WorldWideScience

Sample records for investigating in-vessel cooling

  1. Investigation of vessel exterior air cooling for a HLMC reactor

    International Nuclear Information System (INIS)

    Sienicki, J. J.; Spencer, B. W.

    2000-01-01

    The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink

  2. Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, K.I.; Kim, B.S.; Kim, D.H. [Korea Atomic Energy Research Inst., Thermal Hydraulic Safety Research, Taejon (Korea, Republic of)

    2001-07-01

    A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

  3. Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

    International Nuclear Information System (INIS)

    Ahn, K.I.; Kim, B.S.; Kim, D.H.

    2001-01-01

    A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

  4. Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae

    2016-01-01

    If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study

  5. Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

    2016-10-15

    If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study.

  6. An experimental study on feasibility of ex-vessel cooling through the external guide vessel

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Kim, Jong-Hwan; Park, Rae-Jun; Kim, Sang-Baik

    2000-01-01

    This paper presents the results of a series of experiments for assessing the efficacy of ex-vessel cooling through the external guide vessel during a severe accident. Four tests were performed in the LAVA test facility at KAERI, varying the boundary conditions at the outer surface of the vessel. The first test was a dry condition test conducted without cooling the outside of the vessel. On the other hand, in the second test, the cooling of the vessel surface was produced by gravity-driven forced injection of water along the annular gap of 25 mm between the vessel and the external guide vessel. Water flow rate was about 0.85 kg/s and total mass of available water was 300 kg. For the evaluation of the water flow rate effect, the third test was performed with a pool type cooling in the annulus without any circulation of water. These two external cooling tests were performed under elevated pressure of about 1.6 MPa. Finally, the fourth test was conducted under atmospheric pressure to evaluate the effect of system pressure on boiling heat transfer characteristics. In the dry test and the pool type ex-vessel cooling test performed under atmospheric pressure, the vessel was failed by a melt penetration at about 40 degree upper position from the vessel bottom, which is coincident with the boundary of the Al 2 O 3 /Fe melt separated layers. On the other hand, in both of the ex-vessel cooling tests conducted under elevated pressure of about 1.6 MPa, the vessel didn't fail. Compared with the pool boiling test, the vessel experienced effective cooling due to the inlet flow in the forced flow test. Synthesized the results of the tests, it was shown that the heat removal with ex-vessel cooling through the guide vessel is feasible, but the additional evaluations should be performed to guarantee enough thermal margin. (author)

  7. Integration of cooking and vacuum cooling of carrots in a same vessel

    Directory of Open Access Journals (Sweden)

    Luiz Gustavo Gonçalves Rodrigues

    2012-03-01

    Full Text Available Cooked vegetables are commonly used in the preparation of ready-to-eat foods. The integration of cooking and cooling of carrots and vacuum cooling in a single vessel is described in this paper. The combination of different methods of cooking and vacuum cooling was investigated. Integrated processes of cooking and vacuum cooling in a same vessel enabled obtaining cooked and cooled carrots at the final temperature of 10 ºC, which is adequate for preparing ready-to-eat foods safely. When cooking and cooling steps were performed with the samples immersed in boiling water, the effective weight loss was approximately 3.6%. When the cooking step was performed with the samples in boiling water or steamed, and the vacuum cooling was applied after draining the boiling water, water loss ranged between 15 and 20%, which caused changes in the product texture. This problem can be solved with rehydration using a small amount of sterile cold water. The instrumental textural properties of carrots samples rehydrated at both vacuum and atmospheric conditions were very similar. Therefore, the integrated process of cooking and vacuum cooling of carrots in a single vessel is a feasible alternative for processing such kind of foods.

  8. Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Kim, Min-Hwan; Lee, Won-Jae

    2006-01-01

    The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

  9. Computational study of the mixed cooling effects on the in-vessel retention of a molten pool in a nuclear reactor

    International Nuclear Information System (INIS)

    Kim, Byung Seok; Sohn, Chang Hyun; Ahn, Kwang Il

    2004-01-01

    The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a pressurized water reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure

  10. Experimental study of in-and-ex-vessel melt cooling during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Baik; Yoo, K J; Park, C K; Seok, S D; Park, R J; Yi, S J; Kang, K H; Ham, Y S; Cho, Y R; Kim, J H; Jeong, J H; Shin, K Y; Cho, J S; Kim, D H

    1997-07-01

    After code damage during a severe accident in a nuclear reactor, the degraded core has to be cooled down and the decay heat should be removed in order to cease the accident progression and maintain a stable state. The cooling of core melt is divided into in-vessel and ex-vessel cooling depending on the location of molten core which is dependent on the timing of vessel failure. Since the cooling mechanism varies with the conditions of molten core and surroundings and related phenomena, it contains many phenomenological uncertainties so far. In this study, an experimental study for verification of in-vessel corium cooling and several separate effect experiments for ex-vessel cooling are carried out to verify in- and ex-vessel cooling phenomena and finally to develop the accident management strategy and improve engineered reactor design for the severe accidents. SONATA-IV (Simulation of Naturally Arrested Thermal Attack in Vessel) program is set up for in-vessel cooling and a progression of the verification experiment has been done, and an integral verification experiment of the containment integrity for ex-vessel cooling is planned to be carried out based on the separate effect experiments performed in the first phase. First phase study of SONATA-IV is proof of principle experiment and it is composed of LALA (Lower-plenum Arrested Vessel Attack) experiment to find the gap between melt and the lower plenum during melt relocation and to certify melt quenching and CHFG (Critical Heat Flux in Gap) experiment to certify heat transfer mechanism in an artificial gap. As separate effect experiments for ex-vessel cooling, high pressure melt ejection experiment related to the initial condition for debris layer formation in the reactor cavity, crust formation and heat transfer experiment in the molten pool and molten core concrete interaction experiment are performed. (author). 150 refs., 24 tabs., 127 figs.

  11. Investigating the cooling ability of reactor vessel head injection in the Maanshan PWR using CFD simulation

    International Nuclear Information System (INIS)

    Tseng Yungshin; Lin Chihhung; Wan Jongrong; Shih Chunkuan; Tsai, F. Peter

    2011-01-01

    In order to reduce the crack growth rate on the welding of penetration pipe, Pressurized Water Reactor (PWR) of Maanshan nuclear power plant (NPP) uses vessel head injection to cool vessel lid and control rod driving components. The injection flow from the cold leg is drained by the pressure difference between cold leg and upper internal components. In this study, 10 million meshes model with 4 sub-models have been developed to simulate the thermal-hydraulic behavior by commercial CFD program FLUENT. The results indicate that the injection nozzles can provide good cooling ability to reduce the maximum temperature for lid on the vessel head. The maximum temperature of vessel lid is about 293.81degC. Based on the simulated temperature, ASME CODE N-729-1 was further used to recount the effective degradation years (EDY) and reinspection years (RIY) factors. It demonstrates that the EDY and RIY factors are still less than 1.0. Therefore, the re-inspection period for Maanshan PWR would not be significantly affected by the miner temperature difference. (author)

  12. Thermal Behavior of the Reactor Vessel Penetration Under External Vessel Cooling During a Severe Accident

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Jong-Tae; Min, Byung-Tae; Lee, Ki-Young; Kim, Sang-Baik

    2004-01-01

    Experimental and analytical studies on the thermal behavior of reactor vessel penetration have been performed under external vessel cooling during a severe accident in the Korean next-generation reactor APR1400. Two types of tests, SUS-EXT and SUS-DRY with and without external vessel cooling, respectively, have been performed using sustained heating by an induction heater. Three tests have been carried out varying the cooling conditions at the vessel outer surface in the SUS-EXT tests. The experimental results have been thermally estimated using the LILAC computer code. The experimental results indicate that the inner surface of the vessel was ablated by the 45-mm thickness in the SUS-DRY test. Despite the total ablation of the welding material, the penetration was not ejected outside the vessel, which could be attributed to the thermal expansion of the penetration. Unlike the SUS-DRY test, the thickness of the ablation was ∼15 to 20 mm at most, so the welding was preserved in the SUS-EXT tests. It is concluded from the experimental results that the external vessel cooling highly affected the ablation configuration and the thermal behaviors of the vessel and the penetration. An increase in coolant mass flow rate from 0.047 to 0.152 kg/s had effects on the thermal behavior of the lower head vessel and penetration in the SUS-EXT tests. The LILAC analytical results on temperature distribution and ablation depth in the lower head vessel and penetration were very similar to the experimental results

  13. Cooling system for the connecting rings of a fast neutron reactor vessel

    International Nuclear Information System (INIS)

    Martin, J.-P.; Malaval, Claude

    1974-01-01

    A description is given of a cooling system for the vessel connecting rings of a fast neutron nuclear reactor, particularly of a main vessel containing the core of the reactor and a volume of liquid metal coolant at high temperature and a safety vessel around the main vessel, both vessels being suspended to a rigid upper slab kept at a lower temperature. It is mounted in the annular space between the two vessels and includes a neutral gas circuit set up between the wall of the main vessel to be cooled and that of the safety vessel itself cooled from outer. The neutral gas system comprises a plurality of ventilators fitted in holes made through the thickness of the upper slab and opening on to the space between the two vessels. It also includes two envelopes lining the walls of these vessels, establishing with them small section channels for the circulation of the neutral gas cooled against the safety vessel and heated against the main vessel [fr

  14. System for cooling the containment vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Costes, Didier.

    1982-01-01

    The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

  15. Provision of reliable core cooling in vessel-type boiling reactors

    International Nuclear Information System (INIS)

    Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

    1987-01-01

    Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

  16. Effect of in-core instrumentation mounting location on external reactor vessel cooling

    International Nuclear Information System (INIS)

    Suh, Jungsoo; Ha, Huiun

    2017-01-01

    Highlights: • Numerical simulations were conducted for the evaluation of an IVR-ERVC application. • The ULPU-V experiment was simulated for the validation of numerical method. • The effect of ICI mounting location on an IVR-ERVC application was investigated. • TM-ICI is founded to be superior to BM-ICI for successful application of IVR-ERVC. - Abstract: The effect of in-core instrumentation (ICI) mounting location on the application of in-vessel corium retention through external reactor vessel cooling (IVR-ERVC), used to mitigate severe accidents in which the nuclear fuel inside the reactor vessel becomes molten, was investigated. Numerical simulations of the subcooled boiling flow within an advanced pressurized-water reactor (PWR) in IVR-ERVC applications were conducted for the cases of top-mounted ICI (TM-ICI) and bottom-mounted ICI (BM-ICI), using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. To validate the numerical method for IVR applications, numerical simulations of ULPU-V experiments were also conducted. The BM-ICI reactor vessel was modeled using a simplified design of an advanced PWR with BM-ICI; the TM-ICI counterpart was modeled by removing the ICI parts from the original geometry. It was found that TM-ICI was superior to BM-ICI for successful application of IVR-ERVC. For the BM-ICI case, the flow field was complicated because of the existence of ICIs and a significant temperature gradient was observed near the ICI nozzles on the lower part of the reactor vessel, where the ICIs were attached. These observations suggest that the existence of ICI below the reactor vessel hinders reactor vessel cooling.

  17. System for cooling the upper wall of a nuclear reactor vessel

    International Nuclear Information System (INIS)

    Pailla, Henri; Schaller, Karl; Vidard, Michel.

    1974-01-01

    A system for cooling the upper wall of the main vessel of a fast neutron reactor is described. This vessel is suspended from an upper shield by the upper wall. It includes coils carrying a coolant which are immersed in an intermediate liquid bathing the wall and contained in a tank integral with the vessel. At least one of the two cooling and intermediate liquids is a liquid metal. The main vessel is contained in a safety vessel, the space between the main and safety vessels is occluded in its upper part by an insulating shield placed under the tank. There is a liquid metal seal between the upper wall and the upper shield under the tank. This system has been specially designed for sodium cooled fast neutron reactors [fr

  18. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  19. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2001-03-15

    In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report.

  20. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    International Nuclear Information System (INIS)

    Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin; Park, Jae Hong

    2001-03-01

    In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report

  1. External Reactor Vessel Cooling Evaluation for Severe Accident Mitigation in NPP Krsko

    International Nuclear Information System (INIS)

    Mihalina, M.; Spalj, S.; Glaser, B.

    2016-01-01

    The In-Vessel corium Retention (IVR) through the External Reactor Vessel Cooling (ERVC) is mean for maintaining the reactor vessel integrity during a severe accident, by cooling and retaining the molten material inside the reactor vessel. By doing this, significant portion of severe accident negative phenomena connected with reactor vessel failure could be avoided. In this paper, analysis of NPP Krsko applicability for IVR strategy was performed. It includes overview of performed plant related analysis with emphasis on wet cavity modification, plant's site specific walk downs, new applicable probabilistic and deterministic analysis, evaluation of new possibilities for ERVC strategy implementation regarding plant's post-Fukushima improvements and adequacy with plant's procedures for severe accident mitigation. Conclusion is that NPP Krsko could perform in-vessel core retention by applying external reactor vessel cooling strategy with reasonable confidence in success. Per probabilistic and deterministic analysis, time window for successful ERVC strategy performance for most dominating plant damage state scenarios is 2.5 hours, when onset of core damage is observed. This action should be performed early after transition to Severe Accident Management Guidance's (SAMG). For loss of all AC power scenario, containment flooding could be initiated before onset of core damage within related emergency procedure. To perform external reactor vessel cooling, reactor water storage tank gravity drain with addition of alternate water is needed to be injected into the containment. ERVC strategy will positively interfere with other severe accident strategies. There are no negative effects due to ERVC performance. New flooding level will not threaten equipment and instrumentation needed for long term SAMGs performance and eventually diluted containment sump borated water inventory will not cause return to criticality during eventual recirculation phase due to the

  2. Experimental investigation of creep behavior of reactor vessel lower head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1998-03-01

    The objective of the USNRC supported Lower Head Failure (LHF) Experiment Program at Sandia National Laboratories is to experimentally investigate and characterize the failure of the reactor pressure vessel (RPV) lower head due to the thermal and pressure loads of a severe accident. The experimental program is complemented by a modeling program focused on the development of a constitutive formulation for use in standard finite element structure mechanics codes. The problem is of importance because: lower head failure defines the initial conditions of all ex-vessel events; the inability of state-of-the-art models to simulate the result of the TMI-II accident (Stickler, et al. 1993); and TMI-II results suggest the possibility of in-vessel cooling, and creep deformation may be a precursor to water ingression leading to in-vessel cooling

  3. Optimized design of an ex-vessel cooling thermosyphon for decay heat removal in SFR

    International Nuclear Information System (INIS)

    Choi, Jae Young; Jeong, Yong Hoon; Song, Sub Lee; Chang, Soon Heung

    2017-01-01

    Passive decay heat removal and sodium fire are two major key issues of nuclear safety in sodium-cooled fast reactor (SFR). Several decay heat removal systems (DHR) were suggested for SFR around the world so far. Those DHRS mainly classified into two concepts: Direct reactor cooling system and ex-vessel cooling system. Direct reactor cooling method represented by PDHRS from PGSFR has disadvantages on its additional in-vessel structure and potential sodium fire risk due to the sodium-filled heat exchanger exposed to air. Contrastively, ex-vessel cooling method represented by RVACS from PRISM has low decay heat removal performance, which cannot be applicable to large scale reactors, generally over 1000 MWth. No passive DHRSs which can solve both side of disadvantages has been suggested yet. The goal of this study was to propose ex-vessel cooling system using two-phase closed thermosyphon to compensate the disadvantages of the past DHRSs. Reference reactor was Innovative SFR (iSFR), a pool-type SFR designed by KAIST and featured by extended core lifetime and increased thermal efficiency. Proposed ex-vessel cooling system consisted of 4 trains of thermosyphons and designed to remove 1% of thermal power with 10% of margin. The scopes of this study were design of proposed passive DHRS, validation of system analysis and optimization of system design. Mercury was selected as working fluid to design ex-vessel thermosyphon in consideration of system geometry, operating temperature and required heat flux. SUS 316 with chrome coated liner was selected as case material to resist against high corrosivity of mercury. Thermosyphon evaporator was covered on the surface of reactor vessel as the geometry of hollow shell filled with mercury. Condenser was consisted of finned tube bundles and was located in isolated water pool, the ultimate heat sink. Operation limits and thermal resistance was estimated to guarantee whether the design was adequate. System analysis was conducted by in

  4. Heat dissipation research on the water-cooling channel of HL-2M in-vessel coils

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J., E-mail: jiangjiaming@swip.ac.cn; Liu, Y.; Chen, Q.; Ji, X.Q.

    2017-04-15

    Highlights: • The joule heat of in-vessel coils is very difficult to dissipate inside HL-2M vacuum vessel. • Heat dissipation model of the coil includes the joule heat model, the heat conduction model and the heat transfer model. • The CFD analysis has been done for the coil-water cooling, with comparison with the date of theoretical analysis and experiment. • The result shows water-cooling channel is good for the joule heat transfer and taken away. - Abstract: HL-2M in-vessel coils are positioned in high vacuum circumstance, and they will generate joule heat when they carry 15 kA electrical current, but joule heat is very difficult to dissipate in vacuum, so a hollow cable with 8 mm inner diameter is design as water-cooling channel for heat convection. By using the methods of the theoretical derivation, together with CFD numeric simulation method and the experiment of the heat transfer, the water channel of HL-2M in-vessel coils has been studied, and the temperature of HL-2M in-vessel coils under different cooling water flow rates is obtained and acceptable. Simultaneously, the external cooling water supply system parameters for the water-cooling channel of the coils are estimated. Three methods’ results are in good agreement; the theoretical model is verified and could be popularized for predicting the temperature rise of HL-2M in-vessel coils.

  5. A study of the external cooling capability for the prevention of reactor vessel failure

    Energy Technology Data Exchange (ETDEWEB)

    Chang, S H; Baek, W P; Moon, S K; Yang, S H; Kim, S H [Korea Advanced Institute of Science Technology, Daejeon (Korea, Republic of)

    1994-07-15

    This study (a 3-year program) aims to perform a comprehensive assessment of the feasibility of external vessel flooding with respect to advanced pressurized water reactor plants to be built in Korea. During the first year, review of the relevant phenomena and preliminary assessment of the concept have been performed. Also performed is a review of heat transfer correlations for the computer program that will be developed for assessment of the cooling capability of external vessel flooding. Important phenomena that determine the cooling capability of external vessel flooding are (a) the initial transient before formation of molten corium pool, (b) natural convection of in-vessel molten corium pool, (c) radiative heat exchange between the molten corium pool and the upper vessel structures, (d) thermal hydraulics outside the vessel, (e) structural integrity consideration, and (f) long-term phenomena. The adoption of the concept should be decided by considering several factors such as (a) vessel submergence procedure, (b) cooling requirements, (c) vessel design features, (d) steam production, (e) instrumentation needs, and (f) an overall accident management strategy. The external vessel cooling concept looks to be promising. However, further study is required for a reliable decision making. Several correlations are available for the prediction of cooling capability of the present concept. However, it is difficult to define a sufficiently reliable set of correlations; sensitivity studies would be required in assessing the cooling capability with the computer program.

  6. A feasibility experiment for assessing the efficacy of ex-vessel cooling through the external gap structure

    International Nuclear Information System (INIS)

    Kang, K. H.; Kim, J. H.; Park, L. J.; Kim, S. B.; Hwang, I. S.

    1999-01-01

    This paper presents the results of a feasibility experiment for assessing the efficacy of ex-vessel cooling through the external gap structure during a severe accident. In this study, a 1/8 linear scale mockup of a lower plenum was used with Al2O3/Fe thermite melt as a corium simulant. The results show that in dry case test conducted without cooling the outside of the vessel, after about thirty second from the thermite ignition the vessel was heated to cause a complete melt penetration at about 30 degree upper position from the bottom. Whereas in wet case test conducted cooling the outside of the vessel with 0.85 kg/s of water flow rate using 2.5 cm of uniform gap structure, the vessel effectively cooled down with 23.7 K/s of cooling rate by nucleate boiling at the surface of the vessel. The results of two-dimensional analyses using FLUENT code show a similar trend of vessel thermal behavior presented in the tests. Synthesized the results of the tests and analyses work, a natural convection of the melt pool could cause the formation of hot spot at the upper portion of the vessel, but the vessel could effectively cool down by heat removal with ex-vessel cooling

  7. Conjugate heat transfer analysis for in-vessel retention with external reactor vessel cooling

    International Nuclear Information System (INIS)

    Park, Jong-Woon; Bae, Jae-ho; Song, Hyuk-Jin

    2016-01-01

    Highlights: • A conjugate heat transfer analysis method is applied for in-vessel corium retention. • 3D heat diffusion has a formidable effect in alleviating focusing heat load from metallic layer. • The focusing heat load is decreased by about 2.5 times on the external surface. - Abstract: A conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue for in-vessel retention. The method calculates steady-state three-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel three-layered stratified corium (metallic pool, oxide pool and heavy metal and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel). The three-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method. For the ex-vessel boiling boundary conditions, nucleate, transition and film boiling are considered. The thermal integrity of a reactor vessel is addressed in terms of heat flux at the outer-most nodes of the vessel and remaining thickness profile. The vessel three-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate

  8. 1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Kim, Jae Cheol

    2007-02-01

    During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

  9. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 3. Numerical investigation for thermal stratification phenomena in the upper plenum

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Yamaguchi, Akira

    2002-06-01

    A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermal stratification characteristics in the upper plenum, and to investigate trade-off relations between gas entrainment and thermal stratification phenomena on in-vessel structures for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) Dummy plug insertion to a slit of the upper core structure is one of the effective measures to stabilize the in-vessel flow patterns and to mitigate in-vessel thermal shocks. (2) Though flow guide device such as a baffle ring attached to reactor vessel wall is an effective measure to eliminate impinging jet to dipped plate, rising characteristics of the thermal stratification interface are affected by the baffle ring devise. (3) Thermal stratification characteristics are not influenced very much by the installation of a partial inner barrel to the dipped plate, which is an effective measure to reduce the horizontal flow velocity components at free surface. (4) Labyrinth structures to the gap between the reactor vessel wall and the outer dipped plate have direct effects upon in-vessel thermal shock characteristics including thermal stratification phenomena due to the closing of flow path between the upper plenum and the free surface plenum. (author)

  10. The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

    2001-03-01

    In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary.

  11. The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

    International Nuclear Information System (INIS)

    Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

    2001-03-01

    In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary

  12. Characteristics analysis on a superconductor resonance coil WPT system according to cooling vessel materials in different distances

    International Nuclear Information System (INIS)

    Jeong, In-Sung; Lee, Yu-Kyeong; Choi, Hyo-Sang

    2016-01-01

    Highlights: • WPT using the superconductor coil was needed research for cooling vessel. FRP, bakelite, polystyrene, aluminum, and iron were applied as the cooling vessel material to analyze the WPT distance efficiency. • When the distance between the transmitter and receiver coils was 2000 mm, FRP being used for the cooling vessel made the transmission efficiency higher than any other materials. The efficiency and distance of sending power can be improved in the superconductor coil if the cooling vessel is made with FRP. - Abstract: The interest in wireless power transfer (WPT) that can send power without using wires has been increasing recently. Especially, there is a great interest in the wireless power devices for portable IT devices. The WPT devices that have been developed so far use the magnetic induction method, and they are not active due to their distance problem. A magnetic resonance WPT method was developed and has been actively researched to resolve this problem. A superconductor coil was applied in this study to increase the efficiency of the magnetic resonance WPT. FRP, bakelite, polystyrene, aluminum, and iron were applied as the cooling vessel material to analyze the WPT distance. The distance between the transmitter and receiver coils started from 800 mm and was increased by 200 mm. The reflection coefficient was measured at each distance. As a result, FRP, bakelite, plastic PVC, polystyrene of the reflection coefficient was similar. From among these FRP being used for the cooling vessel made the transmission characteristics higher than any other materials when the distance between the transmitter and receiver coils was 2,000 mm. On the other hand, the reflection coefficient dropped when iron was used. It is estimated based on the experimental results that the wireless power transmission characteristics and distance of sending power can be improved in the superconductor coil if the cooling vessel is made with FRP.

  13. Characteristics analysis on a superconductor resonance coil WPT system according to cooling vessel materials in different distances

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, In-Sung, E-mail: no21park@hanmail.net; Lee, Yu-Kyeong; Choi, Hyo-Sang, E-mail: hyosang@chosun.ac.kr

    2016-11-15

    Highlights: • WPT using the superconductor coil was needed research for cooling vessel. FRP, bakelite, polystyrene, aluminum, and iron were applied as the cooling vessel material to analyze the WPT distance efficiency. • When the distance between the transmitter and receiver coils was 2000 mm, FRP being used for the cooling vessel made the transmission efficiency higher than any other materials. The efficiency and distance of sending power can be improved in the superconductor coil if the cooling vessel is made with FRP. - Abstract: The interest in wireless power transfer (WPT) that can send power without using wires has been increasing recently. Especially, there is a great interest in the wireless power devices for portable IT devices. The WPT devices that have been developed so far use the magnetic induction method, and they are not active due to their distance problem. A magnetic resonance WPT method was developed and has been actively researched to resolve this problem. A superconductor coil was applied in this study to increase the efficiency of the magnetic resonance WPT. FRP, bakelite, polystyrene, aluminum, and iron were applied as the cooling vessel material to analyze the WPT distance. The distance between the transmitter and receiver coils started from 800 mm and was increased by 200 mm. The reflection coefficient was measured at each distance. As a result, FRP, bakelite, plastic PVC, polystyrene of the reflection coefficient was similar. From among these FRP being used for the cooling vessel made the transmission characteristics higher than any other materials when the distance between the transmitter and receiver coils was 2,000 mm. On the other hand, the reflection coefficient dropped when iron was used. It is estimated based on the experimental results that the wireless power transmission characteristics and distance of sending power can be improved in the superconductor coil if the cooling vessel is made with FRP.

  14. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 1. Numerical investigation for the rationalization of hydrodynamics in the upper plenum

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Yamaguchi, Akira

    2002-02-01

    A large-scale sodium-cooled fast breeder reactor in feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for a gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, a thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermohydraulic characteristics in the upper plenum, and to investigate suitable in-vessel structure for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) It is difficult to rationale in-vessel flow patterns through adjustments of porous ratio and pressure loss for a hold down plate and baffle plates installed in an upper core structure. (2) Dummy plug insertion to a slit of the upper core structure is one of effective measures to stabilize in-vessel flow patterns. (3) Flow guide devices such as a baffle ring and a partial inner barrel are also effective measures to eliminate impinging jet to a dipped plate (D/P) and to reduce horizontal flow velocity components at free surface. (4) Installations of labyrinth structures to a R/V - D/P gap is successful for decreasing of free surface horizontal flows. (5) Gap closing of an in-vessel fuel pot and two cold trap components has the effects of reductions for free surface horizontal flows and for the difference of free surface levels. Following future investigations are important preventive measures against the gas entrainment from the free surface. (1) Flattening of spatial axial velocity distributions at the R/V - D/P gap. (2) Alleviation measures of vortex concentration at free surface. (3) Separation measures of 3-dimensional vortex

  15. Prediction of thermal margin for external cooling of reactor vessel lower head during a severe accident

    International Nuclear Information System (INIS)

    Yoon, Ho Jun; Suh, Kune Y.

    1998-01-01

    In the TMI-2 accident, approximately nineteen (19) tons of molten core material drained into the lower plenum. One of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 .deg. C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident management strategies. As an advanced in-vessel design concept, the COrium Attak Syndrome Immunization Structures (COASIS) are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in -vessel (COASISI) and ex-vessel (COASISO) were demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the TMI-2 and the Korean Standard Nuclear Power Plant (KSNPP) reactors. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. In studying the in-vessel severe accident phenomena, one of the main goals is to verify the cooling mechanism in the reactor vessel lower plenum and thereby to prevent the vessel failure from thermal attack by the molten debris. This paper presents the first-principle calculation results for the thermal margin for the case of external cooling of the reactor vessel lower head. Adopting the method presented by F.B. Cheung, et al., we calculated the departure from nucleate boiling ratio (DNBR) for the three cases of pool boiling, flow boiling

  16. An experimental study on in-vessel debris retention through gap cooling

    International Nuclear Information System (INIS)

    Kang, Kyung Ho; Kim, Jong Whan; Cho, Young Ro; Chang, Young Cho; Park, Rae Jun; Gu, Kil Mo; Kim, Sang Baik; Kim, Hee Dong

    1999-04-01

    LAVA experiments have been performed using high temperature molten material to be relocated into the 1/8 linear scaled vessel of a reactor lower plenum filled with water. An Al 2 O 3 /Fe tehrmite melt (Al 2 O 3 only) was used as a corium simulant. In this study, the influence of various initial conditions, such as internal pressure load across the vessel wall, the material composition of the melt simulant, water subcooling and depth on gap formation were investigated. As well, the thermal and mechanical behaviors of the vessel were examined. In case the internal pressure load was imposed, the gap formation between the continuous solidified debris and the vessel wall was clearly shown with post-test examination. On the other hand, in case the internal pressure load was not imposed, the iron welded to the inner surface of the vessel and the vessel experienced ablation to about 5 mm. The cooling rate of the vessel was very slow in the tests using Al 2 O 3 /Fe thermite melt but it was rather fast using Al 2 O 3 melt. It is postulated that in the Al 2 O 3 /Fe thermite tests, the iron melt layer is so dense that water ingression into the gap is difficult due to the high pressure of escaping steam. On the other hand, in the porosity of an Al 2 O 3 melt layer could enhance water ingression into the gap by giving the flow path of the evaporated steam through the porous media. The water height and subcooling could affect the melt pool formation and the initial thermal attack to the vessel. However, at present stage, the effect of water subcooling on the thermal behavior of the vessel couldn't be generalized. For clear confirmation of the effect of water subcooling, the tests will be performed at the saturated water condition. (author). 19 refs., 3 tabs., 36 figs

  17. A state of the art on penetration failure estimation under external vessel cooling

    International Nuclear Information System (INIS)

    Min, B. T.; Park, R. J.; Kang, K. H.; Cho, Y. R.; Kim, J. W.; Kim, S. B.; Park, S. Y.; Lee, K. Y.

    2000-04-01

    A state of the art on penetration failure was reviewed and analyzed to establish the direction of the experimental program in the KNGR and to decide the test section design. The interaction between the corium and the reactor vessel and the corium behavior in the lower plenum of the reactor vessel were analyzed to investigate the penetration effect on severe accident progression, and the TMI-2 accident was investigated in the point of penetration failure. Theoretical model and experiment results on penetration failure under the severe accident were investigated and reviewed to establish the direction of the experimental program on the estimation of the penetration failure in the KNGR. These results were compared with the TMI-2 results. The existing test facilities on penetration failure were investigated and reviewed to decide the test section design. It can be said from the state of the art review that penetration in the lower plenum of the reactor vessel is a week point in the reactor vessel failure under the severe accident, but the reactor vessel may not be failed by penetration failure in condition with the coolant supply to the penetration. Since the penetration is different with reactor types and there is no study on estimation of the penetration welding, it is necessary to investigate failure or not of the penetration in condition with external vessel cooling to maintain the reactor vessel integrity in KNGR. In the present experimental program on the integrity estimation of the KNGR penetration, the aluminum oxide melt by thermite reaction and the test section with one penetration of the real size and real material were selected. The melt mass, the pressure of the system, and the vessel geometry were selected as an experimental parameter. (author)

  18. Integrated conjugate heat transfer analysis method for in-vessel retention with external reactor vessel cooling - 15477

    International Nuclear Information System (INIS)

    Park, J.W.; Bae, J.H.; Seol, W.C.

    2015-01-01

    An integrated conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue. The method calculates steady-state 3-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel 3-layered stratified corium (metallic pool, oxide pool and heavy metal) and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel. The 3-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method and ex-vessel boiling regimes are parametrically considered. The thermal integrity of a reactor vessel is addressed in terms of un-molten thickness profile. The vessel 3-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate inside the oxide pool and the 3-dimensional vessel heat transfer provides a much larger minimum vessel wall thickness. (authors)

  19. Study on external reactor vessel cooling capacity for advanced large size PWR

    International Nuclear Information System (INIS)

    Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

    2014-01-01

    External reactor vessel cooling (ERVC) is widely adopted as a part of in- vessel retention (IVR) in severe accident management strategies. In this paper, some flow parameters and boundary conditions, eg., inlet and outlet area, water inlet temperature, heating power of the lower head, the annular gap size at the position of the lower head and flooding water level, were considered to qualitatively study the effect of them on natural circulation capacity of the external reactor vessel cooling for an advanced large size PWR by using RELAP5 code. And the calculation results provide some basis of analysis for the structure design and the following transient response behavior of the system. (authors)

  20. In-Vessel Retention via External Reactor Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Bachrata, Andrea [CTU in Prague, Faculty of nuclear sciences and physical engineering, V Holesovickach 2 180 00, Prague 8 (Czech republic)

    2008-07-01

    In-vessel (corium) retention (IVR) via external reactor pressure vessel (RPV) cooling is considered to be an effective severe accident management strategy for corium localisation and stabilisation. The main idea of IVR strategy consists in flooding the reactor cavity and transferring the decay heat through the wall of RPV to the recirculating water and than to the atmosphere of the containment of nuclear power plant. The aim of this strategy is to localise and to stabilise the corium inside the RPV. Not using this procedure could destroy the integrity of RPV and might cause the interaction of the corium with the concrete at the bed of the reactor cavity. Several experimental facilities and computer codes (MVITA, ASTEC module DIVA and CFD codes) were applied to simulate the IVR strategy for concrete reactor designs. The necessary technical modifications concerning the implementation of IVR concept were applied at the Loviisa NPP (VVER-440/V213). This strategy is also an important part of the advanced reactor designs AP600 and AP1000. (authors)

  1. An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

    2004-01-01

    LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

  2. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    Energy Technology Data Exchange (ETDEWEB)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  3. Design of the prestressed concrete reactor vessel for gas-cooled heating reactors

    International Nuclear Information System (INIS)

    Becker, G.; Notheisen, C.; Steffen, G.

    1987-01-01

    The GHR pebble bed reactor offers a simple, safe and economic possibility of heat generation. An essential component of this concept is the prestressed concrete reactor vessel. A system of cooling pipes welded to the outer surface of the liner is used to transfer the heat from the reactor to the intermediate circuit. The high safety of this vessel concept results from the clear separation of the functions of the individual components and from the design principle of the prestressed conncrete. The prestressed concrete structure is so designed that failure can be reliably ruled out under all operating and accident conditions. Even in the extremely improbable event of failure of all decay heat removal systems when decay heat and accumulated heat are transferred passively by natural convection only, the integrity of the vessel remains intact. For reasons of plant availability the liner and the liner cooling system shall be designed so as to ensure safe elimination of failure over the total operating life. The calculations which were peformed partly on the basis of extremely adverse assumption, also resulted in very low loads. The prestressed concrete vessel is prefabricated to the greatest possible extent. Thus a high quality and optimized fabrication technology can be achieved especially for the liner and the liner cooling system. (orig./HP)

  4. Latest developments in prestressed concrete vessels for gas-cooled reactors

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.

    1979-01-01

    This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

  5. Heating and cooling system for an on-board gas adsorbent storage vessel

    Science.gov (United States)

    Tamburello, David A.; Anton, Donald L.; Hardy, Bruce J.; Corgnale, Claudio

    2017-06-20

    In one aspect, a system for controlling the temperature within a gas adsorbent storage vessel of a vehicle may include an air conditioning system forming a continuous flow loop of heat exchange fluid that is cycled between a heated flow and a cooled flow. The system may also include at least one fluid by-pass line extending at least partially within the gas adsorbent storage vessel. The fluid by-pass line(s) may be configured to receive a by-pass flow including at least a portion of the heated flow or the cooled flow of the heat exchange fluid at one or more input locations and expel the by-pass flow back into the continuous flow loop at one or more output locations, wherein the by-pass flow is directed through the gas adsorbent storage vessel via the by-pass line(s) so as to adjust an internal temperature within the gas adsorbent storage vessel.

  6. Active cooling system for Tokamak in-vessel operation manipulator

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Jianjun, E-mail: yuanjj@sjtu.edu.cn; Chen, Tan; Li, Fashe; Zhang, Weijun; Du, Liang

    2015-10-15

    Highlights: • We summarized most of the challenges of fusion devices to robot systems. • Propose an active cooling system to protect all of the necessary components. • Trial design test and theoretical analysis were conducted. • Overall implementation of the active cooling system was demonstrated. - Abstract: In-vessel operation/inspection is an indispensable task for Tokamak experimental reactor, for a robot/manipulator is more capable in doing this than human being with more precise motion and less risk of damaging the ambient equipment. Considering the demanding conditions of Tokamak, the manipulator should be adaptable to rapid response in the extreme conditions such as high temperature, vacuum and so on. In this paper, we propose an active cooling system embedded into such manipulator. Cameras, motors, gearboxes, sensors, and other mechanical/electrical components could then be designed under ordinary conditions. The cooling system cannot only be a thermal shield since the components are also heat sources in dynamics. We carry out a trial test to verify our proposal, and analyze the active cooling system theoretically, which gives a direction on the optimization by varying design parameters, components and distribution. And based on thermal sensors monitoring and water flow adjusting a closed-loop feedback control of temperature is added to the system. With the preliminary results, we believe that the proposal gives a way to robust and inexpensive design in extreme environment. Further work will concentrate on overall implementation and evaluation of this cooling system with the whole inspection manipulator.

  7. Synthetic analyses of the LAVA experimental results on in-vessel corium retention through gap cooling

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Cho, Young Ro; Koo, Kil Mo; Park, Rae Joon; Kim, Jong Hwan; Kim, Jong Tae; Ha, Kwang Sun; Kim, Sang Baik; Kim, Hee Dong

    2001-03-01

    LAVA(Lower-plenum Arrested Vessel Attack) has been performed to gather proof of gap formation between the debris and lower head vessel and to evaluate the effect of the gap formation on in-vessel cooling. Through the total of 12 tests, the analyses on the melt relocation process, gap formation and the thermal and mechanical behaviors of the vessel were performed. The thermal behaviors of the lower head vessel were affected by the formation of the fragmented particles and melt pool during the melt relocation process depending on mass and composition of melt and subcooling and depth of water. During the melt relocation process 10.0 to 20.0 % of the melt mass was fragmented and also 15.5 to 47.5 % of the thermal energy of the melt was transferred to water. The experimental results address the non-adherence of the debris to the lower head vessel and the consequent gap formation between the debris and the lower head vessel in case there was an internal pressure load across the vessel abreast with the thermal load induced by the thermite melt. The thermal behaviors of the lower head vessel during the cooldown period were mainly affected by the heat removal characteristics through this gap, which were determined by the possibilities of the water ingression into the gap depending on the melt composition of the corium simulant. The enhanced cooling capacity through the gap was distinguished in the Al 2 O 3 melt tests. It could be inferred from the analyses on the heat removal capacity through the gap that the lower head vessel could effectively cooldown via heat removal in the gap governed by counter current flow limits(CCFL) even if 2mm thick gap should form in the 30 kg Al 2 O 3 melt tests, which was also confirmed through the variations of the conduction heat flux in the vessel and rapid cool down of the vessel outer surface in the Al 2 O 3 melt tests. In the case of large melt mass of 70 kg Al 2 O 3 melt, however, the infinite possibility of heat removal through the

  8. Analysis of three ex-vessel loss-of-coolant accidents in the first wall cooling system of NET/ITER

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1993-01-01

    An ex-vessel LOCA may be caused by a rupture of a cooling pipe located outside the vacuum vessel. No plasma shutdown and no other counteractions have been assumed in order to study the worst case conditions of the accidents. The next three ex-vessel LOCAs in the primary cooling system of the first wall have been analysed: 1. a large break ex-vessel LOCA caused by a rupture of the cold leg (inner diameter 0.314 m) of the main circuit; 2. an intermediate break ex-vessel LOCA caused by a rupture of a sector inlet feeder (inner diameter 0.158 m); 3. an intermediate break ex-vessel LOCA caused by a rupture of the surge line (inner diameter 0.180 m) of the pressurizer. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the first two scenarios, melting in the first wall starts about 90 s after break initiation. In the third scenario, melting in the first wall start about 323 s after break initiation. Special emphasis has been paid to the characteristics of the break flows, the transient thermal-hydraulic behaviour of the cooling system, and the temperature development in the first wall. (orig.)

  9. Effectiveness of External Reactor Vessel Cooling (ERVC) strategy for APR1400 and issues of phenomenological uncertainties

    International Nuclear Information System (INIS)

    Oh, S.J.; Kim, H.T.

    2007-01-01

    The APR1400(Advanced Power Reactor 1400) is an evolutionary advanced light water reactor with rated thermal power of 4000 MWt. For APR1400, External Reactor Vessel Cooling (ERVC) is adopted as a primary severe accident management strategy for in-vessel retention (IVR) of corium. The ERVC is a method of IVR by submerging the reactor vessel exterior. At the early stage of the APR1400 design, only ex-vessel cooling, cooling of the core melt outside the vessel after vessel is breached, is considered based on the EPRI Utility Requirement Document for Evolutionary LWR. However, based on the progress in implementation of Severe Accident Management Guidance (SAMG) for operating plants, as well as the research findings related to ERVC, ERVC strategy is adopted as a part of key severe accident management strategies. To improve its success, the strategy is reviewed and we implemented necessary design arrangement to increase its usefulness in managing the severe accident. In this paper, we examine the evolution of ERVC concept and its implementation in APR1400. Then, we review possible approach, including Risk-Oriented Accident Analysis Methodology (ROAAM), to evaluate the effectiveness of the strategy. (authors)

  10. Data on test results of vessel cooling system of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2003-02-01

    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

  11. Limiting Factors for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Cheung, F.B.

    2005-01-01

    The method of external reactor vessel cooling (ERVC) that involves flooding of the reactor cavity during a severe accident has been considered a viable means for in-vessel retention (IVR). For high-power reactors, however, there are some limiting factors that might adversely affect the feasibility of using ERVC as a means for IVR. In this paper, the key limiting factors for ERVC have been identified and critically discussed. These factors include the choking limit for steam venting (CLSV) through the bottleneck of the vessel/insulation structure, the critical heat flux (CHF) for downward-facing boiling on the vessel outer surface, and the two-phase flow instabilities in the natural circulation loop within the flooded cavity. To enhance ERVC, it is necessary to eliminate or relax these limiting factors. Accordingly, methods to enhance ERVC and thus improve margins for IVR have been proposed and demonstrated, using the APR1400 as an example. The strategy is based on using two distinctly different methods to enhance ERVC. One involves the use of an enhanced vessel/insulation design to facilitate steam venting through the bottleneck of the annular channel. The other involves the use of an appropriate vessel coating to promote downward-facing boiling. It is found that the use of an enhanced vessel/insulation design with bottleneck enlargement could greatly facilitate the process of steam venting through the bottleneck region as well as streamline the resulting two-phase motions in the annular channel. By selecting a suitable enhanced vessel/insulation design, not only the CLSV but also the CHF limits could be significantly increased. In addition, the problem associated with two-phase flow instabilities and flow-induced mechanical vibration could be minimized. It is also found that the use of vessel coatings made of microporous metallic layers could greatly facilitate downward-facing boiling on the vessel outer surface. With vessel coatings, the local CHF limits at

  12. Vessel supporting structure for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Mahe, Armel; Jullien, Georges

    1974-01-01

    The supporting structure described is for a liquid metal cooled nuclear reactor, the vessel being of the type suspended to the end slab of the reactor. It includes a ring connected at one of its two ends to a single shell and at the other end to two shells. One of these three shells connected to the lower end of the ring forms the upper part of the vessel to be supported. The two other shells are embedded in two sperate parts of the slab. The ring and shell assembly is housed in an annular space provided in the end slab and separating it into two parts, namely a central part and a peripheral part [fr

  13. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Yun, Sunghwan; Kim, Sang Ji

    2015-01-01

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

  14. Relocation work of temporary thermocouples for measuring the vessel cooling system in the safety demonstration test

    International Nuclear Information System (INIS)

    Shimazaki, Yosuke; Shinohara, Masanori; Ono, Masato; Yanagi, Shunki; Tochio, Daisuke; Iigaki, Kazuhiko

    2012-05-01

    It is necessary to confirm that the temperature of water cooling panel of the vessel cooling system (VCS) is controlled under the allowable working temperature during the safety demonstration test because the water cooling panel temperature rises due to stop of cooling water circulation pumps. Therefore, several temporary thermocouples are relocated to the water cooling panel near the stabilizers of RPV and the side cooling panel outlet ring header of VCS in order to observe the temperature change of VCS. The relocated thermocouples can measure the temperature change with starting of the cooling water circulation pumps of VCS. So it is confirmed that the relocated thermocouples can observe the VCS temperature change in the safety demonstration test. (author)

  15. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

    2005-01-01

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods

  16. An Approach for Selection of Flow Regime and Models for Conservative Evaluation of a Vessel Integrity Monitoring System for Water-Cooled Vacuum Vessels

    International Nuclear Information System (INIS)

    Pointer, W. David; Ruggles, Arthur E.

    2003-01-01

    Thin-walled vacuum containment vessels cooled by circulating water jackets are often utilized in research and industrial applications where isolation of equipment or experiments from the influences of the surrounding environment is desirable. The development of leaks in these vessels can result in costly downtime for the facility. A Vessel Integrity Monitoring System (VIMS) is developed to detect leak formation and estimate the size of the leak to allow evaluation of the risk associated with continued operation. A wide range of leak configurations and fluid flow phenomena are considered in the evaluation of the rate at which a tracer gas dissolved in the cooling jacket water is transported into the vacuum vessel. A methodology is presented that uses basic fluid flow models and careful evaluation of their ranges of applicability to provide a conservative estimate of the transport rates for the tracer gas and hence the time required for the VIMS to detect a leak of a given size

  17. Development and investigation of the prestressed reinforced concrete vessels for the water cooled reactors in the FRG

    International Nuclear Information System (INIS)

    Medovikov, A.I.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Konevskij, V.N.

    1980-01-01

    An analysis of calculation results for characteristics of stress-strained state of reactor vessel made of prestressed reinforced concrete is presented. Experimental data obtained during the investigation into a model of reactor vessel top cover are given. Thermal shielding system both for boiling water and pressurized-water reactors has been considered and its working capacity has been evaluated. An analysis of experimental data show correctness of the method assumed for calculation of the reactor top cover which permits to exactly determine its stressed-strained state as well as the nature of crack propagation in the vessel and the structure supporting power. Ceramics is suggested to be used as a heat-insulating material

  18. Slurry Ice as a Cooling System on 30 GT Fishing Vessel

    Directory of Open Access Journals (Sweden)

    Alam Baheramsyah

    2017-06-01

    Full Text Available Indonesia is the largest archipelago country in the world that has a sea area that is very spacious. Indonesian sea area is 5.8 million square kilometers and a coastline of 95 181 km has huge potential in the fisheries sector. In line with the need to further improve on the quality of the fish catch. One way to preserve fish is to use a slurry of ice. Slurry ice proved more effective preserving fishery products instead of using ice cubes. Ice slurry cooling system was designed and applied to the fishing vessel 30 GT. The cooling system uses a simple vapor compression system consists of five major components consisting of evaporator, condenser, compressor, and two pumps.In designing this system determined the type of refrigerant used in advance which type of refrigerant R-507a. Then do the design or selection of its main components. The design is only done on the evaporator. As for the other major components such as condensers, compressors, and pumps election in accordance with the specification of the power needed. After that dialakukan depiction of each system component. Then subsequently designing the laying of ice slurry cooling system components on a fishing vessel 30 GT.            Through calculations using simple thermodynamic equations obtained cooling load on this system amounted to 32.06 kW. Condenser with a power of 40 kW. Compressor with power 12 kW. Pump with capacity 10 m3 / h. With memepertimbangkan space left on the ship in the ice slurry system design on the main deck of the ship to the efficient use of space on board. The power requirements of the generator vessel increases due to the addition of ice slurry system components therefore do replacement generator into the generator with a power of 100 kW and penambahn fuel tank to 6,000 L.

  19. A study on corium melt pool behavior under external vessel cooling : investigation of the first phase research results in the OECD RASPLAV project

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Kim, Sang Baik; Kim, Hee Dong; Yoo, Kun Joong

    1998-04-01

    The scope and contents of the OECD RASPLAV program are to investigate natural convection heat transfer in the corium, chemical and mechanical interaction between the corium and the reactor vessel, crust formation of the corium, and thermal behaviour of the corium by experiments and model development during external vessel cooling to prevent reactor vessel failure in severe accidents of nuclear power plant. This study includes evaluation and analysis of the RASPLAV V phase I results for three years between July 1, 1994 and June 30, 1997. These results supply technical basis for our experimental program on severe accident research. Two large-scale experiments of RASPLAV-AW-between the corium and the reactor vessel. Several small-scale experiments were conducted to analyze thermal stratification in the corium. The salt experiments were conducted to estimate the crust and the mushy region formation, and natural convection heat transfer in the corium. In the analytical studies, pre and post analysis of the RASPLAV-AW-200 experiments and evaluation of the salt test results have been performed using CONV 2 and 3D computer codes, which were developed during RASPLAV program phase I. Low density corium was separated from the high density corium during the RASPLAV-AW-200 tests and the TULPAN test, which was a new finding in the RASPLAV project phase I. From the salts test, heat flux distribution in the side wall heating case is similar to the direct internal heat generation case, and the crust formation is a little effect on heat transfer rate. The results of CONV 2 and 3 D were very well with with the experimental results. The results of RASLAV project phase I, such as furnace design and the techniques on fuel melting, are very helpful to our severe accident experimental program. (author). 57 refs., 13 tabs., 52 figs.

  20. Feasibility Study on Two-phase Thermosiphon for External Vessel Cooling Application of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Young; Song, Sub Lee; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    This study shows that ex-vessel cooling by two-phase thermosiphon is feasible for large size of SFR. The result presents that further studies to increase heat transfer on condenser-air and gap is necessary and the experiment should be conducted for the validation. Also, the heat loss through evaporator during normal operation, corrosion, consideration of organic fluid to exclude the poison of mercury should be studied. As the necessity of sodium fast reactor in order to reduce spent fuel, the development of designing sodium fast reactor becomes an issue. Even though there is PDRC and RVACS for the decay heat removal (DHR) system, each system has disadvantage of sodium fire and low performance, respectively. Therefore, to increase the safety of SFR, the new passive safety system design is needed without sodium fire and high performance, which can applied for large SFR. The DHR system using two-phase thermosiphon for external vessel cooling application is suggested in this paper. The proposed design have advantage that there is no structure in reactor vessel, which means no system modification and no sodium fire with perfect isolation. Also, it provide the method to mitigate sodium fire in case of sodium leakage from reactor vessel.

  1. In-vessel core debris retention experiments. Final report

    International Nuclear Information System (INIS)

    1998-10-01

    The in-vessel cooling experimental program (Phase 1 and 2) was motivated by the survivability of the TMI lower vessel head during the TMI-2 accident. During that accident, molten debris relocation into the water filled lower head resulted in a localized hot spot in the lower head, but no lower head failure occurred. A postulated set of mechanisms which could be involved in and responsible for the survivability of the TMI lower head were identified and experimentally investigated as part of this program. These mechanisms included: the formation of a gap (contact resistance) between the relocated and frozen debris and the vessel wall was a key aspect of the in-vessel cooling mechanism; wall heatup due to the relocated debris in the presence of wall stress due to a pressure gradient across the vessel wall; gap growth due to a lack of debris adherence to the vessel wall and material creep of the heated vessel wall; and the potential for enhanced wall cooling due to gap growth. Each of these postulated mechanisms was investigated in this experimental program. This report summarizes the several insights and conclusions that were obtained from this experimental program. This report documents the entire set of five experiments completed in Phase 2 of this experimental program. Results from the Phase 1 effort were used to plan and select the Phase 2 test matrix. Conclusions from the Phase 1 and 2 experiments are identified and recommendations for future work are provided

  2. Assessing the feasibility of a high-temperature, helium-cooled vacuum vessel and first wall for the Vulcan tokamak conceptual design

    International Nuclear Information System (INIS)

    Barnard, H.S.; Hartwig, Z.S.; Olynyk, G.M.; Payne, J.E.

    2012-01-01

    The Vulcan conceptual design (R = 1.2 m, a = 0.3 m, B 0 = 7 T), a compact, steady-state tokamak for plasma–material interaction (PMI) science, must incorporate a vacuum vessel capable of operating at 1000 K in order to replicate the temperature-dependent physical chemistry that will govern PMI in a reactor. In addition, the Vulcan divertor must be capable of handling steady-state heat fluxes up to 10 MW m −2 so that integrated materials testing can be performed under reactor-relevant conditions. A conceptual design scoping study has been performed to assess the challenges involved in achieving such a configuration. The Vulcan vacuum system comprises an inner, primary vacuum vessel that is thermally and mechanically isolated from the outer, secondary vacuum vessel by a 10 cm vacuum gap. The thermal isolation minimizes heat conduction between the high-temperature helium-cooled primary vessel and the water-cooled secondary vessel. The mechanical isolation allows for thermal expansion and enables vertical removal of the primary vessel for maintenance or replacement. Access to the primary vessel for diagnostics, lower hybrid waveguides, and helium coolant is achieved through ∼1 m long intra-vessel pipes to minimize temperature gradients and is shown to be commensurate with the available port space in Vulcan. The isolated primary vacuum vessel is shown to be mechanically feasible and robust to plasma disruptions with analytic calculations and finite element analyses. Heat removal in the first wall and divertor, coupled with the ability to perform in situ maintenance and replacement of divertor components for scientific purposes, is achieved by combining existing helium-cooled techniques with innovative mechanical attachments of plasma facing components, either in plate-type helium-cooled modules or independently bolted, helium-jet impingement-cooled tiles. The vacuum vessel and first wall design enables a wide range of potential PFC materials and configurations to

  3. Experimental study on in-vessel debris coolability during severe accident

    International Nuclear Information System (INIS)

    Kim, S. B.; Park, R. J.; Kim, H. D.

    2002-05-01

    A research program, called SONATA-IV(Simulation of Naturally Arrested Thermal Attack In-Vessel), has been performed to verify the gap cooling mechanism of corium in the lower plenum, and to develop management and mitigation strategies under severe accident conditions. For the proof-of-principles experiment, the LAVA(Lower-plenum Arrested Vessel Attack) experiments have been performed to gather proof of gap formation and to evaluate the gap effect on in-vessel cooling, using Al 2 O 3 /Fe (or Al 2 O 3 only) thermite melt as corium simulant. And also the CHFG(Critical Heat Flux in Gap) experiments have been performed to measure the critical power and to investigate the inherent cooling mechanism in the hemispherical narrow gap. In addition to the experiments, LILAC code was developed to analyze and predict the thermo-hydraulic phenomena of the corium relocated in the reactor lower plenum. It could be found from the LAVA and CHFG experimental results that continuous gap ranged from 1 to 5 mm was formed and that maximum heat removal capacity through a gap is a key factor in determining the potentials of the integrity of the vessel. After all the possibility of IVR(In-Vessel corium Retention) through gap cooling highly depends on the melt relocated into the lower plenum and the gap size. So, feasibility experiments have been performed for the assessment of improved IVR concepts using an internal engineered gap device and a dual strategy of In/Ex-vessel cooling using the LAVA facility. It is preliminarily concluded that these cooling measures lead to an enhanced cooling of the corium in the lower plenum of the reactor vessel. The additional studies will be performed to verify the quantitative heat removal capacity for these cooling measures in the 2nd phase of mid- and long term project period

  4. 78 FR 8698 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel COOL BEANS; Invitation for...

    Science.gov (United States)

    2013-02-06

    ... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013 0005] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel COOL BEANS; Invitation for Public Comments AGENCY... BEANS is: Intended Commercial Use of Vessel: Sightseeing and sunset cruises. Geographic Region: Florida...

  5. Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

    International Nuclear Information System (INIS)

    Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

    2012-01-01

    Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

  6. Development of TPNCIRC code for Evaluation of Two-Phase Natural Circulation Flow Performance under External Reactor Vessel Cooling Conditions

    International Nuclear Information System (INIS)

    Choi, A-Reum; Song, Hyuk-Jin; Park, Jong-Woon

    2015-01-01

    During a severe accident, corium is relocated to the lower head of the nuclear reactor pressure vessel (RPV). Design concept of retaining the corium inside a nuclear reactor pressure vessel (RPV) through external cooling under hypothetical core melting accidents is called external reactor vessel cooling (ERVC). In this respect, validated two-phase natural circulation flow (TPNC) model is necessary to determine the adequacy of the ERVC design and operating conditions such as inlet area, form losses, gap distance, riser length and coolant conditions. The most important model generally characterizing the TPNC are void fraction and two-phase friction factors. Typical experimental and analytical studies to be referred to on two-phase circulation flow characteristics are those by Reyes, Gartia et al. based on Vijayan et al., Nayak et al. and Dubey et al. In the present paper, two-phase natural circulation (TPNC) flow characteristics under external reactor vessel cooling (ERVC) conditions are studied using two existing TPNC flow models of Reyes and Gartia et al. incorporating more improved void fraction and two-phase friction models. These models and correlations are integrated into a computer program, TPNCIRC, which can handle candidate ERVC design parameters, such as inlet, riser and downcomer flow lengths and areas, gap size between reactor vessel and surrounding insulations, minor loss factors and operating parameters of decay power, pressure and subcooling. Accuracy of the TPNCIRC program is investigated with respect to the flow rate and void fractions for existing measured data from a general experiment and ULPU specifically designed for the AP1000 in-vessel retention. Also, the effect of some important design parameters are examined for the experimental and plant conditions. Using the flow models and correlations are integrated into a computer program, TPNCIRC, a number of correlations have been examined. This seems coming from the differences of void fractions

  7. Analytical prediction of the heat transfer from a blood vessel near the skin surface when cooled by a symmetrical cooling strip

    Science.gov (United States)

    Chato, J. C.; Shitzer, A.

    1971-01-01

    An analytical method was developed to estimate the amount of heat extracted from an artery running close to the skin surface which is cooled in a symmetrical fashion by a cooling strip. The results indicate that the optimum width of a cooling strip is approximately three times the depth to the centerline of the artery. The heat extracted from an artery with such a strip is about 0.9 w/m-C which is too small to affect significantly the temperature of the blood flow through a main blood vessel, such as the carotid artery. The method is applicable to veins as well.

  8. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru; Kurita, Gen-ichi; Onozuka, Masaki; Suzuki, Masaru.

    1997-01-01

    Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and γ rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

  9. Vacuum vessel for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Mitsuru; Kurita, Gen-ichi [Japan Atomic Energy Research Inst., Tokyo (Japan); Onozuka, Masaki; Suzuki, Masaru

    1997-07-31

    Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and {gamma} rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

  10. In-vessel core melt retention by RPV external cooling for high power PWR. MAAP 4 analysis on a LBLOCA scenario without SI

    International Nuclear Information System (INIS)

    Cognet, C.; Gandrille, P.

    1999-01-01

    In-, ex-vessel reflooding or both simultaneously can be envisaged as Accident Management Measures to stop a Severe Accident (SA) in vessel. This paper addresses the possibility of in-vessel core melt retention by RPV external flooding for a high power PWR (4250 MWth). The reactor vessel is assumed to have no lower head penetration and thermal insulation is neglected. The effects of external cooling of high power density debris, where the margin for such a strategy is low, are investigated with the MAAP4 code. MAAP4 code is used to verify the system capability to flood the reactor pit and to predict simultaneously the corium relocation into the lower head with the thermal and mechanical response of the RPV in transient conditions. The corium pool cooling and holding in the RPV lower head is analysed. Attention is paid to the internal heat exchanges between corium components. This paper focuses particularly the heat transfer between oxidic and metallic phases as well as between the molten metallic phase and the RPV wall of utmost importance for challenging the RPV integrity in vicinity of the metallic phase. The metal segregation has a decisive influence upon the attack of the vessel wall due to a very strong peaking of the lateral flux ('focusing effect'). Thus, the dynamics of the formation of the metallic layer characterized by a growing inventory of steel, both from a partial vessel ablation and the degradation of internals steel structures by the radiative heat flux from the debris, is displayed. The analysed sequence is a surge line rupture near the hot leg (LBLOCA) leading to the fastest accident progression

  11. Conceptual design for Japan Sodium-Cooled Fast Reactor. (4) Developmental study of steel plate reinforced concrete containment vessel for JSFR

    International Nuclear Information System (INIS)

    Hosoya, Takusaburo; Negishi, Kazuo; Satoh, Kenichiro; Somaki, Takahiro; Matsuo, Ippei; Shimizu, Katsusuke

    2009-01-01

    An innovative containment vessel, namely Steel plate reinforced Concrete Containment Vessel (SCCV) is developed for Japan Sodium-Cooled Fast Reactor (JSFR). Reducing plant construction cost is one of the most important issues for commercialization of fast reactors. This study investigated construction issues including the building structure and the construction method as well as design issues in terms of the applicability of SCCV to fast reactors. An experimental study including loading and/or heating tests has been carried out to investigate the fundamental structural features, which would be provided to develop methodology to evaluate the feasibility of SCCV under the severe conditions. In this paper, the test plan is described as well as the first test results. (author)

  12. Fuel-coolant interaction visualization test for in-vessel corium retention external reactor vessel cooling (IVR-ERVC) condition

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Hong, Seong Ho; Song, Jin Ho; Hong, Seong Wan [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

  13. Investigation of the loss of forced cooling test by using the high temperature engineering test reactor (HTTR) (Contract research)

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Inaba, Yoshitomo; Goto, Minoru; Tochio, Daisuke

    2007-09-01

    The three gas circulators trip test and the vessel cooling system stop test as the safety demonstration test by using the High Temperature engineering Test Reactor (HTTR) are under planning to demonstrate inherent safety features of High Temperature Gas-cooled Reactor. All three gas circulators to circulate the helium gas as the coolant are stopped to simulate the loss of forced cooling in the three gas circulators trip test. The stop of the vessel cooling system located outside the reactor pressure vessel to remove the residual heat of the reactor core follows the stop of all three gas circulators in the vessel cooling system stop test. The analysis of the reactor transient for such tests and abnormal events postulated during the test was performed. From the result of analysis, it was confirmed that the three gas circulators trip test and the vessel cooling system stop test can be performed within the region of the normal operation in the HTTR and the safety of the reactor facility is ensured even if the abnormal events would occur. (author)

  14. Development of heat transfer enhancement techniques for external cooling of an advanced reactor vessel

    Science.gov (United States)

    Yang, Jun

    Nucleate boiling is a well-recognized means for passively removing high heat loads (up to ˜106 W/m2) generated by a molten reactor core under severe accident conditions while maintaining relatively low reactor vessel temperature (Critical Heat Flux (CHF), becomes the key to the success of external passive cooling of reactor vessel undergoing core disrupture accidents. In the present study, two boiling heat transfer enhancement methods have been proposed, experimentally investigated and theoretically modelled. The first method involves the use of a suitable surface coating to enhance downward-facing boiling rate and CHF limit so as to substantially increase the possibility of reactor vessel surviving high thermal load attack. The second method involves the use of an enhanced vessel/insulation design to facilitate the process of steam venting through the annular channel formed between the reactor vessel and the insulation structure, which in turn would further enhance both the boiling rate and CHF limit. Among the various available surface coating techniques, metallic micro-porous layer surface coating has been identified as an appropriate coating material for use in External Reactor Vessel Cooling (ERVC) based on the overall consideration of enhanced performance, durability, the ease of manufacturing and application. Since no previous research work had explored the feasibility of applying such a metallic micro-porous layer surface coating on a large, downward facing and curved surface such as the bottom head of a reactor vessel, a series of characterization tests and experiments were performed in the present study to determine a suitable coating material composition and application method. Using the optimized metallic micro-porous surface coatings, quenching and steady-state boiling experiments were conducted in the Sub-scale Boundary Layer Boiling (SBLB) test facility at Penn State to investigate the nucleate boiling and CHF enhancement effects of the surface

  15. High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

    International Nuclear Information System (INIS)

    1983-01-01

    The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented

  16. Divertor cooling device

    International Nuclear Information System (INIS)

    Nakayama, Tadakazu; Hayashi, Katsumi; Handa, Hiroyuki

    1993-01-01

    Cooling water for a divertor cooling system cools the divertor, thereafter, passes through pipelines connecting the exit pipelines of the divertor cooling system and the inlet pipelines of a blanket cooling system and is introduced to the blanket cooling system in a vacuum vessel. It undergoes emission of neutrons, and cooling water in the divertor cooling system containing a great amount of N-16 which is generated by radioactivation of O-16 is introduced to the blanket cooling system in the vacuum vessel by way of pipelines, and after cooling, passes through exit pipelines of the blanket cooling system and is introduced to the outside of the vacuum vessel. Radiation of N-16 in the cooling water is decayed sufficiently with passage of time during cooling of the blanket, thereby enabling to decrease the amount of shielding materials such as facilities and pipelines, and ensure spaces. (N.H.)

  17. Investigation of Focusing Effect according to the Cooling Condition and Height of the Metallic layer in a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Je-Young; Chung, Bum-Jin [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    The Fukushima nuclear power plant accident has led to renewed research interests in severe accidents of nuclear power plants. In-Vessel Retention (IVR) of core melt is one of key severe accident management strategies adopted in nuclear power plant design. The metallic layer is heated from below by the radioactive decay heat generated at the oxide pool, and is cooled from above and side walls. During the IVR process, reactor vessel may be cooled externally (ERVC) and the heat fluxes to the side wall increase with larger temperature difference than above. This {sup F}ocusing effect{sup i}s varied by cooling condition of upper boundary and height of the metallic layer. A sulfuric acid–copper sulfate (H{sub 2}SO{sub 4} - CuSO{sub 4}) electroplating system was adopted as the mass transfer system. Numerical analysis using the commercial CFD program FLUENT 6.3 were carried out with the same material properties and cooling conditions to examine the variation of the cell. The experimental and numerical studies were performed to investigate the focusing effect according to cooling condition of upper boundary and the height in metallic layer. The height of the side wall was varied for three different cooling conditions: top only, side only, and both top and side. Mass transfer experiments, based on the analogy concept, were carried out in order to achieve high Rayleigh number. The experimental results agreed well with the Rayleigh-Benard convection correlations of Dropkin and Somerscales and Globe and Dropkin. The heat transfer on side wall cooling condition without top cooling is highest and was enhanced by decreasing the aspect ratio. The numerical results agreed well with the experimental results. Each cell pattern (cell size, cell direction, central location of cell) differed in the cooling condition. Therefore, it is difficult to predict the internal flow due to complexity of cell formation behavior.

  18. Probabilistic fracture mechanics analysis of boiling water reactor vessel for cool-down and low temperature over-pressurization transients

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jeong Soon; Choi, Young Hwan; Jhung, Myung Jo [Safety Research Division, Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-04-15

    The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

  19. Radioactive liquid containing vessel

    International Nuclear Information System (INIS)

    Sakurada, Tetsuo; Kawamura, Hironobu.

    1993-01-01

    Cooling jackets are coiled around the outer circumference of a container vessel, and the outer circumference thereof is covered with a surrounding plate. A liquid of good conductivity (for example, water) is filled between the cooling jackets and the surrounding plate. A radioactive liquid is supplied to the container vessel passing through a supply pipe and discharged passing through a discharge pipe. Cooling water at high pressure is passed through the cooling water jackets in order to remove the heat generated from the radioactive liquid. Since cooling water at high pressure is thus passed through the coiled pipes, the wall thickness of the container vessel and the cooling water jackets can be reduced, thereby enabling to reduce the cost. Further, even if the radioactive liquid is leaked, there is no worry of contaminating cooling water, to prevent contamination. (I.N.)

  20. Research towards ultrasonic systems to assist in-vessel manipulations in liquid metal cooled reactors

    International Nuclear Information System (INIS)

    Dierckx, Marc; Van-Dyck, Dries

    2013-06-01

    We describe the state of the art of the research towards ultrasonic measurement methods for use in lead-bismuth cooled liquid metal reactors. Our current research activities are highly focused on specific tasks in the MYRRHA system, which is a fast spectrum research reactor cooled with the eutectic mixture of lead and bismuth (LBE) and is conceived as an accelerator driven system capable of operating in both sub-critical and critical mode. As liquid metal is opaque to light, normal visual feedback during fuel manipulations in the reactor vessel is not available and must therefore be replaced by a system that is not hindered by the opacity of the coolant. In this respect ultrasonic measurement techniques have been proposed and even developed in the past for operation in sodium cooled reactors. To our knowledge, no such systems have ever been deployed in lead based reactors and we are the first to have a research program in this direction as will be detailed in this paper. We give an overview of the acoustic properties of LBE and compare them with the properties of sodium and water to theoretically show the feasibility of ultrasonic systems operating in LBE. In the second part of the paper we discuss the results of the validation experiments in water and LBE. A typical scene is ultrasonically probed by a mechanical scanning system while the signals are processed to render a 3D visualization on a computer screen. It will become clear that mechanical scanning is capable of producing acceptable images but that it is a time consuming process that is not fit to solve the initial task to providing feedback during manipulations in the reactor vessel. That is why we propose to use several dedicated ultrasonic systems each adapted to a specific task and capable to provide real-time feedback of the ongoing manipulations, as is detailed in the third and final part of the paper. (authors)

  1. GAREC analyses in Support of In-Vessel Retention Concept

    International Nuclear Information System (INIS)

    Azarian, G.; Gandrille, P.; Dumontet, A.; Grange; Barbier, F; Bellon, M.; Bordier, G.; Boulanger, F.; Cognet, G.; Gatt, J.M.; Humbert, J.M.; Laporte, T.; Lepareux, M.; Richard, P.; Robert, G.; Seiler, J.M.; Szabo, I.; Tourasse, M.; Valin, F.; Van Dorsselaere, J.P.

    1999-01-01

    The authors describe the analyses of the in-vessel retention capability which the GAREC group has performed for present and future French PWR designs. They present the reactor characteristics which are considered, describe the physical situations which are analysed and the relocation processes initiated by a corium flow, discuss the jet impacts, the debris formation and behaviour in the vessel lower head in a dry situation with absence of cooling, in wet situations in absence of external cooling, in wet situation with external cooling, in dry situation with external cooling. In this last case, they discuss the power dissipated in the corium, the molten salt behaviour, the heat flux distribution from the pool, the residual wall thickness, the heat flux distribution from the metal layer, the thermal-hydraulic aspects of water injection in the pool, the effects of crust instabilities, the external cooling, and the vessel mechanical behaviour. Then, they address the vapour explosion which may occur: mechanical loads leading to vessel failure in the cases of an eroded or non-eroded vessel, corium masses participating to the interaction (corium jets to the lower head, reflooding of corium pools with water). They finally briefly discuss the possible design improvements for in-vessel retention

  2. Cooling water distribution system

    Science.gov (United States)

    Orr, Richard

    1994-01-01

    A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

  3. Study on mitigation of in-vessel release of fission products in severe accidents of PWR

    International Nuclear Information System (INIS)

    Huang, G.F.; Tong, L.L.; Li, J.X.; Cao, X.W.

    2010-01-01

    Research highlights: → In-vessel release of fission products in severe accidents for 600 MW PWR is analyzed. → Mitigation effect of primary feed-and-bleed on in-vessel release is investigated. → Mitigation effect of secondary feed-and-bleed on in-vessel release is studied. → Mitigation effect of ex-vessel cooling on in-vessel release is evaluated. - Abstract: During the severe accidents in a nuclear power plant, large amounts of fission products release with accident progression, including in-vessel and ex-vessel release. Mitigation of fission products release is demanded for alleviating radiological consequence in severe accidents. Mitigation countermeasures to in-vessel release are studied for Chinese 600 MW pressurized water reactor (PWR), including feed-and-bleed in primary circuit, feed-and-bleed in secondary circuit and ex-vessel cooling. SBO, LOFW, SBLOCA and LBLOCA are selected as typical severe accident sequences. Based on the evaluation of in-vessel release with different startup time of countermeasure, and the coupling relationship between thermohydraulics and in-vessel release of fission products, some results are achieved. Feed-and-bleed in primary circuit is an effective countermeasure to mitigate in-vessel release of fission products, and earlier startup time of countermeasure is more feasible. Feed-and-bleed in secondary circuit is also an effective countermeasure to mitigate in-vessel release for most severe accident sequences that can cease core melt progression, e.g. SBO, LOFW and SBLOCA. Ex-vessel cooling has no mitigation effect on in-vessel release owing to inevitable core melt and relocation.

  4. Head spray nozzle in reactor pressure vessel

    International Nuclear Information System (INIS)

    Hatano, Shun-ichi.

    1990-01-01

    In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

  5. Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2013-12-15

    Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m{sup 2} s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface.

  6. Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

    International Nuclear Information System (INIS)

    Park, Seong Dae; Bang, In Cheol

    2013-01-01

    Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m 2 s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface

  7. Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

    2002-01-01

    In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

  8. Emergency reactor cooling circuit

    International Nuclear Information System (INIS)

    Araki, Hidefumi; Matsumoto, Tomoyuki; Kataoka, Yoshiyuki.

    1994-01-01

    Cooling water in a gravitationally dropping water reservoir is injected into a reactor pressure vessel passing through a pipeline upon occurrence of emergency. The pipeline is inclined downwardly having one end thereof being in communication with the pressure vessel. During normal operation, the cooling water in the upper portion of the inclined pipeline is heated by convection heat transfer from the communication portion with the pressure vessel. On the other hand, cooling water present at a position lower than the communication portion forms cooling water lumps. Accordingly, temperature stratification layers are formed in the inclined pipeline. Therefore, temperature rise of water in a vertical pipeline connected to the inclined pipeline is small. With such a constitution, the amount of heat lost from the pressure vessel by way of the water injection pipeline is reduced. Further, there is no worry that cooling water to be injected upon occurrence of emergency is boiled under reduced pressure in the injection pipeline to delay the depressurization of the pressure vessel. (I.N.)

  9. Proceedings of the Workshop on in-vessel core debris retention and coolability

    International Nuclear Information System (INIS)

    1999-01-01

    This conference on in-vessel core debris retention and coolability is composed of 37 papers grouped in three sessions: session 1 (Keynote papers: Key phenomena of late phase core melt progression, accident management strategies and status quo of severe fuel damage codes, In-vessel retention as a severe accident management scheme, GAREC analyses in support of in-vessel retention concept, Latest findings of RASPLAV project); session 2 - Experiments and model development with five sub-sessions: sub-session 1 (Debris bed heat transfer: Debris and Pool Formation/Heat Transfer in FARO-LWR: Experiments and Analyses, Evaporation and Flow of Coolant at the Bottom of a Particle-Bed modelling Relocated Debris, Investigations on the Coolability of Debris in the Lower Head with WABE-2D and MESOCO-2D, Uncertainty and Sensitivity Analysis of the Heat Transfer Mechanisms in the Lower Head, Simulation of the Arrival and Evolution of Debris in a PWR Lower Head with the SFD ICARE2 code), sub-session 2 (Corium properties, molten pool natural convection, and crust formation: Physico-chemistry and corium properties for in-vessel retention, Experimental data on heat flux distribution from volumetrically heated pool with frozen boundaries, Thermal hydraulic phenomena in corium pools - numerical simulation with TOLBIAC and experimental validation with BALI, TOLBIAC code simulations of some molten salt RASPLAV experiments, SIMECO experiments on in-vessel melt pool formation and heat transfer with and without a metallic layer, Numerical investigation of turbulent natural convection heat transfer in an internally-heated melt pool and metallic layer, Current status and validation of CON2D and 3D code, Free convection of heat-generating fluid in a constrained during experimental simulation of heat transfer in slice geometry), sub-session 3 (Gap formation and gap cooling: Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water, Experimental investigations

  10. Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

  11. Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon, E-mail: rjpark@kaeri.re.kr; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

    2016-03-15

    Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m{sup 2} s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

  12. Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

    2016-01-01

    Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m"2 s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

  13. Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

  14. Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (5) Identification of dominant factors in ex-vessel accident sequences

    International Nuclear Information System (INIS)

    Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya

    2009-01-01

    The evaluation of accident progression outside of a reactor vessel (ex-vessel) and subsequent transfer behavior of radioactive materials is of great importance from the viewpoint of Level 2 PSA. Hence typical ex-vessel accident sequences in the JAEA Sodium-cooled Fast Reactor are qualitatively discussed in this paper and dominant behaviors or factors in the sequences are investigated through parametric calculations using the CONTAIN/LMR code. Scenarios to be focused on are, 1) sodium vapor leakage from the reactor vessel and 2) sodium-concrete reaction, which are both to be considered in the accident category of LOHRS (loss of heat removal system) and might be followed by an early containment failure due to the thermal effect of sodium combustion and hydrogen burning respectively. The calculated results clarify that the sodium vapor leak rate and the scale of sodium-concrete reaction are the important factors to dominate the ex-vessel accident progression. In addition to the understandings of the dominant factors, the analyzed results also provide the specific information such as pressure loading value to the containment and the timing of pressurization, which is indispensable as technical base in Level 2 PSA for developing event trees and for quantifying the accident consequences. (author)

  15. Structural integrity investigation for RPV with various cooling water levels under pressurized melting pool

    Directory of Open Access Journals (Sweden)

    J. Mao

    2018-03-01

    Full Text Available The strategy denoted as in-vessel retention (IVR is widely used in severe accident (SA management by most advanced nuclear power plants. The essence of IVR mitigation is to provide long-term external water cooling in maintaining the reactor pressure vessel (RPV integrity. Actually, the traditional IVR concept assumed that RPV was fully submerged into the water flooding, and the melting pool was depressurized during the SA. The above assumptions weren't seriously challenged until the occurrence of Fukushima accident on 2011, suggesting the structural behavior had not been appropriately assessed. Therefore, the paper tries to address the structure-related issue on determining whether RPV safety can be maintained or not with the effect of various water levels and internal pressures created from core meltdown accident. In achieving it, the RPV structural behaviors are numerically investigated in terms of several field parameters, such as temperature, deformation, stress, plastic strain, creep strain, and total damage. Due to the presence of high temperature melt on the inside and water cooling on the outside, the RPV failure is governed by the failure mechanisms of creep, thermal-plasticity and plasticity. The creep and plastic damages are interacted with each other, which further accelerate the failure process. Through detailed investigation, it is found that the internal pressure as well as water levels plays an important role in determining the RPV failure time, mode and site.

  16. Fluid elastic instability analysis of 1/6th experimental model of PFBR main vessel cooling circuit

    International Nuclear Information System (INIS)

    Jalaldeen, S.; Ravi, R.; Chellapandi, P.; Bhoje, S.B.

    1993-01-01

    In reactor assembly of Prototype Fast Breeder Reactor (PFBR), the main vessel (MV) temperature is kept below creep range i.e. less than 427 deg C by way of diverting a small fraction of core flow from the cold pool and sent through the passage between main vessel and an outer cylindrical baffle to cool the vessel. The sodium coning from this, is collected by another inner baffle and then returned to cold pool again. This system is termed as MV cooling circuit. The outer and inner baffles form feeding and restitution collectors respectively. The sodium from the feeding collector flows over the outer baffle and falls through a height of about 0.5 m before impacting on the free surface of sodium in the restitution collector. The fall of sodium may become a source of vibration of the baffles. Such vibrations have been already noted in case of SPX-I during its commissioning stage. For PFBR, the theoretical analysis was done to assess the fluid-elastic instability risks and stability charts were obtained. By this, it was concluded that the operating point (flow rate and fall height) lies within the stable zone. In order to confirm the above analysis results, a series of experiments were proposed. One preliminary experiment on 1/16 th model of MV cooling circuit has been completed. This model has also been analysed theoretically for the fluid- elastic instability, the theoretical analysis involves 2 stage computations. In the first stage, free vibration analysis with fluid structure interaction (FSI) effect for experimental model has been done using INCA (CASTEM 1985) code and all the mode shapes including sloshing are extracted. In the second stage the instability analysis is performed with the free vibration results from INCA. For the instability computations, a code WEIR has been written based on Aita's instability criteria [Aita.S. 1986

  17. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

  18. Storage vessel for radiation contaminated container

    International Nuclear Information System (INIS)

    Sakatani, Tadatsugu.

    1996-01-01

    In a storage vessel of the present invention, a plurality of radiation contaminated material containing bodies are vertically stacked in a cell chamber. Then, the storage vessel comprises a containing tube for containing a plurality of the containing bodies, cooling coils wound around the containing tube, a cooling medium circulating system connected to the cooling coils and circulating cooling medium, and a heat exchanger interposed to the cooling medium circulating system for removing heat of the cooling medium. Heat of the radioactive material containing bodies is transferred to cooling air and cooling coils by way of the container tube, thereby cooling the containing bodies. By the operation of circulating pumps in a cooling medium circulation system, the cooling medium circulates through a circulation channel comprising a cooling medium transfer pipes, cooling medium branching tubes, cooling coils and the heat exchanger, then heat of the cooling medium is transferred to a heat utilizing system by way of the heat exchanger to attain effective utilization of the heat. In this case, heat can be taken out stably even when the storage amount fluctuates and heat releasing amount is reduced, and improvement of heat transfer promotes the cooling of the containing bodies, which enables minimization of the size of the storage vessel. (T.M.)

  19. Numerical investigation of the reactor pressure vessel behaviour under severe accident conditions taking into account the combined processes of the vessel creep and the molten pool natural convection

    International Nuclear Information System (INIS)

    Loktionov, V.D.; Mukhtarov, E.S.; Yaroshenko, N.I.; Orlov, V.E.

    1999-01-01

    Analysis of the WWER lower head behaviour and its failure has been performed for several molten pool structures and internal overpressure levels in a reactor pressure vessel (RPV). The different types of the molten pools (homogeneous, conventionally homogeneous, conventionally stratified, stratified) cover the bounding scenarios during a hypothetical severe accident. The parametric investigations of the failure mode and RPV behaviour for various molten pool types, its heights and internal overpressure levels are presented herein. A coupled treatment in this investigation includes: (i) a 2-D thermohydraulic analysis of a molten pool natural convection. Domestic NARAUFEM code has been used in this detailed analysis for prediction of the heat flux from the molten pool to the RPV inner surface; and (ii) a detailed 3-D transient thermal analysis of the RPV lower head. Domestic 3-D ASHTER-VVR finite element code has been used for the numerical simulations of the high temperature creep and failure of the lower head. The effect of an external RPV cooling, temperature-dependent physical properties of the molten pool and vessel steel, the hydrostatic forces and vessel dead-weight were taken into account in this study. The obtained results show that lower head failure occurs as a result of the vessel creep process which is significantly dependent on both an internal overpressure level and the type of molten pool structure. In particular, it was found that there were combinations of 'overpressure-molten pool structure' when the vessel failure started at the 'hot' layers of the vessel. (orig.)

  20. TMI-2 Vessel Investigation Project integration report

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, J. R.; Rempe, J. L.; Stickler, L. A.; Korth, G. E.; Diercks, D. R.; Neimark, L. A.; Akers, D W; Schuetz, B. K.; Shearer, T L; Chavez, S. A.; Thinnes, G. L.; Witt, R. J.; Corradini, M L; Kos, J. A. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-03-01

    The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel`s condition after the accident.

  1. Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Carroll, D.G.; Chen, C.; Crane, C.; Dalton, R.; Taylor, J.R.; Tosunoglu, S.; Weymouth, T.

    1993-01-01

    One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS

  2. Analytical investigation of multicavity prestressed concrete pressure vessels for elastic loading conditions

    International Nuclear Information System (INIS)

    Fanning, D.N.

    1978-09-01

    A three-dimensional finite-element analysis of a commercial high-temperature gas-cooled reactor (HTGR) was made using the finite-element code STATIC-SAP. Four loading conditions were analyzed elastically to evaluate the behavior of the concentric core prestressed concrete reactor vessel (PCRV) of the HTGR. The results of the analysis were evaluated in accordance with Section III, Division 2, of the ASME Code for Reactor Vessels and Containments. The calculated maximum stresses were found to be well within the Code-allowable values. The analysis was preceded by an evaluation of candidate computer codes using comparisons of experimental data with analytical results for the Ohbayashi-Gumi multicavity PCRV model. This vessel was chosen as a basis for comparison because of its geometrical similarity to the large multicavity PCRV and the anticipated availability of a complete set of the original experimental data. The three-dimensional finite-element codes NONSAP and STATIC-SAP were used for the analysis of the Ohbayashi-Gumi vessel

  3. Experimental investigations on the contribution of the splash-zones in counter-flow cooling towers for water cooling

    International Nuclear Information System (INIS)

    Vladea, I.; Barbu, V.

    1976-01-01

    The relatively high cost of cooling tower packs has led to investigate the contribution of the splash-zones in counter-flow cooling towers, and thereby to determine whether the pack could not be reduced so far, as to be - under certain circumstance - completely eliminated. In this case, one would come to a pure splash cooling tower which would contain inside the equipment required for drop formation only. This problem was investigated experimentally, and it was found that the pack of such a cooling tower could not be eliminated without a reduction in tower effectiveness. (orig.) [de

  4. Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Araj, K.

    1983-01-01

    The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs

  5. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    International Nuclear Information System (INIS)

    Hunsbedt, A.; Boardman, C.E.

    1993-01-01

    A dual passive cooling system for liquid metal cooled nuclear fission reactors is described, comprising the combination of: a reactor vessel for containing a pool of liquid metal coolant with a core of heat generating fissionable fuel substantially submerged therein, a side wall of the reactor vessel forming an innermost first partition; a containment vessel substantially surrounding the reactor vessel in spaced apart relation having a side wall forming a second partition; a first baffle cylinder substantially encircling the containment vessel in spaced apart relation having an encircling wall forming a third partition; a guard vessel substantially surrounding the containment vessel and first baffle cylinder in spaced apart relation having a side wall forming a forth partition; a sliding seal at the top of the guard vessel edge to isolate the dual cooling system air streams; a second baffle cylinder substantially encircling the guard vessel in spaced part relationship having an encircling wan forming a fifth partition; a concrete silo substantially surrounding the guard vessel and the second baffle cylinder in spaced apart relation providing a sixth partition; a first fluid coolant circulating flow course open to the ambient atmosphere for circulating air coolant comprising at lent one down comer duct having an opening to the atmosphere in an upper area thereof and making fluid communication with the space between the guard vessel and the first baffle cylinder and at least one riser duct having an opening to the atmosphere in the upper area thereof and making fluid communication with the space between the first baffle cylinder and the containment vessel whereby cooling fluid air can flow from the atmosphere down through the down comer duct and space between the forth and third partitions and up through the space between the third and second partition and the riser duct then out into the atmosphere; and a second fluid coolant circulating flow

  6. The TPX vacuum vessel and in-vessel components

    International Nuclear Information System (INIS)

    Heitzenroeder, P.; Bialek, J.; Ellis, R.; Kessel, C.; Liew, S.

    1994-01-01

    The Tokamak Physics Experiment (TPX) is a superconducting tokamak with double-null diverters. TPX is designed for 1,000-second discharges with the capability of being upgraded to steady state operation. High neutron yields resulting from the long duration discharges require that special consideration be given to materials and maintainability. A unique feature of the TPX is the use of a low activation, titanium alloy vacuum vessel. Double-wall vessel construction is used since it offers an efficient solution for shielding, bakeout and cooling. Contained within the vacuum vessel are the passive coil system, Plasma Facing Components (PFCs), magnetic diagnostics, and the internal control coils. All PFCs utilize carbon-carbon composites for exposed surfaces

  7. Surveillance tests for light-water cooled nuclear power reactor vessels in IMEF

    International Nuclear Information System (INIS)

    Choo, Yong-Sun; Ahn, Sang-Bok; Park, Dae-Gyu; Jung, Yang-Hong; Yoo, Byung-Ok; Oh, Wan-Ho; Baik, Seung-Je; Koo, Dae-Seo; Lee, Key-Soon

    1999-01-01

    The surveillance tests for light-water cooled nuclear power reactor vessels were established to monitor the radiation-induced changes in the mechanical properties of ferritic materials in the beltline according to US NRC 10 CFR 50 App. G, US NRC RG1.99-rev.2, ASTM E185-82 and E185-94 in Irradiated Materials Examination Facility(IMEF). The surveillance capsule was transported from NPPs pool sites to KAERI IMEF by using a shipping cask. The capsule was cut and dismantled by capsule cutting machine and milling machine in M2 hot cell. Charpy tests and tension tests were performed in M5a and M5b hot cells respectively. Especially the EPMA located at hot lab was used to analyze the Ni and Cu wt% composition of base metal and weld for predicting the adjusted reference temperature(ART). The established process and test results were summarized in this paper. (author)

  8. TPX heating and cooling system

    International Nuclear Information System (INIS)

    Kungl, D.J.; Knutson, D.S.; Costello, J.; Stoenescu, S.; Yemin, L.

    1995-01-01

    TPX, while having primarily super-conducting coils that do not require water cooling, still has very significant water cooling requirements for the plasma heating systems, vacuum vessel, plasma facing components, diagnostics, and ancillary equipment. This is accentuated by the 1000-second pulse requirement. Two major design changes, which have significantly affected the TPX Heating and Cooling System, have been made since the conceptual design review in March of 1993. This paper will discuss these changes and review the current status of the conceptual design. The first change involves replacing the vacuum vessel neutron shielding configuration of lead/glass composite tile by a much simpler and more reliable borated water shield. The second change reduces the operating temperature of the vacuum vessel from 150 C to ≥50 C. With this temperature reduction, all in-vessel components and the vessel will be supplied by coolant at a common ≥50 C inlet temperature. In all, six different heating and cooling supply requirements (temperature, pressure, water quality) for the various TPX components must be met. This paper will detail these requirements and provide an overview of the Heating and Cooling System design while focusing on the ramifications of the TPX changes described above

  9. Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 1. Comprehending the vacuum vessel structure

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Nakahira, Masataka

    2006-01-01

    The functions, conditions and structure of vacuum vessel using tokamak fusion machines are explained. The structural standard and code of vacuum vessel, process of vacuum vessel design, and design of ITER vacuum vessel are described. Production and maintenance of ultra high vacuum, confinement of radioactive materials, support of machines in vessel and electromagnetic force, radiation shield, plasma vertical stability, one-turn electric resistance, high temperature baking heat and remove of nuclear heat, reduce of troidal ripple, structural standard, features of safety of nuclear fusion machines, subjects of structural standard of fusion vacuum vessel, design flow of vacuum vessel, establishment of radial build, selections of materials, baking and cooling method, basic structure, structure of special parts, shield structure, and of support structure, and example of design of structure, ITER, are stated. (S.Y.)

  10. Phenomenological vessel burst investigations

    International Nuclear Information System (INIS)

    Hippelein, K.W.; Julisch, P.; Muz, J.; Schiedermaier, J.

    1985-07-01

    Fourteen burst experiments have been carried out using vessels with circumferential and longitudinal flaws, for investigation of the fracture behaviour, i.e. the time-related fracture opening. The vessels had dimensions (outer diameter x wall thickness = 800 x 47 mm) which correspond to the dimensions of the main coolant piping of a 1300 MW e PWR. The test specimens had been made of the base-safe material 20 MnMoNi 55 and of a special, 22 NiMoCr 37 base alloy. The experimental conditions with regard to pressure and temperature have been chosen so as to correspond to normal operating conditions of a PWR (p∝17.5 MPa, T∝300 0 C), i.e. the flaws have been so dimensioned that failure was to be expected at a pressure of p∝17.5 MPa. As a rule, water has been used as the pressure medium, or in some cases air, in order to influence the time-dependent pressure decrease. Fluid and structural dynamics calculations have also been made. In order to determine the impact of a fast propagating crack on the leak-to-fracture curve, which normally is defined by quasistationary experiments, suitable tests have been made with large-volume, cylindrical vessels (outer diameter x wall thickness x length = 3000 x 21 x 14000 mm) made of the material WSt E 43. The leak-before-fracture criterion has been confirmed. (orig./HP) [de

  11. TMI-2 Vessel Investigation Project integration report

    International Nuclear Information System (INIS)

    Wolf, J.R.; Rempe, J.L.; Stickler, L.A.; Korth, G.E.; Diercks, D.R.; Neimark, L.A.; Akers, D.W.; Schuetz, B.K.; Shearer, T.L.; Chavez, S.A.; Thinnes, G.L.; Witt, R.J.; Corradini, M.L.; Kos, J.A.

    1994-03-01

    The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel's condition after the accident

  12. Analysis code for pressure in reactor containment vessel of ATR. CONPOL

    International Nuclear Information System (INIS)

    1997-08-01

    For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

  13. Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

    International Nuclear Information System (INIS)

    Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

    1979-01-01

    Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

  14. A study of the external cooling capability for the prevention of reactor vessel failure

    Energy Technology Data Exchange (ETDEWEB)

    Chang, S H; Baek, W P; Moon, S K; Yang, S H; Kim, S H [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1995-07-15

    This study (a 3-year program) aims to perform a comprehensive assessment of the feasibility of external vessel flooding with respect to advanced pressurized water reactor plants to be built in Korea. During the second year, appropriate correlations have been chosen to describe the phenomena resulted from the external flooding on the basis of review works. Also performed is to develop the computer program using the chosen correlations and to accomplish the thermal analysis for assessment of the cooling capability of external flooding. Accomplished works for second year are as follows. Review of analytical and experimental works related to the external flooding are performed, appropriate correlations are chosen to describe the phenomena resulted from the external flooding on the basis of first and second year review works. A computer program is also developed to predict the temperature distribution of reactor vessel lower head. Thermal analyses are performed to judge the feasibility of external flooding using developed computer program.

  15. TMI-2 Vessel Investigation Project Metallurgical Program

    International Nuclear Information System (INIS)

    Diercks, D.R.; Neimark, L.A.

    1990-01-01

    The TMI-2 [Three Mile Island unit 2] Vessel Investigation Project Metallurgical Program at Argonne National Laboratory is a part of the international TMI-2 Vessel Investigation Project being conducted jointly by the U.S. Nuclear Regulatory Commission and the Organization for Economic Co-operation and Development (OECD). The overall project consists of three phases, namely (1) recovery of material samples from the lower head of the TMI-2 reactor, (2) examination and analysis of the lower head samples and the preparation and testing of archive material subjected to a similar thermal history, and (3) procurement, examination, and analysis of companion core material located adjacent to or near the lower head material. The specific objectives of the ANL Metallurgical Program, which accounts for a major portion of Phase 2, are to prepare metallographic and mechanical test specimen blanks from the TMI-2 lower head material, prepare similar test specimen blanks from suitable archive material subjected to the appropriate thermal processing, determine the mechanical properties of the lower vessel head and archive materials under the conditions of the core-melt accident, and assess the lower head integrity and margin-to-failure during the accident. The ANL work consists of three tasks: (1) archive materials program, (2) fabrication of metallurgical and mechanical test specimens from the TMI-2 pressure vessel samples, and (3) mechanical property characterization of TMI-2 lower pressure vessel head and archive material

  16. Nuclear reactor vessel fuel thermal insulating barrier

    Science.gov (United States)

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  17. Passive containment cooling water distribution device

    Science.gov (United States)

    Conway, Lawrence E.; Fanto, Susan V.

    1994-01-01

    A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using a series of radial guide elements and cascading weir boxes to collect and then distribute the cooling water into a series of distribution areas through a plurality of cascading weirs. The cooling water is then uniformly distributed over the curved surface by a plurality of weir notches in the face plate of the weir box.

  18. Study on the welding continuous cooling transformation and weldability of SA508Gr4 steel for nuclear pressure vessels

    International Nuclear Information System (INIS)

    Bai, Qingwei; Ma, Yonglin; Xing, Shuqing; Chen, Zhongyi

    2017-01-01

    SA508Gr4 is a newly developed high-strength steel for nuclear reactor pressure vessels. Its welding characteristics remain largely unexplored. In this work, the simulated heat affected zone continuous cooling transformation (SH-CCT) diagram of SA508Gr4 steel was constructed and the high-temperature cooling phase compositions and the properties of the heat affected zone (HAZ) were characterized using dilatometry and microscopic tests. The results show that the phase transformation in the HAZ was divided into bainite and martensite transformation stages. When 4.6 ≤ t_8_/_5 (the HAZ cooling time from 800 C to 500 C) ≤ 15 s, lath-shaped martensite was fully developed, resulting in extensive hardening and cold cracking in the HAZ, while the cooling time required to form the bainite completely exceeds 1 200 s. Thus, to improve weld quality, preheating to 196 C or higher is recommended.

  19. Study on the welding continuous cooling transformation and weldability of SA508Gr4 steel for nuclear pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bai, Qingwei; Ma, Yonglin; Xing, Shuqing; Chen, Zhongyi [Inner Mongolia Univ. of Science and Technology, Baotou (China). School of Material and Metallurgy; Kang, Xiaolan [Baotou Vocational and Technical College (China)

    2017-02-15

    SA508Gr4 is a newly developed high-strength steel for nuclear reactor pressure vessels. Its welding characteristics remain largely unexplored. In this work, the simulated heat affected zone continuous cooling transformation (SH-CCT) diagram of SA508Gr4 steel was constructed and the high-temperature cooling phase compositions and the properties of the heat affected zone (HAZ) were characterized using dilatometry and microscopic tests. The results show that the phase transformation in the HAZ was divided into bainite and martensite transformation stages. When 4.6 ≤ t{sub 8/5} (the HAZ cooling time from 800 C to 500 C) ≤ 15 s, lath-shaped martensite was fully developed, resulting in extensive hardening and cold cracking in the HAZ, while the cooling time required to form the bainite completely exceeds 1 200 s. Thus, to improve weld quality, preheating to 196 C or higher is recommended.

  20. Inelastic Cyclic Deformation Behaviors of Type 316H Stainless Steel for Reactor Pressure Vessel of Sodium-Cooled Fast Reactor at Elevated Temperatures

    International Nuclear Information System (INIS)

    Yoon, Ji-Hyun; Hong, Seokmin; Koo, Gyeong-Hoi; Lee, Bong-Sang; Kim, Young-Chun

    2015-01-01

    Type 316H stainless steel is a primary candidate material for a reactor pressure vessel of a sodium-cooled fast (SFR) reactor which is under development in Korea. The reactor pressure vessel for a SFR is subjected to inelastic deformation induced by cyclic thermal stress. Fully reversed cyclic testing and ratcheting testing at elevated temperatures were performed to characterize the inelastic cyclic deformation behaviors of Type 316H stainless steel at the SFR operating temperature. It was found that cyclic hardening of Type 316H stainless steel was enhanced, and the accumulation of ratcheting deformation of Type 316H stainless steel was retarded at around the SFR operating temperature. The results of the tensile testing and the microstructural investigation for dislocated structures after the inelastic deformation testing showed that dynamic strain aging affected the inelastic cyclic deformation behavior of Type 316 stainless steel at around the SFR operating temperature.

  1. Investigations of combined used of cooling ponds with cooling towers or spraying systems

    International Nuclear Information System (INIS)

    Farforovsky, V.B.

    1990-01-01

    Based on a brief analysis of the methods of investigating cooling ponds, spraying systems and cooling towers, a conclusion is made that the direct modelling of the combined use of cooling systems listed cannot be realized. An approach to scale modelling of cooling ponds is proposed enabling all problems posed by the combined use of coolers to be solved. Emphasized is the importance of a proper choice of a scheme of including a cooler in a general water circulation system of thermal and nuclear power plants. A sequence of selecting a cooling tower of the type and spraying system of the size ensuring the specified temperature regime in a water circulation system is exemplified by the water system of the Ghorasal thermal power plant in Bangladesh

  2. Thermal-hydraulic analyses of pressurized-thermal-shock-induced vessel ruptures

    International Nuclear Information System (INIS)

    Dobranich, D.

    1982-05-01

    A severe overcooling transient was postulated to produce vessel wall temperatures below the nil-ductility transition temperature which in conjunction with system repressurization, led to vessel rupture at the core midplane. Such transients are referred to as pressurized-thermal-shock transients. A wide range of vessel rupture sizes were investigated to assess the emergency system's ability to cool the fuel rods. Ruptures greater than approximately 0.015 m 2 produced flows greater than those of the emergency system and resulted in core uncovery and subsequent core damage

  3. Emergency cooling device for reactors

    International Nuclear Information System (INIS)

    Inoue, Hisamichi; Naito, Masanori; Sato, Chikara; Chino, Koichi.

    1975-01-01

    Object: To pour high pressure cooling water into a core, when coolant is lost in a boiling water reactor, thereby restraining the rise of fuel cladding. Structure: A control rod guiding pipe, which is moved up and down by a control rod, is mounted on the bottom of a pressure vessel, the control rod guiding pipe being communicated with a high pressure cooling water tank positioned externally of the pressure vessel, and a differential in pressure between the pressure vessel and the aforesaid tank is detected when trouble of coolant loss occurs, and the high pressure cooling water within the tank is poured into the core through the control rod guiding pipe to restrain the rise of fuel cladding. (Kamimura, M.)

  4. The 1500 MW fast breeder reactor the double envelope-vessel anchored in concrete

    International Nuclear Information System (INIS)

    Bolvin, M.

    1981-01-01

    This paper givers an account of EDF investigations to reduce the investment costs of the 1500 MW Fast Reactor (RNR 1500) without prejudice to the safety requirements. It deals with the double envelope-vessel, designed to minimize radiation consequences in the event of accidental leakage in the main vessel. In the Fast Reactors in operation (PHOENIX), under construction (CRYS-MALVILLE), and under project (NR 1500), the double envelope-steel vessel hangs down from the upper part of the reactor block, its weight being approximately 300 t. In the new design, the vessel is fixed into the concrete which supports the main vessel, by means of steel anchors. A thermal insulation isolates it from the main vessel. The installation of coils in the concrete, next to the lining, allows for water circulation to cool the concrete. (orig./GL)

  5. Development of an integrated prestressed-concrete pressure vessel for water-cooled reactors (SBB type 'STERN' (star) with supporting boiler)

    International Nuclear Information System (INIS)

    Jueptner, G.; Kumpf, H.; Molz, G.; Neunert, B.; Seidl, O.

    1976-01-01

    This report goes into the reasons for selecting a 'STERN' (star) vessel configuration for accommodating a complete primary circuit including PWR, this involving the grouping of cylindrical pressure vessels of independent design into a star-shaped configuration with the central vessel housing the reactor core in the middle. This arrangement was made possible by application of the DYWIDAG-radial prestressing process generating controlled annular prestressing using existing presses and by an organic coupling of individual vessels. The liner, heat insulating and cooling system required for each vessel comprises a so-called support boiler, i.e. a hot liner not handicapped by the disadvantages of other systems. The support boiler is placed in the and PCV and has flat floor and cover surfaces. Temperature constraints are reduced to specific design requirements by means of radial gap permitting precise adjustment in conjunction with an axial expanding element comprising a multilayer diaphragm which is supported in operation. A detailed description is given of the PCPV, the support boiler and the cover used in the center vessel as well as of their design, the assembly and construction work is described and a summary presented of the quantities and estimated prices involved. Due to the absence of steam raising facilities adapted to meet the star-shaped configuration requirements, a study of satellite vessels was dispensed with, the design of which is in full accord with that of the center vessel. One part of the report is concerned with the calculation of the center vessel. (orig./HP) [de

  6. The impact of microwave stray radiation to in-vessel diagnostic components

    Energy Technology Data Exchange (ETDEWEB)

    Hirsch, M.; Laqua, H. P.; Hathiramani, D.; Baldzuhn, J.; Biedermann, C.; Cardella, A.; Erckmann, V.; König, R.; Köppen, M.; Zhang, D. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, EURATOM Association, D-17489 Greifswald (Germany); Oosterbeek, J.; Brand, H. von der; Parquay, S. [Technische Universiteit Eindhoven, department Technische Natuurkunde, working group for Plasma Physics and Radiation Technology, Den Doelch 2, 5612 AZ Eindhoven (Netherlands); Jimenez, R. [Centro de Investigationes Energeticas, Medioambientales y Technológicas, Association EURATOM/CIEMAT, Avenida Complutense 22, Madrid 28040 (Spain); Collaboration: W7-X Teasm

    2014-08-21

    Microwave stray radiation resulting from unabsorbed multiple reflected ECRH / ECCD beams may cause severe heating of microwave absorbing in-vessel components such as gaskets, bellows, windows, ceramics and cable insulations. In view of long-pulse operation of WENDELSTEIN-7X the MIcrowave STray RAdiation Launch facility, MISTRAL, allows to test in-vessel components in the environment of isotropic 140 GHz microwave radiation at power load of up to 50 kW/m{sup 2} over 30 min. The results show that both, sufficient microwave shielding measures and cooling of all components are mandatory. If shielding/cooling measures of in-vessel diagnostic components are not efficient enough, the level of stray radiation may be (locally) reduced by dedicated absorbing ceramic coatings on cooled structures.

  7. TMI-2 Vessel Investigation Project (VIP) Metallurgical Program

    International Nuclear Information System (INIS)

    Diercks, D.R.; Neimark, L.A.

    1990-06-01

    The TMI-2 Vessel Investigation Project (VIP) Metallurgical Program is a part of the international TMI-2 Vessel Investigation Project being conducting jointly by the US Nuclear Regulatory Commission and the Organization for Economic Co-operation and Development (OECD). The overall project consists of three phases, namely (1) recovery of material samples from the lower head of the TMI-2 reactor, (2) examination and analysis of the lower head samples and the preparation and testing of archive material subjected to a similar thermal history, and (3) procurement, examination, and analysis of companion core material located adjacent to or near the lower head material. The specific objectives of the ANL Metallurgical Program, which comprises a major portion of Phase 2, are to prepare metallographic and mechanical test specimen blanks from the TMI-2 lower head material, prepare similar test specimen blanks from suitable archive material subjected to the appropriate thermal processing, determine the mechanical properties of the lower vessel head and archive materials under the conditions of the core-melt accident, and assess the lower head integrity and margin-to-failure during the accident. The ANL work consists of three tasks: (1) archive materials program, (2) fabrication of metallurgical and mechanical test specimens from the TMI-2 pressure vessel samples, and (3) mechanical property characterization of TMI-2 lower pressure vessel head and archive material

  8. RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

    International Nuclear Information System (INIS)

    Gencheva, Rositsa V.; Stefanova, Antoaneta E.; Groudev, Pavlin P.

    2005-01-01

    During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is 'Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed and Bleed procedure initiation. For the purpose of this, operator action with 'Reactor vessel off-gas valve - 0.032 m' opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety

  9. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  10. Ex-vessel corium coolability sensitivity study with the CORQUENCH code

    International Nuclear Information System (INIS)

    Robb, Kevin; Corradini, Michael

    2009-01-01

    An unresolved safety issue for light water reactor beyond design basis accidents is the coolability and stabilization of ex-vessel core melt debris by top flooding. Several experimental programs, including the OECD MACE, MCCI-1, and the current MCCI-2 program, have investigated core-concrete interactions and debris cooling of ex-vessel core melts. As part of the OECD programs, the CORQUENCH computer model was developed based on phenomena identified from the experiments. Predictions by CORQUENCH have previously been compared against experiments and have also been extrapolated to reactor scale. The current study applied statistical techniques to investigate the importance of initial system parameters and cooling phenomena in CORQUENCH 3.01 on the accident progression of ex-vessel core melts. The purpose of this sensitivity study is to identify parameters that are of major importance, any code peculiarities over the range of inputs, and where modeling improvements may produce the most gain in prediction accuracy. The sensitivity studies were carried out over a range of input conditions, in 1-D and 2-D geometries, and for two concrete compositions. In terms of initial system parameters, the melt height had the most importance on concrete ablation and melt coolability. With respect to cooling phenomena, the amount of melt entrainment through the crust had the most importance on concrete ablation and melt coolability. (author)

  11. External cooling: The SWR 1000 severe accident management strategy. Part 1: motivation, strategy, analysis: melt phase, vessel integrity during melt-water interaction

    International Nuclear Information System (INIS)

    Kolev, Nikolay Ivanov

    2004-01-01

    This paper provides the description of the basics behind design features for the severe accident management strategy of the SWR 1000. The hydrogen detonation/deflagration problem is avoided by containment inertization. In-vessel retention of molten core debris via water cooling of the external surface of the reactor vessel is the severe accident management concept of the SWR 1000 passive plant. During postulated bounding severe accidents, the accident management strategy is to flood the reactor cavity with Core Flooding Pool water and to submerge the reactor vessel, thus preventing vessel failure in the SWR 1000. Considerable safety margins have determined by using state of the art experiment and analysis: regarding (a) strength of the vessel during the melt relocation and its interaction with water; (b) the heat flux at the external vessel wall; (c) the structural resistance of the hot structures during the long term period. Ex-vessel events are prevented by preserving the integrity of the vessel and its penetrations and by assuring positive external pressure at the predominant part of the external vessel in the region of the molten corium pool. Part 1 describes the motivation for selecting this strategy, the general description of the strategy and the part of the analysis associated with the vessel integrity during the melt-water interaction. (author)

  12. Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

    International Nuclear Information System (INIS)

    Bodmann, E.

    1984-01-01

    Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

  13. A resting bottom sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Costes, D.

    2012-01-01

    This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

  14. Preliminary calculation with code CONTEMPT-LT for spray cooling tests with JAERI model containment vessel

    International Nuclear Information System (INIS)

    Tanaka, Mitsugu

    1978-01-01

    LWR plants have a containment spray system to reduce the escape of radioactive material to the environment in a loss-of-coolant accident (LOCA) by washing out fission products, especially radioiodine, and condensing the steam to lower the pressure. For carrying out the containment spray tests, pressure and temperature behaviour of the JAERI Model Containment Vessel in spray cooling has been calculated with computer program CONTEMPT-LT. The following could be studied quantitatively: (1) pressure and temperature raise rates for steam addition rate and (2) pressure fall rate for spray flow rate and spray heat transfer efficiency. (auth.)

  15. Natural Convection Heat Transfer of Oxide Pool During In-Vessel Retention of Core Melts

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hae-Kyun; Chung, Bum-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The integrity of reactor vessel may be threatened by the heat generation at the oxide pool and to the natural convection heat transfer to the reactor vessel by those two layers. Therefore, External Reactor Vessel Cooling (ERVC) is performed in order to secure the integrity of the reactor vessel. Whether the IVR(In-Vessel Retention) Strategy can be applicable to a larger reactor is the technical concern, which nourished the research interest for the natural convection heat transfer of metal and oxide pool and ERVC performance. Especially, it is hard to simulate oxide pool by experimentally due to the high level of buoyancy. Moreover, the volumetrically exothermic working fluid should be adopted to simulate the behavior of the core melts. Therefore, the volumetric heat sources that immersed in the working fluid have been adopted to simulate oxide pool by experiment. We investigated oxide pool with two different designs of the volumetric heat sources that adopted previous experiments. The investigation was performed by mass transfer experiment using analogy between heat and mass transfers. The results were compared to previous studies. We simulated the natural convection heat transfer of the oxide pool by mass transfer experiment. The isothermally cooled condition was established by limiting current technique firstly. The results were compared to previous studies under identical design of the volumetric heat sources. The average Nu's of the curvature and the top plate were close to the previous studies.

  16. Emergency core cooling system

    International Nuclear Information System (INIS)

    Abe, Nobuaki.

    1993-01-01

    A reactor comprises a static emergency reactor core cooling system having an automatic depressurization system and a gravitationally dropping type water injection system and a container cooling system by an isolation condenser. A depressurization pipeline of the automatic depressurization system connected to a reactor pressure vessel branches in the midway. The branched depressurizing pipelines are extended into an upper dry well and a lower dry well, in which depressurization valves are disposed at the top end portions of the pipelines respectively. If loss-of-coolant accidents should occur, the depressurization valve of the automatic depressurization system is actuated by lowering of water level in the pressure vessel. This causes nitrogen gases in the upper and the lower dry wells to transfer together with discharged steams effectively to a suppression pool passing through a bent tube. Accordingly, the gravitationally dropping type water injection system can be actuated faster. Further, subsequent cooling for the reactor vessel can be ensured sufficiently by the isolation condenser. (I.N.)

  17. Thermographic venous blood flow characterization with external cooling stimulation

    Science.gov (United States)

    Saxena, Ashish; Ng, E. Y. K.; Raman, Vignesh

    2018-05-01

    Experimental characterization of blood flow in a human forearm is done with the application of continuous external cooling based active thermography method. Qualitative and quantitative detection of the blood vessel in a thermal image is done, along with the evaluation of blood vessel diameter, blood flow direction, and velocity in the target blood vessel. Subtraction based image manipulation is performed to enhance the feature contrast of the thermal image acquired after the removal of external cooling. To demonstrate the effect of occlusion diseases (obstruction), an external cuff based occlusion is applied after the removal of cooling and its effect on the skin rewarming is studied. Using external cooling, a transit time method based blood flow velocity estimation is done. From the results obtained, it is evident that an external cooling based active thermography method can be used to develop a diagnosis tool for superficial blood vessel diseases.

  18. Study on ex-vessel cooling of RPV (behavior of coalesced bubbles and trigger condition of critical heat flux on inclined plate)

    International Nuclear Information System (INIS)

    Ohtake, H.; Koizumi, Y.; Takano, K.I.

    2001-01-01

    The Ex-vessel cooling of Reactor-Pressure-Vessel in Light-Water-Reactor at the severe accident have been proposed for future nuclear reactors. The estimation of Critical-Heat-Flux on a downward-facing curvilinear surface, like a hemisphere, is important to the assessment of the cooling. In this study, the CHFs on inclined surfaces were examined experimentally focusing on orientation of the heating surface. In order to discuss detailed mechanism of the CHF, the behaviors of coalesced bubbles near the heating surface were investigated through visual observations. The critical heat flux obtained in the present experiments increased with the inclined angle over the present experimental range. The dependence of the inclined angle on the critical heat flux was q CHF,R-113 [q] = f (q 0.33 ) for the present experimental results. The effect of the surface orientation on the critical heat flux was roughly explained by using the simple analytical model based on the macro-layer model and Kelvin-Helmholtz instability. From visual observations for behavior of bubbles near the heating surface, whereas the coalesced bubble covered over the heating surface for the inclined angle of 0 degree, the coalesced bubble moved upward to avoid packing the bubble on the surface above 5 degree. As the inclined angle increased, the velocity of the coalesced bubble was high, the period covered the heater and the bubble length were small. The results suggested that the CHF was closely related to forming the coalesced bubble and the behavior of the bubble. (author)

  19. Commissioning result of the KSTAR in-vessel cryo-pump

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. B.; Lee, H. J.; Park, Y.M. [National Fusion Research Institute, Daejeon (Korea, Republic of); and others

    2013-12-15

    KSTAR in-vessel cryo-pump has been installed in the vacuum vessel top and bottom side with up-down symmetry for the better plasma density control in the D-shape H-mode. The cryogenic helium lines of the in-vessel cryo-pump are located at the vertical positions from the vacuum vessel torus center 2,000 mm. The inductive electrical potential has been optimized to reduce risk of electrical breakdown during plasma disruption. In-vessel cryo-pump consists of three parts of coaxial circular shape components; cryo-panel, thermal shield and particle shield. The cryo-panel is cooled down to below 4.5 K. The cryo-panel and thermal shields were made by Inconel 625 tube for higher mechanical strength. The thermal shields and their cooling tubes were annealed in air environment to improve the thermal radiation emissivity on the surface. Surface of cryo-panel was electro-polished to minimize the thermal radiation heat load. The in-vessel cryo-pump was pre-assembled on a test bed in 180 degree segment base. The leak test was carried out after the thermal shock between room temperature to LN2 one before installing them into vacuum vessel. Two segments were welded together in the vacuum vessel and final leak test was performed after the thermal shock. Commissioning of the in-vessel cryo-pump was carried out using a temporary liquid helium supply system.

  20. Investigation of the cooling film distribution in liquid rocket engine

    Directory of Open Access Journals (Sweden)

    Luís Antonio Silva

    2011-05-01

    Full Text Available This study presents the results of the investigation of a cooling method widely used in the combustion chambers, which is called cooling film, and it is applied to a liquid rocket engine that uses as propellants liquid oxygen and kerosene. Starting from an engine cooling, whose film is formed through the fuel spray guns positioned on the periphery of the injection system, the film was experimentally examined, it is formed by liquid that seeped through the inner wall of the combustion chamber. The parameter used for validation and refinement of the theoretical penetration of the film was cooling, as this parameter is of paramount importance to obtain an efficient thermal protection inside the combustion chamber. Cold tests confirmed a penetrating cold enough cooling of the film for the length of the combustion chamber of the studied engine.

  1. Analysis of stress in reactor core vessel under effect of pressure lose shock wave

    International Nuclear Information System (INIS)

    Li Yong; Liu Baoting

    2001-01-01

    High Temperature gas cooled Reactor (HTR-10) is a modular High Temperature gas cooled Reactor of the new generation. In order to analyze the safety characteristics of its core vessel in case of large rupture accident, the transient performance of its core vessel under the effect of pressure lose shock wave is studied, and the transient pressure difference between the two sides of the core vessel and the transient stresses in the core vessel is presented in this paper, these results can be used in the safety analysis and safety design of the core vessel of HTR-10. (author)

  2. An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing

    International Nuclear Information System (INIS)

    Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T.

    1996-05-01

    The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE's Office of Nuclear Energy, Science and Technology; DOE's Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute's Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454 degrees C [850'F], all sensors measured the same temperature within about ±5% (23.6 degrees C [42.5 degrees F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes

  3. Cooling device for reactor container

    International Nuclear Information System (INIS)

    Arai, Kenji.

    1996-01-01

    Upon assembling a static container cooling system to an emergency reactor core cooling system using dynamic pumps in a power plant, the present invention provides a cooling device of lowered center of gravity and having a good cooling effect by lowering the position of a cooling water pool of the static container cooling system. Namely, the emergency reactor core cooling system injects water to the inside of a pressure vessel using emergency cooling water stored in a suppression pool as at least one water source upon loss of reactor coolant accident. In addition, a cooling water pool incorporating a heat exchanger is disposed at the circumference of the suppression pool at the outside of the container. A dry well and the heat exchanger are connected by way of steam supply pipes, and the heat exchanger is connected with the suppression pool by way of a gas exhaustion pipe and a condensate returning pipeline. With such a constitution, the position of the heat exchanger is made higher than an ordinary water level of the suppression pool. As a result, the emergency cooling water of the suppression pool water is injected to the pressure vessel by the operation of the reactor cooling pumps upon loss of coolant accident to cool the reactor core. (I.S.)

  4. Three Mile Island unit 2 vessel investigation project. Conclusions and significance

    International Nuclear Information System (INIS)

    Beckjord, E.S.

    1994-01-01

    At the conclusion of the TMI-2 Vessel Investigation Project, additional insights about the accident have been gained, specifically in the area of reactor vessel integrity and the conditions of the lower head of the reactor vessel. This paper discusses three topics: the evolving views about the TMI-2 accident scenario over time, the technical conclusions of the TMI-2 VIP (recovery of samples from the vessel lower head), and the broad significance of these findings (accident management). 4 refs

  5. Investigation of vessel visibility of iterative reconstruction method in coronary computed tomography angiography using simulated vessel phantom

    International Nuclear Information System (INIS)

    Inoue, Takeshi; Uto, Fumiaki; Ichikawa, Katsuhiro; Hara, Takanori; Urikura, Atsushi; Hoshino, Takashi; Miura, Youhei; Terakawa, Syouichi

    2012-01-01

    Iterative reconstruction methods can reduce the noise of computed tomography (CT) images, which are expected to contribute to the reduction of patient dose CT examinations. The purpose of this study was to investigate impact of an iterative reconstruction method (iDose 4 , Philips Healthcare) on vessel visibility in coronary CT angiography (CTA) by using phantom studies. A simulated phantom was scanned by a CT system (iCT, Philips Healthcare), and the axial images were reconstructed by filtered back projection (FBP) and given a level of 1 to 7 (L1-L7) of the iterative reconstruction (IR). The vessel visibility was evaluated by a quantitative analysis using profiles across a 1.5-mm diameter simulated vessel as well as visual evaluation for multi planar reformation (MPR) images and volume rendering (VR) images in terms of the normalized-rank method with analysis of variance. The peak CT value of the profiles decreased with IR level and full width at half maximum of the profile also decreased with the IR level. For normalized-rank method, there was no statistical difference between FBP and L1 (20% dose reduction) for both MPR and VR images. The IR levels higher than L1 sacrificed the spatial resolution for the 1.5-mm simulated vessel, and their visual vessel visibilities were significantly inferior to that of the FBP. (author)

  6. Finite element analysis of thermal stresses of the reactor vessel in a severe light water reactor accident

    International Nuclear Information System (INIS)

    Borovkov, A.I.; Semenov, A.S.; Granovsky, V.S.; Kovtunova, S.V.

    1995-01-01

    The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs

  7. Finite element analysis of thermal stresses of the reactor vessel in a severe light water reactor accident

    Energy Technology Data Exchange (ETDEWEB)

    Borovkov, A.I.; Semenov, A.S. [St. Petersburg State Technical Univ. (Russian Federation); Granovsky, V.S.; Kovtunova, S.V. [Research Inst. of Technology, Sosnovy Bor (Russian Federation)

    1995-12-31

    The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs.

  8. Gamma dose rate estimation and operation management suggestions for decommissioning the reactor pressure vessel of HTR-PM

    Energy Technology Data Exchange (ETDEWEB)

    Sheng Fang; Hong Li; Jianzhu Cao; Wenqian Li; Feng Xie; Jiejuan Tong [Institute of Nuclear and New Energy Technology, Tsinghua, University, Beijing (China)

    2013-07-01

    China is now designing and constructing a high temperature gas cooled reactor-pebble bed module (HTR-PM). In order to investigate the future decommissioning approach and evaluate possible radiation dose, gamma dose rate near the reactor pressure vessel was calculated for different cooling durations using QAD-CGA program. The source term of this calculation was provided by KORIGEN program. Based on the calculated results, the spatial distribution and temporal changes of gamma dose rate near reactor pressure vessel was systematically analyzed. A suggestion on planning decommissioning operation of reactor pressure vessel of HTRPM was given based on calculated dose rate and the Chinese Standard GB18871-2002. (authors)

  9. Emergency core cooling system in BWR type reactors

    International Nuclear Information System (INIS)

    Takizawa, Yoji

    1981-01-01

    Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

  10. Evaluation of heat exchange performance for the auxiliary component cooling water system cooling tower in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Kameyama, Yasuhiko; Shimizu, Atsushi; Inoi, Hiroyuki; Yamazaki, Kazunori; Shimizu, Yasunori; Aragaki, Etsushi; Ota, Yukimaru; Fujimoto, Nozomu

    2006-09-01

    The auxiliary component cooling water system (ACCWS) is one of the cooling system in High Temperature Engineering Test Reactor (HTTR). The ACCWS has main two features, many facilities cooling, and heat sink of the vessel cooling system which is one of the engineering safety features. Therefore, the ACCWS is required to satisfy the design criteria of heat removal performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the ACCWS cooling tower was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that ACCWS cooling tower has the required heat exchange performance in the design. (author)

  11. Advanced in-vessel retention design for next generation risk management

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    1998-12-31

    In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

  12. Advanced in-vessel retention design for next generation risk management

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y.; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    1997-12-31

    In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

  13. Evaluation of In-Vessel Corium Retention under a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon; Kang, Kyung-Ho; Ha, Kwang-Soon; Kim, Jong-Tae; Koo, Kil-Mo; Cho, Young-Ro; Hong, Seong-Wan; Kim, Sang-Baik; Kim, Hee-Dong

    2008-02-15

    The current study on In-Vessel corium Retention and its application activities to the actual nuclear power plant have been reviewed and discussed in this study. Severe accident sequence which determines an initial condition of the IVR has been evaluated and late phase melt progression, heat transfer on the outer reactor vessel, and in-vessel corium cooling mechanism have been estimated in detail. During the high pressure sequence of the reactor coolant system, a natural circulation flow of the hot steam leads to a failure of the pressurizer surge line before the reactor vessel failure, which leads to a rapid decrease of the reactor coolant system pressure. The results of RASPLAV/MASCA study by OECD/NEA have shown that a melt stratification has occurred in the lower plenum of the reactor vessel. In particular, laver inversion has occurred, which is that a high density of the metal melt moves to the lower part of the oxidic melt layer. A method of heat transfer enhancement on the outer reactor vessel is an optimal design of the reactor vessel insulation for an increase of the natural circulation flow between the outer reactor vessel and the its insulation, and an increase of the critical Heat flux on the outer reactor vessel by using various method, such as Nono fluid, coated reactor vessel, and so on. An increase method of the in-vessel melt cooling is a development of the In-vessel core catcher and a decrease of focusing effect in the metal layer.

  14. Outline of design, manufacturing and installation experience of pressure vessel structure for the prototype heavy water moderated boiling light water cooled reactor 'FUGEN'

    International Nuclear Information System (INIS)

    Shibato, Eizo; Oguchi, Isao; Kishi, Toshikazu; Kitagawa, Yuji

    1977-01-01

    After component installation completed in June 1977 and various functional tests to be conducted later, the prototype heavy water moderated, boiling light water cooled reactor ''FUGEN'' is scheduled to reach first criticality in March 1978. Since the pressure vessel of ''FUGEN'' is completely different from that of a light water reactor in structure and materials, through research and development work was carried out prior to fabrication and construction. Based on these studies, installation of the actual pressure vessel was completed. Functional tests are now under way. This article describes examples in which our research and development results are reflected on design, manufacture, and installation of the pressure vessel. Also it introduces noteworthy achievements relevant to production techniques in manufacture and installation. (auth.)

  15. Emergency core cooling device

    International Nuclear Information System (INIS)

    Suzaki, Kiyoshi; Inoue, Akihiro.

    1979-01-01

    Purpose: To improve core cooling effect by making the operation region for a plurality of water injection pumps more broader. Constitution: An emergency reactor core cooling device actuated upon failure of recycling pipe ways is adapted to be fed with cooling water through a thermal sleeve by way of a plurality of water injection pump from pool water in a condensate storage tank and a pressure suppression chamber as water feed source. Exhaust pipes and suction pipes of each of the pumps are connected by way of switching valves and the valves are switched so that the pumps are set to a series operation if the pressure in the pressure vessel is high and the pumps are set to a parallel operation if the pressure in the pressure vessel is low. (Furukawa, Y.)

  16. Investigation on integrity of JMTR concrete structures, cooling system and utility facilities

    International Nuclear Information System (INIS)

    Ebisawa, Hiroyuki; Tobita, Kenji; Fukasaku, Akitomi; Kaminaga, Masanori

    2010-02-01

    The condition of facilities and components to be used for re-operation of the Japan Materials Testing Reactor (JMTR) from FY2011, was investigated before the refurbishment work. An investigation of aged components (aged-investigation) was carried out for concrete structures of the JMTR reactor building, exhaust stack, trench, canal, filter banks and for aged components of tanks in the primary cooling system, heat exchangers, pipes in the secondary cooling system, cooling tower, emergency generators and so on, in order to identify their integrity. The aged-investigation was carried out from the beginning of FY2007. As a result, cracks of concrete structures such as the exhaust stack, a foundation of the UCL (Utility Cooling Line) elevated water tank were repaired and pipe linings of secondary cooling system were replaced. Motors of primary cooling pumps, pumps in the secondary cooling system and in other systems were decided to replace from viewpoints of future maintenance and improvement of reliability. Other components and the reactor building were decided to use continuously for a long-term by appropriate maintenance activities based on the long-term maintenance plan. In this paper, the aged-investigation for the JMTR reactor building, heat exchangers and emergency generators is presented. (author)

  17. RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Chang Oh; Cliff Davis; Goon C. Park

    2007-01-01

    A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls

  18. Improvements in or relating to cooling systems for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Ljubivy, A.G.; Batjukov, V.I.; Shkhian, T.G.; Fadeev, A.I.

    1980-01-01

    A cooling system is proposed which can be used to cool a set of nuclear fuel assemblies arranged in a reactor core or placed in a container for spent fuel assemblies. The object of the invention is to provide a system which would prevent leakage of coolant from the vessel in the event of a rupture of the coolant supply pipeline externally of the vessel. In the case of the reactor cooling system the level of the coolant is stopped from dropping below the level of the active portion of the fuel assemblies and thus prevents a breakdown of the reactor. (UK)

  19. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  20. In-vessel coolability and steam explosion in Nordic BWRs

    International Nuclear Information System (INIS)

    Ma, W.; Hansson, R.; Li, L.; Kudinov, P.; Cadinu, F.; Tran, C-.T.

    2010-05-01

    The INCOSE project is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in Nordic BWR plants with the cavity flooding as a severe accident management (SAM) measure. During 2009 substantial advances and new insights into physical mechanisms were gained for studies of: (i) in-vessel corium coolability - development of the methodologies to assess the efficiency of the control rod guide tube (CRGT) cooling as a potential SAM measure; (ii) debris bed coolability - characterization of the effective particle diameter of multi-size particles and qualification of friction law for two-phase flow in the beds packed with multi-size particles; and (iii) steam explosion - investigation of the effect of binary oxides mixtures properties on steam explosion. An approach for coupling of ECM/PECM models with RELAP5 was developed to enhance predictive fidelity for melt pool heat transfer. MELCOR was employed to examine the CRGT cooling efficiency by considering an entire accident scenario, and the simulation results show that the nominal flowrate (∼10kg/s) of CRGT cooling is sufficient to maintain the integrity of the vessel in a BWR of 3900 MWth, if the water injection is activated no later than 1 hour after scram. The POMECO-FL experimental data suggest that for a particulate bed packed with multi-size particles, the effective particle diameter can be represented by the area mean diameter of the particles, while at high velocity (Re>7) the effective particle diameter is closer to the length mean diameter. The pressure drop of two-phase flow through the particulate bed can be predicted by Reed's model. The steam explosion experiments performed at high melt superheat (>200oC) using oxidic mixture of WO3-CaO didn't detect an apparent difference in steam explosion energetics and preconditioning between the eutectic and noneutectic melts. This points out that the next step of MISTEE experiment will be conducted at lower superheat. (author)

  1. Investigation of Water-spray Cooling of Turbine Blades in a Turbojet Engine

    Science.gov (United States)

    Freche, John C; Stelpflug, William J

    1953-01-01

    An analytical and experimental investigation was made with a J33-A-9 engine to determine the effectiveness of spray cooling as a means of increasing thrust by permitting engine operation at inlet-gas temperatures and speeds above rated. With the assumption of adequate spray cooling at a coolant-to-gas flow ratio of 3 percent, calculations for the sea-level static condition indicated a thrust may be achieved by engine operation at an inlet-gas temperature of 2000 degrees F and an overspeed of 10 percent. Of the water-injection configurations investigated experimentally, those located in the inner ring of the stator diaphragm provided the best cooling at rated engine speed.

  2. Distribution of the In-Vessel Diagnostics in ITER Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    González, Jorge, E-mail: Jorge.Gonzalez@iter.org [Rüecker Lypsa, Carretera del Prat, 65, Cornellá de Llobregat (Spain); Clough, Matthew; Martin, Alex; Woods, Nick; Suarez, Alejandro [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France); Martinez, Gonzalo [Technical University Of Catalonia (UPC), Barcelona-Tech, Barcelona (Spain); Stefan, Gicquel; Yunxing, Ma [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France)

    2017-01-15

    The ITER In-Vessel Diagnostics have been distributed around the In-Vessel shell to understand burning plasma physics and assist in machine operation. Each diagnostics component has its own requirements, constraints, and even exclusion among them for the highly complex In-Vessel environment. The size of the plasma, the requirement to be able to align the blanket system to the magnetic centre of the machine, the cooling requirements of the blanket system and the size of the pressure vessel itself all add to the difficulties of integrating these systems into the remaining space available. The available space for the cables inside the special trays (in-Vessel looms) is another constraint to allocate In-Vessel electrical sensors. Besides this, there are issues with the Assembly sequences and surface & volumetric neutron heating considerations that have imposed several additional restrictions.

  3. Emergency core cooling system

    International Nuclear Information System (INIS)

    Arai, Kenji; Oikawa, Hirohide.

    1990-01-01

    The device according to this invention can ensure cooling water required for emerency core cooling upon emergence such as abnormally, for example, loss of coolant accident, without using dynamic equipments such as a centrifugal pump or large-scaled tank. The device comprises a pressure accumulation tank containing a high pressure nitrogen gas and cooling water inside, a condensate storage tank, a pressure suppression pool and a jet stream pump. In this device there are disposed a pipeline for guiding cooling water in the pressure accumulation tank as a jetting water to a jetting stream pump, a pipeline for guiding cooling water stored in the condensate storage tank and the pressure suppression pool as pumped water to the jetting pump and, further, a pipeline for guiding the discharged water from the jet stream pump which is a mixed stream of pumped water and jetting water into the reactor pressure vessel. In this constitution, a sufficient amount of water ranging from relatively high pressure to low pressure can be supplied into the reactor pressure vessel, without increasing the size of the pressure accumulation tank. (I.S.)

  4. Integration of cooking and vacuum cooling of carrots in a same vessel Integração dos processos de cozimento e resfriamento a vácuo de cenouras em um mesmo tanque

    Directory of Open Access Journals (Sweden)

    Luiz Gustavo Gonçalves Rodrigues

    2012-03-01

    Full Text Available Cooked vegetables are commonly used in the preparation of ready-to-eat foods. The integration of cooking and cooling of carrots and vacuum cooling in a single vessel is described in this paper. The combination of different methods of cooking and vacuum cooling was investigated. Integrated processes of cooking and vacuum cooling in a same vessel enabled obtaining cooked and cooled carrots at the final temperature of 10 ºC, which is adequate for preparing ready-to-eat foods safely. When cooking and cooling steps were performed with the samples immersed in boiling water, the effective weight loss was approximately 3.6%. When the cooking step was performed with the samples in boiling water or steamed, and the vacuum cooling was applied after draining the boiling water, water loss ranged between 15 and 20%, which caused changes in the product texture. This problem can be solved with rehydration using a small amount of sterile cold water. The instrumental textural properties of carrots samples rehydrated at both vacuum and atmospheric conditions were very similar. Therefore, the integrated process of cooking and vacuum cooling of carrots in a single vessel is a feasible alternative for processing such kind of foods.Para a preparação de refeições rápidas é comum o uso de legumes cozidos. A integração dos processos de cozimento e resfriamento de cenouras em um mesmo tanque pelo uso do resfriamento a vácuo é descrito neste artigo. A combinação de diferentes métodos de cozimento e resfriamento a vácuo foi investigada. O processo integrado de cozimento-resfriamento a vácuo em um mesmo tanque permitiu obter cenouras cozidas-resfriadas com temperaturas finais de 10 ºC, o que é adequado à preparação de refeições rápidas com segurança. Quando o processo de cozimento-resfriamento foi realizado com amostras imersas em água de cozimento, a perda efetiva de massa foi de aproximadamente 3,6%. Quando o processo de cozimento-resfriamento foi

  5. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 4. Numerical analysis of 1/10 scaled water experiment with the AQUA code

    International Nuclear Information System (INIS)

    Muramatu, Toshiharu; Yamaguchi, Akira

    2004-01-01

    A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. A numerical analysis was carried out with a multi-dimensional code AQUA to confirm an applicability to the evaluations for the in-vessel thermohydraulic phenomena using a 1/10 scaled water experiment simulating the large-scale fast breeder reactor in the feasibility studies. From the analysis, the following results were obtained. (1) In-vessel thermohydraulics characterized by a radiated flow pattern to the reactor vessel wall and a strong upward flow through a slit of the upper core structures were evaluated. These characteristics agreed approximately with the water experiment. (2) The upward velocity values at the slit agreed well with the experimental data under a condition of γ z = 0.3 and ξ z = 0.5, though overall evaluations of the in-vessel thermohydraulics were failed to predict quantitatively. (3) The AQUA code is applicable to the in-vessel thermohydraulics evaluations in the feasibility studies, though it is necessary to make further modifications of the calculational models for accurate evaluations. On the one hand, it was confirmed that calculated results for the 1/10 water experimental model and the 1/1 actual-scaled model agreed quantitatively for the in-vessel thermohydraulics characteristics indicated above. (author)

  6. A water inner circulation device for a reactor vessel

    International Nuclear Information System (INIS)

    Eriksson, O.

    1976-01-01

    A water inner circulation device for a reactor vessel comprising a pump mounted in the reactor vessel and driven by a water-cooled electric motor mounted in a housing outside the reactor vessel, the shaft of the pump passing through the reactor-vessel bottom and being coupled to the motor shaft in a member mechanically connected to the bottom of the reactor vessel in the vicinity of the motor housing, the pump shaft being surrounded by a resilient sealing ring, the reactor vessel communicating with the cooling channels of the pump, when the latter is operating, via a slot surrounding the pump hollow cylindrical shaft, characterized in that the slot inner end is used for/forming a circular space surrounding the pump shaft and surrounded by the motorhousing, in which is coaxially mounted a separating cylindral wall, the upper edge of which is tightly applied against the inner wall of the motor-housing to which it is fastened vertically, the inner surface of said wall being turned towards the outer surface of a circular packing-box, the outer surface of said separating wall constituting a separating radical inner surface for a circular chamber through which flow the motor cooling water. (author)

  7. An investigation of the flow dependence of temperature gradients near large vessels during steady state and transient tissue heating

    International Nuclear Information System (INIS)

    Kolios, M.C.; Worthington, A.E.; Hunt, J.W.; Holdsworth, D.W.; Sherar, M.D.

    1999-01-01

    Temperature distributions measured during thermal therapy are a major prognostic factor of the efficacy and success of the procedure. Thermal models are used to predict the temperature elevation of tissues during heating. Theoretical work has shown that blood flow through large blood vessels plays an important role in determining temperature profiles of heated tissues. In this paper, an experimental investigation of the effects of large vessels on the temperature distribution of heated tissue is performed. The blood flow dependence of steady state and transient temperature profiles created by a cylindrical conductive heat source and an ultrasound transducer were examined using a fixed porcine kidney as a flow model. In the transient experiments, a 20 s pulse of hot water, 30 deg. C above ambient, heated the tissues. Temperatures were measured at selected locations in steps of 0.1 mm. It was observed that vessels could either heat or cool tissues depending on the orientation of the vascular geometry with respect to the heat source and that these effects are a function of flow rate through the vessels. Temperature gradients of 6 deg. C mm -1 close to large vessels were routinely measured. Furthermore, it was observed that the temperature gradients caused by large vessels depended on whether the heating source was highly localized (i.e. a hot needle) or more distributed (i.e. external ultrasound). The gradients measured near large vessels during localized heating were between two and three times greater than the gradients measured during ultrasound heating at the same location, for comparable flows. Moreover, these gradients were more sensitive to flow variations for the localized needle heating. X-ray computed tomography data of the kidney vasculature were in good spatial agreement with the locations of all of the temperature variations measured. The three-dimensional vessel path observed could account for the complex features of the temperature profiles. The flow

  8. Assessment of reactor vessel integrity (ARVI)

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R. E-mail: sehgal@ne.kth.se; Theerthan, A.; Giri, A.; Karbojian, A.; Willschuetz, H.G.; Kymaelaeinen, O.; Vandroux, S.; Bonnet, J.M.; Seiler, J.M.; Ikkonen, K.; Sairanen, R.; Bhandari, S.; Buerger, M.; Buck, M.; Widmann, W.; Dienstbier, J.; Techy, Z.; Kostka, P.; Taubner, R.; Theofanous, T.; Dinh, T.N

    2003-04-01

    The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme. Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.

  9. Research and development of the prestressed concrete reactor vessel

    International Nuclear Information System (INIS)

    Shiozawa, Shoji; Omata, Ippei; Nakamura, Norio

    1975-01-01

    Compared with the steel reactor vessel, the prestressed concrete reactor vessel (PCRV) is said to be superior in safety and economy. One of the characteristics of the high temperature gas cooled reactor (HTGR) is the adoption of the PCRV instead of the steel reactor vessel to ensure safety. In order to improve safety characteristics, it is necessary for the PCRV to be provided with more reliable functions. When the multi-purpose HTGR or the gas cooled fast breeder reactor (GCFR) are realized in future, more severe conditions of technology will be imposed on the PCRV, and accordingly, technical developments are now increasingly required. IHI is now proceeding with the technical research and development on the PCRV, in which a basic study of its liner cooling system has already been completed. In this study applying a large cylindrical PCRV model, comparison was made between experimental data and analyses concerning the liner cooling system, and the results of analytical technique have been evaluated. The analytical technique established this time is applicable to the estimation of temperature distribution in the concrete of a large PCRV and also to the evaluation of the liner cooling system. (auth.)

  10. Investigations on efficiency of the emergency cooling by means of large-scale tests

    International Nuclear Information System (INIS)

    Hicken, E.F.

    1982-01-01

    The RSK guidelines contain the maximum permissible loads (max. cladding tube temperature 1200 0 C, max. Zr/H 2 O-reaction of 1% Zr). Their observance implies that only a small number of fuel rods fail. The safety research has to produce the evidence that the limiting loads are not exceeded. The analytical investigations on the emergency cooling behaviour could so far only be verified in scaled-down test facilities. After about 100 tests in four different large-scale test facilities the experimental investigations on the blow-down phase for large cracks are finished in the main. With the refill- and flood process the systems behaviour in scaled down test stands, the multidimensional conditions in the reactor pressure vessel can, however, only be simulated on the original scale. More experiments are planned as part of the 2D/3D-project (CCTF , SCTF, UPTF) and as part of the PKL-tests, so that more than 200 tests in seven plants will be available then. As to the small cracks the physical phenomena are known. The current investigations are used to increase the reliability of statement. After their being finished approximately 300 tests in seven plants will be available. (orig./HP) [de

  11. Study on decay heat removal capability of reactor vessel auxiliary cooling system

    International Nuclear Information System (INIS)

    Nishi, Y.; Kinoshita, I.

    1991-01-01

    The reactor vessel auxiliary cooling system (RVACS) is a simple, Passive decay heat removal system for an LMFBR. However, the heat removal capacity of this system is small compared to that of an immersed type of decay heat exchanger. In this study, a high-porosity porous body is proposed to enhance the RVACS's heat transfer performance to improve its applicability. The objectives of this study are to propose a new method which is able to use thermal radiation effectively, to confirm its heat removal capability and to estimate its applicability limit of RVACS for an LMFBR. Heat transfer tests were conducted in an experimental facility with a 3.5 m heat transfer height to evaluate the heat transfer performance of the high-porosity porous body. Using the experimental results, plant transient analyses were performed for a 300 MWe pool type LMFBR under a Total Black Out (TBO) condition to confirm the heat removal capability. Furthermore, the relationship between heat removal capability and thermal output of a reactor were evaluated using a simple parameter model

  12. Simulation of Two-Phase Natural Circulation Loop for Core Cather Cooling Using Air Water

    International Nuclear Information System (INIS)

    Revankar, S. T.; Huang, S. F.; Song, K. W.; Rhee, B. W.; Park, R. J.; Song, J. H.

    2012-01-01

    A closed loop natural circulation system employs thermally induced density gradients in single phase or two-phase liquid form to induce circulation of the working fluid thereby obviating the need for any mechanical moving parts such as pumps and pump controls. This increases the reliability and safety of the cooling system and reduces installation, operation and maintenance costs. That is the reason natural circulation cooling has been considered in advanced reactor core cooling and in engineered safety systems. Natural circulation cooling has been proposed to remove reactor decay heat by external vessel cooling for in-vessel core retention during sever accident scenario. Recently in APR1400 reactor core catcher design natural circulation cooling is proposed to stabilize and cool the corium ejected from the reactor vessel following core melt and breach of reactor vessel. The natural circulation flow is similar to external vessel cooling where water flows through an inclined narrow gap below hot surface and is heated to produce boiling. The two-phase natural circulation enables cooling of the corium pool collected on core catcher. Due to importance of this problem this paper focuses simulation of the two-phase natural circulation through inclined gap using air-water system. Scaling criteria for air-water loop are derived that enable simulation of the flow regimes and natural circulation flow rates in such systems using air-water system

  13. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

    Larson, E.M.; Elliott, R.D.

    1971-01-01

    System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

  14. Analysis of emergency core cooling capability of direct vessel vertical injection using CFX

    International Nuclear Information System (INIS)

    Yoon, Sang H.; Yu, Yong H.; Suh, Kune Y.

    2003-01-01

    More reliable and efficient safety injection system is of utmost importance in the design of advanced reactors such as the APR1400 (Advanced Power Reactor 1400 MWe). In this work, a new idea is proposed to inject the Emergency Core Cooling (ECC) water utilizing a dedicated nozzle with a vertically downward elbow. The Direct Vessel Injection (DVI) system is located horizontally above the cold leg in the APR1400. However, the horizontal injection method may not always satisfy the ECC penetration requirement into the core on account of rather involved multidimensional thermal and hydraulic phenomena occurring in the annular reactor downcomer such as bypass, impingement, entrainment and sweepout, condensation oscillation, etc. Thus, a novel concept is called for from the reactor safety point of view. The Direct Vessel Vertical Injection (DVVI) system is one of these efforts to penetrate as much the ECC water through the downcomer into the core as is practically achievable. The DVVI system can increase the momentum of the downward flow, thus minimizing the effect of water impingement on the core barrel and the direct bypass though the break. To support the claim of increased downward momentum of flow in the DVVI system, computational fluid dynamics analyses were performed using CFX. The new concept of the DVVI system, which can certainly help increase the core thermal margin, is found to be more efficient than DVI. If the structural problem in the manufacturing process is properly solved, this concept can safely be applied in the advanced nuclear reactor design

  15. Analysis of AP1000{sup TM} reactor vessel cavity and support cooling

    Energy Technology Data Exchange (ETDEWEB)

    Craig, K.J. [Westinghouse Electric South Africa, 32 Park Avenue North, Highway Business Park, Centurion, 0157 (SOUTH AFRICA); Harkness, A.W. [Nuclear Power Plants, Westinghouse Electric Company, LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Kritzinger, H.P.; Hoffmann, J.E. [Pebble Bed Modular Reactor (Pty) Ltd, 1279 Mike Crawford Avenue, Centurion (South Africa)

    2010-07-01

    The paper investigates a Computational Fluid Dynamic (CFD) analysis of the air cooling of the Reactor Vessel (RV) cavity and RV supports. All the Heating, Ventilation and Air Conditioning (HVAC) flow of the RV cavity has to pass through the four RV supports supporting the four cold legs (cold inlets from the two steam generators) of the AP1000{sup TM} reactor. The RV support has a complex flow path leading to significant pressure drops to provide the necessary cooling. The insulation surrounding the RV has a specification on the amount of heat that may be transferred (lost) from the RV in order to maximize the heat transfer to the coolant driving the steam generators. This heat loss is applied as a boundary condition to the solution domain. Another heat source that is considered is that due to nuclear heating. Due to the fact that the heat source is nuclear in nature, gamma and neutron heating have to be considered for the surrounding structures. These include the carbon steel structural module that encapsulates the RV cavity, as well as the concrete poured around this module. The space in the gap between the RV insulation and the structural module steel shell is not only obstructed by the insulation supports, but also by wells or tubes within which power and intermediate ex-core detectors are located. Source-range ex-core detectors are embedded in the concrete surrounding the structural module. All these detectors have a limited operating temperature range, and together with limits on concrete temperatures for safety considerations, necessitate the need for CFD simulations to determine the range of operational temperatures seen by these components. The CFD simulations also provide an estimate of the pressure drop through the cavity between the RV insulation and structural module, as well as that through the four RV supports. Results presented include ANSYS{sup R} FLUENT{sup R} simulations describing the modelling procedure that was followed, namely to combine

  16. OECD/CSNI Workshop on In-Vessel Core Debris Retention and Coolability - Summary and Conclusions

    International Nuclear Information System (INIS)

    Behbahani, Ali-Reza; Drozd, Andrzej; Kim, Sang-Baik; Micaelli, Jean-Claude; Okkonen, Timo; Sugimoto, Jun; Trambauer, Klaus; Tuomisto, Harri

    1999-01-01

    In the spring of 1994 an OECD Workshop on Large Pool Heat transfer was held in Grenoble. The scope of this workshop was the investigation of (1) molten pool heat transfer, (2) heat transfer to the surrounding water, and (3) the feasibility of in-vessel core debris cooling through external cooling of the vessel. Since this time, experimental test series have been completed (e.g., COPO, ULPU, CORVIS) and new experimental programs (e.g., BALI, SONATA, RASPLAV, debris and gap heat transfer) have been established to consolidate and expand the data base for further model development and to improve the understanding of in-vessel debris retention and coolability in a nuclear power plant. Discussions within the CSNI's PWG-2 and the Task Group on Degraded Core Cooling (TG-DCC) have led to the conclusion that the time was ripe for organizing a new international Workshop with the objectives: - to review the results of experimental research that has been conducted in this area; - to exchange information on the results of member countries experiments and model development on in-vessel core debris retention and coolability; - to discuss areas where additional experimental research is needed in order to provide an adequate data base for analytical model development for core debris retention and coolability. The scope of this workshop was limited to the phenomena connected to in-vessel core debris retention and coolability and did not include steam explosion and fission product issues. The workshop was structured into the following sessions: Key note papers; Experiments and model development; Debris bed heat transfer; Corium properties, molten pool convection and crust formation; Gap formation and gap cooling; Creep behaviour of reactor pressure vessel lower head; Ex-vessel boiling and critical heat flux phenomena; Scaling to reactor severe accident conditions and reactor applications. Compared to the previous workshop held in Grenoble in 1994, large progress has been made in the

  17. Testing plan for critical heat flux measurement during in-vessel retention

    International Nuclear Information System (INIS)

    Aoki, Kazuyoshi; Iwaki, Chikako; Sato, Hisaki; Mimura, Satoshi; Kanamori, Daisuke

    2015-01-01

    In-Vessel Retention (IVR) is a method to maintain molten debris in a reactor vessel (RV) by RV outer surface cooling. Structural integrity of RV and cooling capacity on RV outer surface are important to verify IVR strategy. Critical Heat Flux (CHF) data is necessary to estimate cooling capacity on the RV outer surface. And there are some CHF data to estimate cooling capacity on the RV outer surface. However, these data were obtained for specific plants. Thus, the objective of this study is developing a CHF correlation for various PWR plants. The objectives of this paper are developing test equipment and testing plan for the CHF correlation. Firstly, plant conditions during severe accidents were organized. Then, ranges of testing parameters were estimated with the plant conditions. And specifications of the test equipment were set to cover the range of parameters. Secondly, testing cases were set based on design of experiments. The test cases are suitable to develop experimental correlations. (author)

  18. PH adjustment of power plant cooling water with flue gas/fly ash

    Science.gov (United States)

    Brady, Patrick V.; Krumhansl, James L.

    2015-09-22

    A system including a vessel including a heat source and a flue; a turbine; a condenser; a fluid conduit circuit disposed between the vessel, the turbine and the condenser; and a diverter coupled to the flue to direct a portion of an exhaust from the flue to contact with a cooling medium for the condenser water. A method including diverting a portion of exhaust from a flue of a vessel; modifying the pH of a cooling medium for a condenser with the portion of exhaust; and condensing heated fluid from the vessel with the pH modified cooling medium.

  19. Study on Heat Transfer Characteristics of One Side Heated Vertical Channel Applied as Vessel Cooling System

    International Nuclear Information System (INIS)

    Kuriyama, Shinji; Takeda, Tetsuaki; Funatani, Shumpei

    2014-01-01

    The inherent properties of the Very-High-Temperature Reactor facilitate the design of the VHTR with high degree of passive safe performances, compared to other type of reactors. However; it is still not clear if the VHTR can maintain a passive safe function during the severe accident, or what would be a design criterion to guarantee the VHTR with the high degree of passive safe performances during the accidents. In the Very High Temperature Reactor (VHTR) which is a next generation nuclear reactor system, ceramics and graphite are used as a fuel coating material and a core structural material, respectively. Even if the depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change slowly. This is because the thermal capacity of the core is so large. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel (RPV). This study is to develop the passive cooling system for the VHTR using the vertical channel inserting porous materials. The objective of this study is to investigate heat transfer characteristics of natural convection of a one-side heated vertical channel inserting the porous materials with high porosity. In order to obtain the heat transfer and fluid flow characteristics of a vertical channel inserting porous material, we have also carried out a numerical analysis using the commercial CFD code. From the analytical results obtained in the natural convection cooling, an amount of removed heat enhanced inserting the copper wire. It was found that an amount of removed heat inserting the copper wire (porosity = 0.9972) was about 10% higher than that without the copper wire. This paper describes a thermal performance of the one-side heated vertical channel inserting copper wire with high porosity. (author)

  20. Validation of heat transfer models for gap cooling

    International Nuclear Information System (INIS)

    Okano, Yukimitsu; Nagae, Takashi; Murase, Michio

    2004-01-01

    For severe accident assessment of a light water reactor, models of heat transfer in a narrow annular gap between overheated core debris and a reactor pressure vessel are important for evaluating vessel integrity and accident management. The authors developed and improved the models of heat transfer. However, validation was not sufficient for applicability of the gap heat flux correlation to the debris cooling in the vessel lower head and applicability of the local boiling heat flux correlations to the high-pressure conditions. Therefore, in this paper, we evaluated the validity of the heat transfer models and correlations by analyses for ALPHA and LAVA experiments where molten aluminum oxide (Al 2 O 3 ) at about 2700 K was poured into the high pressure water pool in a small-scale simulated vessel lower head. In the heating process of the vessel wall, the calculated heating rate and peak temperature agreed well with the measured values, and the validity of the heat transfer models and gap heat flux correlation was confirmed. In the cooling process of the vessel wall, the calculated cooling rate was compared with the measured value, and the validity of the nucleate boiling heat flux correlation was confirmed. The peak temperatures of the vessel wall in ALPHA and LAVA experiments were lower than the temperature at the minimum heat flux point between film boiling and transition boiling, so the minimum heat flux correlation could not be validated. (author)

  1. Reactor vessel head permanent shield

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Leduc, R.J.; Richard, J.W.; Malandra, L.J.

    1989-01-01

    A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs

  2. Emergency core cooling system

    International Nuclear Information System (INIS)

    Kato, Ken.

    1989-01-01

    In PWR type reactors, a cooling water spray portion of emergency core cooling pipelines incorporated into pipelines on high temperature side is protruded to the inside of an upper plenum. Upon rupture of primary pipelines, pressure in a pressure vessel is abruptly reduced to generate a great amount of steams in the reactor core, which are discharged at a high flow rate into the primary pipelines on high temperature side. However, since the inside of the upper plenum has a larger area and the steam flow is slow, as compared with that of the pipelines on the high temperature side, ECCS water can surely be supplied into the reactor core to promote the re-flooding of the reactor core and effectively cool the reactor. Since the nuclear reactor can effectively be cooled to enable the promotion of pressure reduction and effective supply of coolants during the period of pressure reduction upon LOCA, the capacity of the pressure accumulation vessel can be decreased. Further, the re-flooding time for the reactor is shortened to provide an effect contributing to the improvement of the safety and the reduction of the cost. (N.H.)

  3. A computational study for investigating acoustic streaming and tissue heating during high intensity focused ultrasound through blood vessel with an obstacle

    Science.gov (United States)

    Parvin, Salma; Sultana, Aysha

    2017-06-01

    The influence of High Intensity Focused Ultrasound (HIFU) on the obstacle through blood vessel is studied numerically. A three-dimensional acoustics-thermal-fluid coupling model is employed to compute the temperature field around the obstacle through blood vessel. The model construction is based on the linear Westervelt and conjugate heat transfer equations for the obstacle through blood vessel. The system of equations is solved using Finite Element Method (FEM). We found from this three-dimensional numerical study that the rate of heat transfer is increasing from the obstacle and both the convective cooling and acoustic streaming can considerably change the temperature field.

  4. Experimental Investigation of Creep Behavior of Reactor Vessel Lower Head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1999-01-01

    The authors report a study which aimed at experimentally and numerically investigating and characterizing the failure of a reactor pressure vessel (RPV) lower head due to thermal and pressure loads generated by a severe accident. They present the experimental apparatus which is based on a scaled version of the lower part of a TMI-like reactor pressure vessel without vessel skirt. They report and comment the results obtained during the first five experiments: uniform heating and non penetrations, centre-peaked heat flux and no penetrations, edge-peaked heat flux and no penetrations, uniform heating with penetrations, edge-peaked heat flux with penetrations. They compare the third and fifth experience (those with edge-peaked heat flux)

  5. Pressurized wet digestion in open vessels (T11)

    International Nuclear Information System (INIS)

    Kettisch, P.; Maichin, P.; Zischka, M.; Knapp, G.

    2002-01-01

    Full text: Pressurized wet digestion in closed vessels, microwave assisted or with conventional conductive heating, is the most important sample preparation technique for digestion or leaching procedures in element analysis. In comparison to open vessel digestion closed vessel digestion methods have many advantages, but there is one disadvantage - complex and expensive vessel designs. A new technique - pressurized wet digestion in open vessels - combine the advantages of closed vessel sample digestion with the application of simple and cheap open vessels made of quartz or PFA. The vessels are placed in a high pressure Asher HPA, which is adapted with a Teflon liner and filled partly with water. The analytical results with 30 ml quartz vessels, 22 ml PFA vessels and 1.5 ml PIA auto sampler cups will be shown. In principle every dimensions of vessels can be used. The vessels are loaded with sample material (max. 1.5 g with quartz vessels, max. 0.5 g with PFA vessels and 50 mg with auto sampler cups) and digestion reagent. Afterwards the vessels are simply covered with PTFE stoppers and not sealed. The vessels are transferred into a special adapted HPA and digested at temperatures up to 270 o C. The digestion time is 90 min. and cooling down to room temperature 30 min. The analytical results of CRM's are within the certified values and no cross contamination and losses of volatile elements could be observed. (author)

  6. Nuclear reactor with a suspended vessel

    International Nuclear Information System (INIS)

    Lemercier, Guy.

    1977-01-01

    This invention relates to a nuclear reactor with a suspended vessel and applies in particular when this is a fast reactor, the core or active part of the reactor being inside the vessel and immersed under a suitable volume of flowing liquid metal to cool it by extracting the calories released by the nuclear fission in the fuel assemblies forming this core [fr

  7. Vascular Patterns in Iguanas and Other Squamates: Blood Vessels and Sites of Thermal Exchange.

    Directory of Open Access Journals (Sweden)

    William Ruger Porter

    Full Text Available Squamates use the circulatory system to regulate body and head temperatures during both heating and cooling. The flexibility of this system, which possibly exceeds that of endotherms, offers a number of physiological mechanisms to gain or retain heat (e.g., increase peripheral blood flow and heart rate, cooling the head to prolong basking time for the body as well as to shed heat (modulate peripheral blood flow, expose sites of thermal exchange. Squamates also have the ability to establish and maintain the same head-to-body temperature differential that birds, crocodilians, and mammals demonstrate, but without a discrete rete or other vascular physiological device. Squamates offer important anatomical and phylogenetic evidence for the inference of the blood vessels of dinosaurs and other extinct archosaurs in that they shed light on the basal diapsid condition. Given this basal positioning, squamates likewise inform and constrain the range of physiological thermoregulatory mechanisms that may have been found in Dinosauria. Unfortunately, the literature on squamate vascular anatomy is limited. Cephalic vascular anatomy of green iguanas (Iguana iguana was investigated using a differential-contrast, dual-vascular injection (DCDVI technique and high-resolution X-ray microcomputed tomography (μCT. Blood vessels were digitally segmented to create a surface representation of vascular pathways. Known sites of thermal exchange, consisting of the oral, nasal, and orbital regions, were given special attention due to their role in brain and cephalic thermoregulation. Blood vessels to and from sites of thermal exchange were investigated to detect conserved vascular patterns and to assess their ability to deliver cooled blood to the dural venous sinuses. Arteries within sites of thermal exchange were found to deliver blood directly and through collateral pathways. The venous drainage was found to have multiple pathways that could influence neurosensory

  8. Vascular Patterns in Iguanas and Other Squamates: Blood Vessels and Sites of Thermal Exchange.

    Science.gov (United States)

    Porter, William Ruger; Witmer, Lawrence M

    2015-01-01

    Squamates use the circulatory system to regulate body and head temperatures during both heating and cooling. The flexibility of this system, which possibly exceeds that of endotherms, offers a number of physiological mechanisms to gain or retain heat (e.g., increase peripheral blood flow and heart rate, cooling the head to prolong basking time for the body) as well as to shed heat (modulate peripheral blood flow, expose sites of thermal exchange). Squamates also have the ability to establish and maintain the same head-to-body temperature differential that birds, crocodilians, and mammals demonstrate, but without a discrete rete or other vascular physiological device. Squamates offer important anatomical and phylogenetic evidence for the inference of the blood vessels of dinosaurs and other extinct archosaurs in that they shed light on the basal diapsid condition. Given this basal positioning, squamates likewise inform and constrain the range of physiological thermoregulatory mechanisms that may have been found in Dinosauria. Unfortunately, the literature on squamate vascular anatomy is limited. Cephalic vascular anatomy of green iguanas (Iguana iguana) was investigated using a differential-contrast, dual-vascular injection (DCDVI) technique and high-resolution X-ray microcomputed tomography (μCT). Blood vessels were digitally segmented to create a surface representation of vascular pathways. Known sites of thermal exchange, consisting of the oral, nasal, and orbital regions, were given special attention due to their role in brain and cephalic thermoregulation. Blood vessels to and from sites of thermal exchange were investigated to detect conserved vascular patterns and to assess their ability to deliver cooled blood to the dural venous sinuses. Arteries within sites of thermal exchange were found to deliver blood directly and through collateral pathways. The venous drainage was found to have multiple pathways that could influence neurosensory tissue temperature

  9. Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, Tyler A.; Edwards, Michael J.

    2018-01-23

    A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

  10. Neutronics studies for the design of the European DEMO vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Flammini, Davide, E-mail: davide.flammini@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Moro, Fabio; Pizzuto, Aldo [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Bachmann, Christian [EUROfusion Consortium, Boltzmannstr. 2, 85748 Garching (Germany)

    2016-11-01

    Highlights: • MCNP calculation of nuclear heating, damage, helium production and neutron flux in DEMO HCLL and HCPB vacuum vessel at the inboard equatorial plane. • Study of impact of the poloidal gap between blanket modules, for several gap width, on vacuum vessel nuclear quantities. • Effect of the gap on nuclear heating result to be moderate, however high values of nuclear heating are found, even far from the gap with HCLL blanket. • Radiation damage limit of 2.75 DPA is met with a 1 cm wide gap. Helium production results very sensitive to the gap width. • Comparison between HCLL and HCPB blankets is shown for nuclear heating and neutron flux in the vacuum vessel. - Abstract: The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572 MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5 cm. High values of the nuclear heating (≈1 W/cm{sup 3}), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75 DPA on the vessel is almost met with 1 cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided.

  11. Thermal Load Analysis of Multilayered Corium in the Lower Head of Reactor Pressure Vessel during Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Whang, Seok Won; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of); Hwang, Tae Suk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    In-Vessel Retention (IVR) is one of the severe accident management strategies to terminate or mitigate the severe accident which is also called 'core-melt accident'. The reactor vessel would be cooled by flooding the cavity with water. The molten core mixture is divided into two or three layers due to the density difference. Light metal layer which contains Fe and Zr is on the oxide layer which is consist of UO{sub 2} and ZrO{sub 2}. Heavy metal layer which contains U, Fe and Zr is located under the oxide layer. In oxide layer, the crust which is solidified material is formed along the boundary. The assessment of IVR for nuclear power plant has been conducted with lumped parameter method by Theofanous, Rempe and Esmaili. In this paper, the numerical analysis was performed and verified with the Esmaili's work to analyze thermal load of multilayered corium in pressurized reactor vessel and also to examine the condition of in-vessel corium characteristic before the vessel failure that lead to ex-vessel severe accident progression for example, ex-vessel debris bed cooling. The in-vessel coolability analysis for several scenarios is conducted for the plant which has higher power than AP1000. Two sensitivity analyses are conducted, the first is emissivity of light metal layer and the second is the heat transfer coefficient correlations of oxide layer. The effect of three layered system also investigated. In this paper, the numerical analysis was performed and verified with Esmaili's model to analyze thermal load of multilayered corium in pressurized reactor vessel. For two layered system, thermal load was analyzed according to the severe accident scenarios, emissivity of the light metal layer and heat transfer correlations of the.

  12. Experimental analysis of ex-vessel core catcher cooling system performance for EU-APR1400 during severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Song, K. W.; Park, H. S.; Revankar, S. T. [POSTECH, Pohang (Korea, Republic of); Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In the coolant channel which has a unique design and large scale flow paths, natural circulation is passively activated by buoyancy driven force. Since two-phase flow behavior in a large scale channel is different from that in a small scale channel, the two-phase flow affecting the cooling capability is difficult to be predicted in the large channel. Therefore, cooling experiment in the core catcher coolant path is necessary. Cooling Experiment - Passive Ex-vessel corium retaining and Cooling System(CE-PECS) is constructed in full scale(in height and width) slice of half prototype. It actually simulates steam-water flow in the coolant channel for different decay heat condition of the corium. In this study, thermal power considering of total amount of decay heat 190 kW which corresponds to 40MW of thermal power in the prototype is loaded on the top wall of the CE-PECS coolant channel. Natural circulation flow rate and pressure drops at the two-phase region are measured in various power level. Temperatures of heater block and working fluid in various position along the flow path enable to calculate heat fluxes and heat transfer coefficients distribution. These results are used for evaluating heat removal capability of core catcher facility. Two-phase natural circulation experiment is carried out in CE-PECS facility. Based on the prototypic condition, 190 kW of total power is supplied to the top of the coolant path. Uniform distribution of heat load on the downward facing heater bock produces -300 kW/m2 at 100 % power ratio. Although the experiment should consider the heat loss and heat flux uniformity, several noticeable conclusions have been made as followings; 1. Mass flow rate and two-phase pressure drop are measured in various power conditions. 2. Slightly inclined top wall at the downstream of the channel shows better heat exchange performance than horizontal top wall because enhanced convection due to the increase of void fraction improves local cooling. This

  13. Development of LILAC-meltpool for the thermo-hydraulic analysis of core melt relocated in a reactor vessel

    International Nuclear Information System (INIS)

    Kim, Jong Tae; Kim, Sang Baik; Kim, Hee Dong

    2002-03-01

    LILAC-meltpool has been developed to study thermo-hydraulic behavior of molten pool and thermal behavior of vessel wall during severe accident. To validate LILAC-meltpool code several two and three dimensional thermo-hydraulic problems were selected and solved. The benchmark problems have experimental results or verified numerical results. Through the validation it was found that LILAC-meltpool reproduces very accurate numerical results. Two-layered semicircular pool was solved to study thermal and hydraulic characteristics of pool stratification. The LAVA experiment using alumina/ferrite molten pool was calculated and compared with computed results. Cooling of alumina/ferrite two-layered pool was affected by stratification. In the numerical results temperature of vessel inner was highest at a location below the interface. Crust was developed from upper surface and lower outer surface, but in the area near the interface corium simulant existed as molten state for long time. LAVA-4 experiment was studied using gap-cooling model in LILAC-meltpool code. Temperature increase of LAVA vessel after alumina melt relocation was strongly dependent on gap formation mechanism. Calculated cooling rates of the vessel were very similar to experimental results. For LAVA experiments which do not have heat generation coolant penetrates easily into a gap and it is found that gap-cooling is very effective for cooling of vessel, but it is thought that coolant penetration could be limited near upper part of gap because of decay heat and high temperature of corium crust

  14. Review of the TMI-2 accident evaluation and vessel investigation projects

    Energy Technology Data Exchange (ETDEWEB)

    Ladekarl Thomsen, Knud

    1998-03-01

    The results of the TMI-2 Accident Evaluation Programme and the Vessel Investigation Project have been reviewed as part of a literature study on core meltdown and in-vessel coolability. The emphasis is placed on the late phase melt progression, which is of special relevance to the NKS-sponsored RAK-2.1 project on Severe Accident Phenomenology. The body of the report comprises three main sections, The TMI-2 Accident Scenario, Core Region and Relocation Path Investigations, and Lower Head Investigations. In the final discussion, the lower head gap formation mechanism is explained in terms of thermal contraction and fracturing of the debris crust. This model seems more plausible than the MAAP model based on creep expansion of the lower head. (au) 1 tab., 33 ills., 31 refs.

  15. Minimization of radioactive material deposition in water-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Ruiz, C.P.; Blaies, D.M.

    1988-01-01

    This patent describes the method for inhibiting the deposition of radioactive cobalt in a water-bearing vessel of a water-cooled nuclear reactor which comprises adding zinc ion to water entering the water-bearing vessel. The improvement contains a substantially lower proportion of the /sup 64/Zn isotope than naturally occurring zinc

  16. Hydroaerothermal investigations conducted in the USSR to justify the construction of large cooling towers

    International Nuclear Information System (INIS)

    Goncharov, V.V.

    1989-01-01

    The multi-purpose task of improving water cooling systems of thermal and nuclear power plants is aimed at the development of efficient designs of cooling towers and other types of industrial coolers which call for comprehensive scientific justification. Cooling towers of 60-70 thou m 3 /h capacity with a chimney height of 130 m and those of 80-100 thou m 3 /h capacity with a chimney height of 150 m were developed. For circulating water systems of large power plants the design of a counterflow chimney cooling tower of 180 thou m 3 /h capacity has been recently developed. At present the work is being conducted on developing a new three-cell cooling tower featuring high reliability, operational flexibility and cost-effectiveness of the design. This cooling tower, besides having higher operating reliability than the conventional one of circular shape, can ensure the commissioning, current repairs and overhauls of water cooling arrangements in a cell-wise sequence, i.e. without shutting down the power generating units. Laboratory and field investigations of the spray-type cooling towers having no packing (fill), studies on heat and mass exchanges processes, aerodynamics of droplet flows and new designs of sprayers made it possible to come to a conclusion that their cooling capacity can be substantially increased and brought up to the level of the cooling towers with film packings. The pilot cooling towers were designed according to the counterflow, crossflow and cross-counterflow schemes. The basic investigation method remains to be the experimental one. On the test rigs and aerodynamic models the heat and mass transfer and aerodynamic resistance coefficients are determined. These studies and subsequent calculations are based on the heat balance equation

  17. Applications of nano-fluids to enhance LWR accidents management in in-vessel retention and emergency core cooling systems

    International Nuclear Information System (INIS)

    Chupin, A.; Hu, L. W.; Buongiorno, J.

    2008-01-01

    Water-based nano-fluid, colloidal dispersions of nano-particles in water; have been shown experimentally to increase the critical heat flux and surface wettability at very low concentrations. The use of nano-fluids to enhance accidents management would allow either to increase the safe margins in case of severe accidents or to upgrade the power of an existing power plant with constant margins. Building on the initial work, computational fluid dynamics simulations of the nano-fluid injection system have been performed to evaluate the feasibility of a nano-fluid injection system for in-vessel retention application. A preliminary assessment was also conducted on the emergency core cooling system of the European Pressurized Reactor (EPR) to implement a nano-fluid injection system for improving the management of loss of coolant accidents. Several design options were compared/or their respective merits and disadvantages based on criteria including time to injection, safety impact, and materials compatibility. (authors)

  18. Experimental Investigation of Double Effect Evaporative Cooling Unit

    Directory of Open Access Journals (Sweden)

    Ahmed Abd Mohammad Saleh

    2018-03-01

    Full Text Available This work presents the experimental investigation of double effect evaporative cooling unit with approximate capacity 7 kW. The unit consisted of two stages, the sensible heat exchanger and the cooling tower composing the external indirect regenerative evaporative cooling stage where a direct evaporative cooler represent the second stage. Testing results showed a maximum capacity and lowest supplied air temperature when the water flow rate in heat exchanger was 0.1 L/s. The experiment recorded the unit daily readings at two airflow rates (0.425 m3/s, 0.48 m3/s. The reading shows that unit inlet DBT is effect positively on unit wet bulb effectiveness and unit COP at constant humidity ratio. The air extraction ratio effected positively on the unit wet bulb effectiveness within a certain limit where maximum COP recorded 11.4 when the extraction ratio equal to 40%.

  19. Evaluation of a cavity flooding strategy for the prevention of reactor vessel failure in a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Je, Moo Sung; Park, Chang Kyoo [Korea Atomic Energy Research Institute, TaeJon (Korea, Republic of)

    1994-10-01

    As a part of the evaluation of accident management strategies for severe accident prevention or mitigation in a station blackout scenario for YGN 3 and 4, an external vessel cooling strategy for the prevention of reactor vessel failure has been estimated using the MAAP4 computer code. The sensitivity studies have been performed such as actuating timings and the number of spray pumps used. To explore external vessel cooling strategies, containment spray pumps were actuated by varying time spanning core uncovery, core melting and relocation of molten core material. It was shown that flooding of the reactor cavity using the containment spray system may prevent reactor vessel failure but may not prevent the failure of the relocation of molten core material during the station blackout sequence of YGN 3 and 4. Reactor vessel failure can be prevented by external vessel cooling using condensed water from the operation of two containment spray pumps at the time of core melting and using water from the operation of one containment spray pumps at the time of core melting and using water from the operation of one containment spray pump at the time of core uncovery. (Author) 46 refs., 26 figs., 5 tabs.

  20. Emergency reactor cooling device

    International Nuclear Information System (INIS)

    Arakawa, Ken.

    1993-01-01

    An emergency nuclear reactor cooling device comprises a water reservoir, emergency core cooling water pipelines having one end connected to a water feeding sparger, fire extinguishing facility pipelines, cooling water pressurizing pumps, a diesel driving machine for driving the pumps and a battery. In a water reservoir, cooling water is stored by an amount required for cooling the reactor upon emergency and for fire extinguishing, and fire extinguishing facility pipelines connecting the water reservoir and the fire extinguishing facility are in communication with the emergency core cooling water pipelines connected to the water feeding sparger by system connection pipelines. Pumps are operated by a diesel power generator to introduce cooling water from the reservoir to the emergency core cooling water pipelines. Then, even in a case where AC electric power source is entirely lost and the emergency core cooling system can not be used, the diesel driving machine is operated using an exclusive battery, thereby enabling to inject cooling water from the water reservoir to a reactor pressure vessel and a reactor container by the diesel drive pump. (N.H.)

  1. Debris interactions in reactor vessel lower plena during a severe accident. II. Integral analysis

    International Nuclear Information System (INIS)

    Suh, K.Y.; Henry, R.E.

    1996-01-01

    For pt.I see ibid., p.147-63, 1996. The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 vessel inspection program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation. (orig.)

  2. Prestressed reactor vessel for nuclear power plants

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1982-01-01

    With usual pressure vessels for nuclear reactor plants, especially for gas-cooled nuclear reactors, the load occurring due to the inner overpressure, especially the tensile load affecting the vessel top and/or bottom, their axis of inertia being horizontal, shall be compensated without a supplementary modification in design of the top and/or the bottom. This is attained by choosing an appropriate prestressing system of the vessel wall in the field the top and/or the bottom, so that the top and/or the bottom form a tension vault directed towards the interior of the vessel. (orig.) [de

  3. Design characteristics of pantograph type in vessel fuel handling system in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

  4. Design characteristics of pantograph type in vessel fuel handling system in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

  5. A homogeneous cooling scheme investigation for high power slab laser

    Science.gov (United States)

    He, Jianguo; Lin, Weiran; Fan, Zhongwei; Chen, Yanzhong; Ge, Wenqi; Yu, Jin; Liu, Hao; Mo, Zeqiang; Fan, Lianwen; Jia, Dan

    2017-10-01

    The forced convective heat transfer with the advantages of reliability and durability is widely used in cooling the laser gain medium. However, a flow direction induced temperature gradient always appears. In this paper, a novel cooling configuration based on longitudinal forced convective heat transfer is presented. In comparison with two different types of configurations, it shows a more efficient heat transfer and more homogeneous temperature distribution. The investigation of the flow rate reveals that the higher flow rate the better cooling performance. Furthermore, the simulation results with 20 L/min flow rate shows an adequate temperature level and temperature homogeneity which keeps a lower hydrostatic pressure in the flow path.

  6. A method for increasing the homogeneity of the temperature distribution during magnetic fluid hyperthermia with a Fe-Cr-Nb-B alloy in the presence of blood vessels

    Energy Technology Data Exchange (ETDEWEB)

    Tang, Yundong [College of Physics and Information Engineering, Fuzhou University, Fuzhou 350116 (China); Flesch, Rodolfo C.C. [Departamento de Automação e Sistemas, Universidade Federal de Santa Catarina, 88040-900 Florianópolis, SC (Brazil); Jin, Tao, E-mail: jintly@fzu.edu.cn [College of Electrical Engineering and Automation, Fuzhou University, Fuzhou 350116 (China)

    2017-06-15

    Highlights: • The effects of blood vessels on temperature field distribution are investigated. • The critical thermal energy of hyperthermia is computed by the Finite Element Analysis. • A treatment method is proposed by using the MNPs with low Curie temperature. • The cooling effects due to the blood flow can be controlled. - Abstract: Magnetic hyperthermia ablates tumor cells by absorbing the thermal energy from magnetic nanoparticles (MNPs) under an external alternating magnetic field. The blood vessels (BVs) within tumor region can generally reduce treatment effectiveness due to the cooling effect of blood flow. This paper aims to investigate the cooling effect of BVs on the temperature field of malignant tumor regions using a complex geometric model and numerical simulation. For deriving the model, the Navier-Stokes equation for blood flow is combined with Pennes bio-heat transfer equation for human tissue. The effects on treatment temperature caused by two different BV distributions inside a mammary tumor are analyzed through numerical simulation under different conditions of flow rate considering a Fe-Cr-Nb-B alloy, which has low Curie temperature ranging from 42 °C to 45 °C. Numerical results show that the multi-vessel system has more obvious cooling effects than the single vessel one on the temperature field distribution for hyperthermia. Besides, simulation results show that the temperature field within tumor area can also be influenced by the velocity and diameter of BVs. To minimize the cooling effect, this article proposes a treatment method based on the increase of the thermal energy provided to MNPs associated with the adoption of low Curie temperature particles recently reported in literature. Results demonstrate that this approach noticeably improves the uniformity of the temperature field, and shortens the treatment time in a Fe-Cr-Nb-B system, thus reducing the side effects to the patient.

  7. Inherent Safety Features and Passive Prevention Approaches for Pb/Bi-cooled Accelerator-Driven Systems

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Johan

    2003-03-01

    This thesis is devoted to the investigation of passive safety and inherent features of subcritical nuclear transmutation systems - accelerator-driven systems. The general objective of this research has been to improve the safety performance and avoid elevated coolant temperatures in worst-case scenarios like unprotected loss-of-flow accidents, loss-of-heat-sink accidents, and a combination of both these accident initiators. The specific topics covered are emergency decay heat removal by reactor vessel auxiliary cooling systems, beam shut-off by a melt-rupture disc, safety aspects from locating heat-exchangers in the riser of a pool-type reactor system, and reduction of pressure resistance in the primary circuit by employing bypass routes. The initial part of the research was focused on reactor vessel auxiliary cooling systems. It was shown that an 80 MW{sub th} Pb/Bi-cooled accelerator-driven system of 8 m height and 6 m diameter vessel can be well cooled in the case of loss-of-flow accidents in which the accelerator proton beam is not switched off. After a loss-of-heat-sink accident the proton beam has to be interrupted within 40 minutes in order to avoid fast creep of the vessel. If a melt-rupture disc is included in the wall of the beam pipe, which breaks at 150 K above the normal core outlet temperature, the grace period until the beam has to be shut off is increased to 6 hours. For the same vessel geometry, but an operating power of 250 MW{sub th} the structural materials can still avoid fast creep in case the proton beam is shut off immediately. If beam shut-off is delayed, additional cooling methods are needed to increase the heat removal. Investigations were made on the filling of the gap between the guard and the reactor vessel with liquid metal coolant and using water spray cooling on the guard vessel surface. The second part of the thesis presents examinations regarding an accelerator-driven system also cooled with Pb/Bi but with heat-exchangers located

  8. Inherent Safety Features and Passive Prevention Approaches for Pb/Bi-cooled Accelerator-Driven Systems

    International Nuclear Information System (INIS)

    Carlsson, Johan

    2003-03-01

    This thesis is devoted to the investigation of passive safety and inherent features of subcritical nuclear transmutation systems - accelerator-driven systems. The general objective of this research has been to improve the safety performance and avoid elevated coolant temperatures in worst-case scenarios like unprotected loss-of-flow accidents, loss-of-heat-sink accidents, and a combination of both these accident initiators. The specific topics covered are emergency decay heat removal by reactor vessel auxiliary cooling systems, beam shut-off by a melt-rupture disc, safety aspects from locating heat-exchangers in the riser of a pool-type reactor system, and reduction of pressure resistance in the primary circuit by employing bypass routes. The initial part of the research was focused on reactor vessel auxiliary cooling systems. It was shown that an 80 MW th Pb/Bi-cooled accelerator-driven system of 8 m height and 6 m diameter vessel can be well cooled in the case of loss-of-flow accidents in which the accelerator proton beam is not switched off. After a loss-of-heat-sink accident the proton beam has to be interrupted within 40 minutes in order to avoid fast creep of the vessel. If a melt-rupture disc is included in the wall of the beam pipe, which breaks at 150 K above the normal core outlet temperature, the grace period until the beam has to be shut off is increased to 6 hours. For the same vessel geometry, but an operating power of 250 MW th the structural materials can still avoid fast creep in case the proton beam is shut off immediately. If beam shut-off is delayed, additional cooling methods are needed to increase the heat removal. Investigations were made on the filling of the gap between the guard and the reactor vessel with liquid metal coolant and using water spray cooling on the guard vessel surface. The second part of the thesis presents examinations regarding an accelerator-driven system also cooled with Pb/Bi but with heat-exchangers located in the

  9. Investigation of light gas effects on passive containment cooling system in ALWR

    International Nuclear Information System (INIS)

    Paladino, D.; Auban, O.; Huggenberger, M.; Andreani, M.

    2003-01-01

    The large-scale thermal-hydraulic PANDA facility has been used for the last years for investigating passive decay-heat removal systems and related containment phenomena relevant for current and next generation of light water reactors. Passive Containment Cooling System (PCCS) systems operate by transferring heat from the containment to a water pool located outside the containment by steam condensation, and serve to mitigate long-term pressure build-up in the event of steam discharge from the primary circuit. As part of the 5 th Euratom framework program project TEMPEST, a new series of tests was performed in the PANDA facility to experimentally investigate the distribution of non-condensable gases inside the containment and their effect on the performance of PCCS of the European Simplified Boiling Water Reactor (ESBWR). The influence of light gas(hydrogen) on the PCCs performance is of special interest. Hydrogen release caused by the metalwater reaction in case of severe accident was simulated in PANDA by injecting helium into the lines feeding the break flow from the reactor pressure vessel to the Drywells. The paper combines the presentation of experimental results for a number of PANDA tests and the analysis performed using the GOTHIC code. As GOTHIC has 3-D modeling capabilities, gas distribution effects could be studied. The comparison of GOTHIC calculations (two pre-test and one post-test with the same model) with selected TEMPEST tests showed that the code is capable to predict well gas stratification in the drywell, while the system pressure increase due to the release of light gas is slightly overestimated. The analysis aiming to clarify the discordance between the GOTHIC simulation and the experimental results is included in this paper

  10. Cooling methods of station blackout scenario for LWR plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The objective of this study is to analyze the cooling method of station blackout scenario for both the BWR and PWR plants by RELAP5 code and to check the validity of the cooling method proposed by the utilities. In the BWR plant cooling scenario, the Reactor Core Isolation Cooling System (RCIC), which is operated with high pressure steam from the reactor, injects cooling water into the reactor to keep the core water level. The steam generated in the core is released into the suppression pool at containment vessel to condense. To restrict the containment vessel pressure rising, the ventilation from the wet-well is operated. The scenario is analyzed by RELAP5 code. In the PWR plant scenario, the primary pressure is decreased by the turbine-driven auxiliary feed water system operated with secondary side steam of the steam generators (SGs). And the core cooling is kept by the natural circulation flow at the primary loop. From the RELAP5 code analysis, it was shown that the primary system cooling was practicable by using the turbine-driven auxiliary feed water system. (author)

  11. An efficient continuous flow helium cooling unit for Moessbauer experiments

    International Nuclear Information System (INIS)

    Herbert, I.R.; Campbell, S.J.

    1976-01-01

    A Moessbauer continuous flow cooling unit for use with liquid helium over the temperature range 4.2 to 300K is described. The cooling unit can be used for either absorber or source studies in the horizontal plane and it is positioned directly on top of a helium storage vessel. The helium transfer line forms an integral part of the cooling unit and feeds directly into the storage vessel so that helium losses are kept to the minimum. The helium consumption is 0.12 l h -1 at 4.2 K decreasing to 0.055 l h -1 at 40 K. The unit is top loading and the exchange gas cooled samples can be changed easily and quickly. (author)

  12. Cleaning device for recycling pump motor cooling system in nuclear reactor

    International Nuclear Information System (INIS)

    Katayama, Kenjiro; Kondo, Takahisa; Shindo, Kenjiro; Akimoto, Jun.

    1996-01-01

    The cleaning device of the present invention comprises a cleaning water supply pump, a filter for filtering the cleaning water and a cap member for isolating the inside of a motor casing from the inside of a reactor pressure vessel. A motor in the motor casing and a pump in the reactor pressure vessel are removed, the cap member is attached to the upper end of the motor casing to isolate the inside of the motor casing from the inside of the reactor pressure vessel. If the cleaning water supply pump is operated in this state, the cleaning water flows from a returning pipeline for cooling water circulation, connected to the motor casing to supply pipelines through a heat exchange and is discharged. The discharged water passes through a filter and is sent again, as the cleaning water, to the cleaning water supply pump. With such procedures, the recycling pump motor cooling system in the BWR type reactor can be cleaned without disposing a cyclone separator and irrespective of presence or absence of reactor coolants in the reactor pressure vessel. (I.N.)

  13. Review of the Technical Status on the Debris Bed Cooling Model

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

    2007-09-15

    Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris0.

  14. Review of the Technical Status on the Debris Bed Cooling Model

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

    2007-09-01

    Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris

  15. A study on passive containment cooling condensers in SBWR

    International Nuclear Information System (INIS)

    Kuran, S.; Soekmen; C. N.

    2001-01-01

    The passive containment cooling condensers (PCCC) are the crucial part of several new reactor designs, like European Simplified Boiling Water Reactor (ESBWR) and the SBWR. In a hypothetical accident, the pressurised steam non-condensable mixture from drywell is condensed in PCCCs, and condensate is returned to reactor vessel while non-condensable is vented through wet well. In this study, in order to examine the performance of PCCCs, condensation with presence of noncondensable is investigated. Condensation with different noncondensable types and conditions is studied on a PCCC model, which is developed by using RELAP5 Mod3.2 computer code

  16. An Individualized, Perception-Based Protocol to Investigate Human Physiological Responses to Cooling

    Science.gov (United States)

    Coolbaugh, Crystal L.; Bush, Emily C.; Galenti, Elizabeth S.; Welch, E. Brian; Towse, Theodore F.

    2018-01-01

    Cold exposure, a known stimulant of the thermogenic effects of brown adipose tissue (BAT), is the most widely used method to study BAT physiology in adult humans. Recently, individualized cooling has been recommended to standardize the physiological cold stress applied across participants, but critical experimental details remain unclear. The purpose of this work was to develop a detailed methodology for an individualized, perception-based protocol to investigate human physiological responses to cooling. Participants were wrapped in two water-circulating blankets and fitted with skin temperature probes to estimate BAT activity and peripheral vasoconstriction. We created a thermoesthesia graphical user interface (tGUI) to continuously record the subject's perception of cooling and shivering status during the cooling protocol. The protocol began with a 15 min thermoneutral phase followed by a series of 10 min cooling phases and concluded when sustained shivering (>1 min duration) occurred. Researchers used perception of cooling feedback (tGUI ratings) to manually adjust and personalize the water temperature at each cooling phase. Blanket water temperatures were recorded continuously during the protocol. Twelve volunteers (ages: 26.2 ± 1.4 years; 25% female) completed a feasibility study to evaluate the proposed protocol. Water temperature, perception of cooling, and shivering varied considerably across participants in response to cooling. Mean clavicle skin temperature, a surrogate measure of BAT activity, decreased (−0.99°C, 95% CI: −1.7 to −0.25°C, P = 0.16) after the cooling protocol, but an increase in supraclavicular skin temperature was observed in 4 participants. A strong positive correlation was also found between thermoesthesia and peripheral vasoconstriction (ρ = 0.84, P < 0.001). The proposed individualized, perception-based protocol therefore has potential to investigate the physiological responses to cold stress applied across populations with

  17. 46 CFR 168.15-45 - Heating and cooling.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Heating and cooling. 168.15-45 Section 168.15-45 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) NAUTICAL SCHOOLS CIVILIAN NAUTICAL SCHOOL VESSELS Accommodations § 168.15-45 Heating and cooling. All quarters must be adequately heated and cooled...

  18. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  19. Integral experiments on in-vessel coolability and vessel creep: results and analysis of the FOREVER-C1 test

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A. [Division of Nuclear Power Safety, Royal Institute of Technology, Drottning Kristinas Vaeg., Stockholm (Sweden)

    1999-07-01

    This paper describes the FOREVER (Failure Of REactor VEssel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The paper presents the experimental results and analysis of the first FOREVER-C1 test. During this experiment, the 1/10th scale pressure vessel, heated to about 900degC and pressurized to 26 bars, was subjected to creep deformation in a non-stop 24-hours test. The vessel wall displacement data clearly shows different stages of the vessel deformation due to thermal expansion, elastic, plastic and creep processes. The maximum displacement was observed at the lowermost region of the vessel lower plenum. Information on the FOREVER-C1 measured thermal characteristics and analysis of the observed thermal and structural behavior is presented. The coupled nature of thermal and mechanical processes, as well as the effect of other system conditions (such as depressurization) on the melt pool and vessel temperature responses are analyzed. (author)

  20. Cooling device for reactor container

    International Nuclear Information System (INIS)

    Akiba, Miyuki.

    1996-01-01

    In a cooling device for a reactor container, a low pressure vessel is connected to an incondensible gas vent tube by way of an opening/closing valve. Upon occurrence of a loss of coolant accident, among steams and incondensible gases contained in the reactor container, steams are cooled and condensed in a heat exchanger. The incondensible gases are at first discharged from the heat exchanger to a suppression pool by way of the incondensible gas vent tube, but subsequently, they are stagnated in the incondensible gas vent tube to hinder heat exchanging and steam cooling and condensing effects in the heat exchanger thereby raising temperature and pressure in the reactor. However, if the opening/closing valve is opened when the incondensible gases are stagnated in the incondensible gas vent tube, since the incondensible gases stagnated in the heat exchanger are sucked and discharged to the low pressure vessel, the performance of the heat exchanger is maintained satisfactorily thereby enabling to suppress elevation of temperature and pressure in the reactor container. (N.H.)

  1. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  2. Analyses for passive safety of fusion reactor during ex-vessel loss of coolant accident

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Maki, Koichi; Uda, Tatuhiko; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki.

    1995-01-01

    Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident. (author)

  3. Preliminary structural evaluations of the STAR-LM reactor vessel and the support design

    International Nuclear Information System (INIS)

    Koo, Gyeong-Hoi; Sienicki, James J.; Moisseytsev, Anton

    2007-01-01

    In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor - Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid-structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules

  4. Cooling methods of station blackout scenario for LWR plants

    International Nuclear Information System (INIS)

    2012-01-01

    The objective of this study is to analyze the cooling method of station blackout scenario for both the BWR and PWR plants by RELAP5 code and to check the validity of the cooling method proposed by the utilities. In the BWR plant cooling scenario, the Reactor Core Isolation Cooling System (RCIC), which is operated with high pressure steam from the reactor, injects cooling water into the reactor to keep the core water level. The steam generated in the core is released into the suppression pool at containment vessel to condense. To restrict the containment vessel pressure rising, the ventilation from the wet-well is operated. The scenario is analyzed by RELAP5 and CONTEMPT-LT code. In the PWR plant scenario, the primary pressure is decreased by the turbine-driven auxiliary feed water system operated with secondary side steam of the steam generators (SGs). And the core cooling is kept by the natural circulation flow at the primary loop. The analytical method of un-uniform flow behavior among the SG U-tubes, which affects the natural circulation flow rate, is developed. (author)

  5. Progress on the MICE Liquid Absorber Cooling and Cryogenic Distribution System

    International Nuclear Information System (INIS)

    Green, M.A.; Baynham, E.; Bradshaw, T.; Drumm, P.; Ivanyushenkov, Y.; Ishimoto, S.; Cummings, M.A.C.; Lau, W.W.; Yang, S.Q.

    2005-01-01

    This report describes the progress made on the design of the cryogenic cooling system for the liquid absorber for the international Muon Ionization Cooling Experiment (MICE). The absorber consists of a 20.7-liter vessel that contains liquid hydrogen (1.48 kg at 20.3 K) or liquid helium (2.59 kg at 4.2 K). The liquid cryogen vessel is located within the warm bore of the focusing magnet for the MICE. The purpose of the magnet is to provide a low beam beta region within the absorber. For safety reasons, the vacuum vessel for the hydrogen absorber is separated from the vacuum vessel for the superconducting magnet and the vacuum that surrounds the RF cavities or the detector. The absorber thin windows separate the liquid in the absorber from the absorber vacuum. The absorber vacuum vessel also has thin windows that separate the absorber vacuum space from adjacent vacuum spaces. Because the muon beam in MICE is of low intensity, there is no beam heating in the absorber. The absorber can use a single 4 K cooler to cool either liquid helium or liquid hydrogen within the absorber

  6. Prestressed concrete pressure vessels for nuclear reactors - 1973

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

  7. Design and issues of the ITER in-vessel components: ITER Joint central team and home teams

    International Nuclear Information System (INIS)

    Parker, R.R.

    1998-01-01

    This paper surveys the status of the design of the in-vessel components for ITER, in particular the major components, namely the vacuum vessel, blanket and first wall, and divertor, and the interface of selected ancillary systems such as those used for RF heating and current drive, and for diagnostics. The vacuum vessel is a double-walled structure constructed from two toroidal shells joined by ribs. The space between the skins is filled with shield plates directly cooled by water. The structural material is 316 LN IG (ITER grade). Toroidal supports joining the vessel midplane ports with the TF structure limit possible differential toroidal displacements, as might occur due to seismic or vertical displacement events (VDEs). A variety of load conditions corresponding to normal and off-normal loads have been considered and in all cases peak vessel stresses are within allowables. The blanket system consists of approximately 700 modules, each weighing ∝4 t. The integrated first wall consists of a beryllium-tiled copper mat bonded to the water-cooled SS shield block. The copper mat functions as a heat sink and has imbedded in it an array of SS tubes providing water cooling. The modules are mechanically attached to a toroidal backplate. Loads due to centered disruptions are reacted via hoop stress in the backplate, whereas net vertical and horizontal loads such as those arising from VDEs are transferred through the backplate and divertor supports to the vessel. (orig.)

  8. Vessel classification method based on vessel behavior in the port of Rotterdam

    NARCIS (Netherlands)

    Zhou, Y.; Daamen, W.; Vellinga, T.; Hoogendoorn, S.P.

    2015-01-01

    AIS (Automatic Identification System) data have proven to be a valuable source to investigate vessel behavior. The analysis of AIS data provides a possibility to recognize vessel behavior patterns in a waterway area. Furthermore, AIS data can be used to classify vessel behavior into several

  9. Modeling of melt retention in EU-APR1400 ex-vessel core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Granovsky, V. S.; Sulatsky, A. A.; Khabensky, V. B.; Sulatskaya, M. B. [Alexandrov Research Inst. of Technology NITI, Sosnovy Bor (Russian Federation); Gusarov, V. V.; Almyashev, V. I.; Komlev, A. A. [Saint Petersburg State Technological Univ. SPbSTU, St.Petersburg (Russian Federation); Bechta, S. [KTH, Stockholm (Sweden); Kim, Y. S. [KHNP, 1312 Gil 70, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Park, R. J.; Kim, H. Y.; Song, J. H. [KAERI, 989 Gil 111, Daedeokdaero, Yuseong-gu, Daejeon (Korea, Republic of)

    2012-07-01

    A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled. (authors)

  10. Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

    1994-01-01

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ''flooded cavity'', is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

  11. Large-Scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

    1994-01-01

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the open-quotes flooded cavityclose quotes, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

  12. Assessment of reactor vessel integrity (ARVI)

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden)]. E-mail: sehgal@ne.kth.se; Karbojian, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Giri, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Kymaelaeinen, O. [FortumEngNP (Finland); Bonnet, J.M. [CEA (France); Ikkonen, K. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Sairanen, R. [VTT (Finland); Bhandari, S. [FRAMATOME (France); Buerger, M. [USTUTT (Germany); Dienstbier, J. [NRI Rez (Czech Republic); Techy, Z. [VEIKI (Hungary); Theofanous, T. [UCSB (United States)

    2005-02-01

    The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.

  13. Numerical investigation of mist/air impingement cooling on ribbed blade leading-edge surface.

    Science.gov (United States)

    Bian, Qingfei; Wang, Jin; Chen, Yi-Tung; Wang, Qiuwang; Zeng, Min

    2017-12-01

    The working gas turbine blades are exposed to the environment of high temperature, especially in the leading-edge region. The mist/air two-phase impingement cooling has been adopted to enhance the heat transfer on blade surfaces and investigate the leading-edge cooling effectiveness. An Euler-Lagrange particle tracking method is used to simulate the two-phase impingement cooling on the blade leading-edge. The mesh dependency test has been carried out and the numerical method is validated based on the available experimental data of mist/air cooling with jet impingement on a concave surface. The cooling effectiveness on three target surfaces is investigated, including the smooth and the ribbed surface with convex/concave columnar ribs. The results show that the cooling effectiveness of the mist/air two-phase flow is better than that of the single-phase flow. When the ribbed surfaces are used, the heat transfer enhancement is significant, the surface cooling effectiveness becomes higher and the convex ribbed surface presents a better performance. With the enhancement of the surface heat transfer, the pressure drop in the impingement zone increases, but the incremental factor of the flow friction is smaller than that of the heat transfer enhancement. Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. In-vessel tritium

    International Nuclear Information System (INIS)

    Ueda, Yoshio; Ohya, Kaoru; Ashikawa, Naoko; Ito, Atsushi M.; Kato, Daiji; Kawamura, Gakushi; Takayama, Arimichi; Tomita, Yukihiro; Nakamura, Hiroaki; Ono, Tadayoshi; Kawashima, Hisato; Shimizu, Katsuhiro; Takizuka, Tomonori; Nakano, Tomohide; Nakamura, Makoto; Hoshino, Kazuo; Kenmotsu, Takahiro; Wada, Motoi; Saito, Seiki; Takagi, Ikuji; Tanaka, Yasunori; Tanabe, Tetsuo; Yoshida, Masafumi; Toma, Mitsunori; Hatayama, Akiyoshi; Homma, Yuki; Tolstikhina, Inga Yu.

    2012-01-01

    The in-vessel tritium research is closely related to the plasma-materials interaction. It deals with the edge-plasma-wall interaction, the wall erosion, transport and re-deposition of neutral particles and the effect of neutral particles on the fuel recycling. Since the in-vessel tritium shows a complex nonlinear behavior, there remain many unsolved problems. So far, behaviors of in-vessel tritium have been investigated by two groups A01 and A02. The A01 group performed experiments on accumulation and recovery of tritium in thermonuclear fusion reactors and the A02 group studied theory and simulation on the in-vessel tritium behavior. In the present article, outcomes of the research are reviewed. (author)

  15. Guide to the periodic inspection of nuclear reactor steel pressure vessels

    International Nuclear Information System (INIS)

    1969-01-01

    This Guide is intended to provide general information and guidance to reactor owners or operators, inspection authorities, certifying authorities or regulatory bodies who are responsible for establishing inspection procedures for specific reactors or reactor types, and for the preparation of national codes or standards. The recommendations of the Guide apply primarily to water-cooled steel reactor vessels which are at a sufficiently early stage of design so that recommendations to provide accessibility for inspection can be incorporated into the early stages of design and inspection planning. However, much of the contents of the Guide are also applicable in part to vessels for other reactor types, such as gas-cooled, pressure-tube, or liquid-metal-cooled reactors, and also to some existing water-cooled reactors and reactors which are in advanced stage of design or construction. 46 refs, figs, 1 tab

  16. Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

    1996-01-01

    Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

  17. System design study of small lead-bismuth cooled reactor

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Hori, Toru; Konomura, Mamoru

    2003-07-01

    In phase II of the feasibility study of JNC, we will make a concept of a dispersion power source reactor with various requirements, such as economical competitiveness and safety. In the study of a small lead-bismuth cooled reactor, a concept whose features are long life core, inherent safety, natural convection of cooling system and steam generators in the reactor vessel has been designed since 2000. The investigations which have been done in 2002 are shown as follows; Safety analysis of UTOP considering uncertainty of reactivity. Possibility of reduction of number of control rods. Estimation of construction cost. Transient analyses of UTOP have been done in considering uncertainty of reactivity in order to show the inherent safety in the probabilistic method. And the inherent safety in UTOP is realized under the condition of considering uncertainty. Transient analyses of UTOP with various numbers of control rods have been done and it is suggested that there is possibility of reduction of the number of control rods considering accident managements. The method of cost estimation is a little modified. The cost of reactor vessel is estimated from that of medium sized lead-bismuth cooled reactor and the estimation of a purity control system is by coolant volume flow rate. The construction cost is estimated 850,000yen/kWe. (author)

  18. Modeling for evaluation of debris coolability in lower plenum of reactor pressure vessel

    International Nuclear Information System (INIS)

    Maruyama, Yu; Moriyama, Kiyofumi; Nakamura, Hideo; Hirano, Masashi

    2003-01-01

    Effectiveness of debris cooling by water that fills a gap between the debris and the lower head wall was estimated through steady calculations in reactor scale. In those calculations, the maximum coolable debris depth was assessed as a function of gap width with combination of correlations for critical heat flux and turbulent natural convection of a volumetrically heated pool. The results indicated that the gap with a width of 1 to 2 mm was capable of cooling the debris under the conditions of the TMI-2 accident, and that a significantly larger gap width was needed to retain a larger amount of debris within the lower plenum. Transient models on gap growth and water penetration into the gap were developed and incorporated into CAMP code along with turbulent natural convection model developed by Yin, Nagano and Tsuji, categorized in low Reynolds number type two-equation model. The validation of the turbulent model was made with the UCLA experiment on natural convection of a volumetrically heated pool. It was confirmed that CAMP code predicted well the distribution of local heat transfer coefficients along the vessel inner surface. The gap cooling model was validated by analyzing the in-vessel debris coolability experiments at JAERI, where molten Al 2 O 3 was poured into a water-filled hemispherical vessel. The temperature history measured on the vessel outer surface was satisfactorily reproduced by CAMP code. (author)

  19. Molten material-containing vessel

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko

    1998-01-01

    The molten material-containing vessel of the present invention comprises a vessel main body having an entrance opened at the upper end, a lid for closing the entrance, an outer tube having an upper end disposed at the lower surface of the lid, extended downwardly and having an closed lower end and an inner tube disposed coaxially with the outer tube. When a molten material is charged from the entrance to the inside of the vessel main body of the molten material-containing vessel and the entrance is closed by the lid, the outer tube and the inner tube are buried in the molten material in the vessel main body, accordingly, a fluid having its temperature elevated by absorption of the heat of the molten material rises along the inner circumferential surface of the outer tube, abuts against the lower surface of the lid and cooled by exchanging heat with the lid and forms a circulating flow. Since the heat in the molten material is continuously absorbed by the fluid, transferred to the lid and released from the lid to the atmospheric air, heat releasing efficiency can be improved compared with conventional cases. (N.H.)

  20. Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor

    International Nuclear Information System (INIS)

    Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.

    2006-10-01

    Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)

  1. TMI-2 Vessel Investigation Project (VIP) Metallurgical Program

    International Nuclear Information System (INIS)

    Diercks, D.R.; Neimark, L.A.

    1991-01-01

    The Three Mile Island Unite 2 (TMI-2) Vessel Investigation Project Metallurgical Program is a part of the international TMI-2 Vessel Investigation Project being conducted jointly by the U.S. Nuclear Regulatory Commission and the Organization for Economic Cooperation and Development. The objectives of the metallurgical program are to deduce the temperatures of, determine the mechanical properties of, and assess the integrity of the TMI-2 lower head during the loss-of-coolant accident. Fifteen samples have been removed from the lower head and are being examined. In addition, archive material from the lower head of the Midland nuclear reactor has been procured for conducting supplemental metallurgical evaluations and mechanical property determinations. Evaluations of the microstructure and mechanical properties of the as-received archive material have been completed, and a series of heat treatment experiments has been conducted to develop standard microstructures to be compared with those present in the TMI-2 samples. Results have been obtained from examinations of two of the fifteen TMI-2 lower head samples. These results indicate that one of these two samples, which contained cracks in the weld cladding extending ∼3 mm into the underlying base metal, apparently reached temperatures on the order of 1000 to 1100C during the accident. A preliminary examination of the core debris deposited on this sample has been performed. The other sample, from an area away from the region of core relocation, did not exceed 727C during the accident

  2. Investigation into a major crack that occurred during fabrication of a thick walled alloy pressure vessel

    International Nuclear Information System (INIS)

    Griffiths, Roger R.

    2002-01-01

    A high pressure thick walled (171 mm+cladding) reactor was under construction when a crack, with a total length of about 2.5 m, occurred at a nozzle. An investigation was conducted to determine how manufacture could safely proceed. This revealed that the primary cause of cracking was the method by which preheat had been applied to the vessel for the welding operation, coupled with the very low impact values achieved by the weld metal in the as-welded condition. Investigation also centred on the use of dehydrogenation heat treatment (DHT) instead of an intermediate stress relief (ISR), and the oxidised nature of the fracture surface. The oxidation could not be satisfactorily explained, and as a result neither the time the fracture occurred nor the significance of applying DHT in place of ISR could be absolutely determined. Nevertheless it was concluded that fracture probably occurred before DHT was applied. It was recommended that the method of preheat be revised and ISR applied without cooling below minimum preheat temperature. Further review of the incident resulted in additional recommendations for prevention of a recurrence in future work. One critical aspect was the lack of response to the poor as-welded toughness properties of the weld deposit

  3. Investigation into a major crack that occurred during fabrication of a thick walled alloy pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, Roger R

    2002-08-01

    A high pressure thick walled (171 mm+cladding) reactor was under construction when a crack, with a total length of about 2.5 m, occurred at a nozzle. An investigation was conducted to determine how manufacture could safely proceed. This revealed that the primary cause of cracking was the method by which preheat had been applied to the vessel for the welding operation, coupled with the very low impact values achieved by the weld metal in the as-welded condition. Investigation also centred on the use of dehydrogenation heat treatment (DHT) instead of an intermediate stress relief (ISR), and the oxidised nature of the fracture surface. The oxidation could not be satisfactorily explained, and as a result neither the time the fracture occurred nor the significance of applying DHT in place of ISR could be absolutely determined. Nevertheless it was concluded that fracture probably occurred before DHT was applied. It was recommended that the method of preheat be revised and ISR applied without cooling below minimum preheat temperature. Further review of the incident resulted in additional recommendations for prevention of a recurrence in future work. One critical aspect was the lack of response to the poor as-welded toughness properties of the weld deposit.

  4. Investigation on flow stability of supercritical water cooled systems

    International Nuclear Information System (INIS)

    Cheng, X.; Kuang, B.

    2006-01-01

    Research activities are ongoing worldwide to develop nuclear power plants with supercritical water cooled reactor (SCWR) with the purpose to achieve a high thermal efficiency and to improve their economical competitiveness. However, the strong variation of the thermal-physical properties of water in the vicinity of the pseudo-critical line results in challenging tasks in various fields, e.g. thermal-hydraulic design of a SCWR. One of the challenging tasks is to understand and to predict the dynamic behavior of supercritical water cooled systems. Although many thermal-hydraulic research activities were carried out worldwide in the past as well as in the near present, studies on dynamic behavior and flow stability of SC water cooled systems are scare. Due to the strong density variation, flow stability is expected to be one of the key items which need to be taken into account in the design of a SCWR. In the present work, the dynamic behavior and flow stability of SC water cooled systems are investigated using both numerical and theoretical approaches. For this purpose a new computer code SASC was developed, which can be applied to analysis the dynamic behavior of systems cooled by supercritical fluids. In addition, based on the assumptions of a simplified system, a theoretical model was derived for the prediction of the onset of flow instability. A comparison was made between the results obtained using the theoretical model and those from the SASC code. A good agreement was achieved. This gives the first evidence of the reliability of both the SASC code and the theoretical model

  5. Toward Cooling Uniformity: Investigation of Spiral, Sweeping Holes, and Unconventional Cooling Paradigms

    Science.gov (United States)

    Shyam, Vikram; Thurman, Douglas R.; Poinsatte, Philip E.; Ameri, Ali A.; Culley, Dennis E.

    2018-01-01

    Surface infrared thermography, hotwire anemometry, and thermocouple surveys were performed on two new film cooling hole geometries: spiral/rifled holes and fluidic sweeping holes. Ways to quantify the efficacy of novel cooling holes that are asymmetric, not uniformly spaced or that show variation from hole to hole are presented. The spiral holes attempt to induce large-scale vorticity to the film cooling jet as it exits the hole to prevent the formation of the kidney shaped vortices commonly associated with film cooling jets. The fluidic sweeping hole uses a passive in-hole geometry to induce jet sweeping at frequencies that scale with blowing ratios. The spiral hole performance is compared to that of round holes with and without compound angles. The fluidic hole is of the diffusion class of holes and is therefore compared to a 777 hole and square holes. A patent-pending spiral hole design showed the highest potential of the nondiffusion type hole configurations. Velocity contours and flow temperature were acquired at discreet cross-sections of the downstream flow field. The passive fluidic sweeping hole shows the most uniform cooling distribution but suffers from low span-averaged effectiveness levels due to enhanced mixing. The data was taken at a Reynolds number of 11,000 based on hole diameter and freestream velocity. Infrared thermography was taken for blowing ratios of 1.0, 1.5, 2.0, and 2.5 at a density ratio of 1.05. The flow inside the fluidic sweeping hole was studied using 3D unsteady RANS. A section on ideas for future work is included that addresses issues of quantifying cooling uniformity and provides some ideas for changing the way we think about cooling such as changing the direction of cooling or coupling acoustic devices to cooling holes to regulate frequency.

  6. Proposal of In-vessel corium retention concept for Paks NPP

    International Nuclear Information System (INIS)

    Elter, J.; Toth, E.; Matejovic, P.

    2011-01-01

    The in-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) seems to be a promising severe accident management strategy not only for new generation of advanced PWRs, but also for VVER-440/V213 reactors, which were designed several years ago. The basic idea of in-vessel retention of corium is to prevent RPV failure by flooding the reactor cavity so that the reactor pressure vessel is submerged in water up to its support structures, and thus the decay heat can be transferred from the corium pool through the vessel wall and into the water surrounding the vessel. An IVR concept with simple ECVR loop based only on minor modifications of existing plant technology was proposed for the Paks Nuclear Power Plant. 2 severe accident (LB and SB LOCA) without availability of HP and LP safety injection in power upgrade (108%) conditions were simulated using the ASTEC code. The analyses show that the proposed solution is effective in preserving RPV integrity in the case of severe accident. Possible uncertainties in code predictions are covered by the applied conservative assumptions

  7. Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

    2017-04-01

    Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

  8. Fast-neutron nuclear reactor vessel

    International Nuclear Information System (INIS)

    Presciuttini, L.

    1984-01-01

    The reactor vessel comprises a cylindrical shell, of which axis is vertical, coupled at its lower part with a dished bottom. The reactor core rests on a support plate bearing on the bottom of the vessel. The above dished bottom comprises a spherical sector having the same radius and the same axis as the cylindrical shell and joining the lower part of the shell, and a spherical head of which radius is a little more important than the spherical sector one. A cylindrical support for the reactor core is attached to the vessel at the joint between the two dished sections. The invention applies more particularly to integrated type reactors cooled by liquid sodium [fr

  9. Thermal-hydraulic investigations on the CEA-ENEA DEMO relevant helium cooled poloidal blanket

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Polazzi, G.; Vallette, F.; Proust, E.; Eid, M.

    1994-01-01

    The CEA-ENEA design of an Helium Cooled Solid Breeder Blanket (HCSBB) for the DEMO reactor, with a breeder in tube (BIT) poloidal arrangement, is based on the use of lithium ceramic pellets, the ENEA γ-LiAlO 2 or the CEA Li 2 ZrO 3 . Due to the geometry of the DEMO reactor plasma chamber, these breeder bundles are adapted to the Vacuum Vessel with a strong poloidal curvature. This curvature influences the thermal-hydraulic behaviour of the coolant flowing inside the bundle. The paper presents the CEA-ENEA first results of the experimental and theoretical programme, aiming at optimizing the breeder module thermal hydraulic design. (author) 6 refs.; 7 figs.; 1 tab

  10. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    Park, Yong Chul; Ryu, Jeong Soo

    2007-12-01

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

  11. Glass Forming Ability of Amorphous Drugs Investigated by Continuous Cooling and Isothermal Transformation.

    Science.gov (United States)

    Blaabjerg, Lasse I; Lindenberg, Eleanor; Löbmann, Korbinian; Grohganz, Holger; Rades, Thomas

    2016-09-06

    The aim of this study was to investigate the glass forming ability of 12 different drugs by the determination of continuous cooling and isothermal transformation diagrams in order to elucidate if an inherent differentiation between the drugs with respect to their the glass forming ability can be made. Continuous-cooling-transformation (CCT) and time-temperature-transformation (TTT) diagrams of the drugs were developed in order to predict the critical cooling rate necessary to convert the drug from the melt into an amorphous form. While TTT diagrams overestimated the actual critical cooling rate, they allowed an inherent differentiation of glass forming ability for the investigated drugs into drugs that are extremely difficult to amorphize (>750 °C/min), drugs that require modest cooling rates (>10 °C/min), and drugs that can be made amorphous even at very slow cooling rates (>2 °C/min). Thus, the glass forming ability can be predicted by the use of TTT diagrams. In contrast to TTT diagrams, CCT diagrams may not be suitable for small organic molecules due to poor separation of exothermic events, which makes it difficult to determine the zone of recrystallization. In conclusion, this study shows that glass forming ability of drugs can be predicted by TTT diagrams.

  12. Numerical Study of Condensation Heat Exchanger Design in a Cooling jacket: Correlation Investigation

    International Nuclear Information System (INIS)

    Kim, Myoung Jun; Lee, Hee Joon; Kang, Han Ok; Lee, Tae Ho; Park, Cheon Tae

    2013-01-01

    In this study, condensing heat transfer correlation of TSCON is evaluated with the existing experimental data set to design condensation heat exchanger without noncondensable gas effect (pure steam condensation) in a cooling jacket. From the investigation of the existing condensation heat transfer correlation to the existing experimental data, the improved Shah's correlation showed most satisfactory result for the condensation heat transfer coefficient with experimental data of Khun in a cooling jacket, whereas the Shah's correlation with experimental data of Lee. Lee et al. reported the improved Shah correlation gave us the best predictor for the condensation heat transfer data of Kim and Henderson in a subcooled and saturated water pool. They suggested the improved Shah correlation should be adopted as condensation heat transfer module in TSCON(Thermal Sizing of CONdenser) to design condensation heat exchanger in secondary passive cooling system of nuclear plant

  13. Mini-Membrane Evaporator for Contingency Spacesuit Cooling

    Science.gov (United States)

    Makinen, Janice V.; Bue, Grant C.; Campbell, Colin; Petty, Brian; Craft, Jesse; Lynch, William; Wilkes, Robert; Vogel, Matthew

    2015-01-01

    The next-generation Advanced Extravehicular Mobility Unit (AEMU) Portable Life Support System (PLSS) is integrating a number of new technologies to improve reliability and functionality. One of these improvements is the development of the Auxiliary Cooling Loop (ACL) for contingency crewmember cooling. The ACL is a completely redundant, independent cooling system that consists of a small evaporative cooler--the Mini Membrane Evaporator (Mini-ME), independent pump, independent feedwater assembly and independent Liquid Cooling Garment (LCG). The Mini-ME utilizes the same hollow fiber technology featured in the full-sized AEMU PLSS cooling device, the Spacesuit Water Membrane Evaporator (SWME), but Mini-ME occupies only approximately 25% of the volume of SWME, thereby providing only the necessary crewmember cooling in a contingency situation. The ACL provides a number of benefits when compared with the current EMU PLSS contingency cooling technology, which relies upon a Secondary Oxygen Vessel; contingency crewmember cooling can be provided for a longer period of time, more contingency situations can be accounted for, no reliance on a Secondary Oxygen Vessel (SOV) for contingency cooling--thereby allowing a reduction in SOV size and pressure, and the ACL can be recharged-allowing the AEMU PLSS to be reused, even after a contingency event. The first iteration of Mini-ME was developed and tested in-house. Mini-ME is currently packaged in AEMU PLSS 2.0, where it is being tested in environments and situations that are representative of potential future Extravehicular Activities (EVA's). The second iteration of Mini-ME, known as Mini-ME2, is currently being developed to offer more heat rejection capability. The development of this contingency evaporative cooling system will contribute to a more robust and comprehensive AEMU PLSS.

  14. Drill core investigations from the TMI-2 pressure vessel. Final report

    International Nuclear Information System (INIS)

    Sturm, D.; Katerbau, K.H.; Maile, K.; Ruoff, H.

    1994-01-01

    For the evaluation of the results obtained in TMI-2 VIP and for the preparation of the continuing discussion in the OECD and of research measures in the national sphere but also for the appraisal of the effect of the results to date on safety philosophy and safety research in Germany, the present research project, inter alia, was commenced. In content was: a) Furtherance of the OECD-NEA-TMI-2 Vessel Investigation Project in dealing with the testing programme by active collaboration in the Programme Review Group, by participation in ad-hoc meetings on the question of specimen extraction, by advice on the conduct of metallographic, metallurgical and mechanical investigations on the specimens from the RPV bottom head and by assessment of the findings. b) Investigation of specimens from the bottom head of the TMI-2 reactor pressure vessel. c) Investigation of specimens from archive material. The investigations reach the widely agreed conclusion that during the accident a hot spot developed in the bottom head of the reactor in which for a time of about 30 minutes a maximum temperature of some 1100 C or greater than 900 C prevailed. Around this zone there is a region with temperatures higher than ca. 730 C (A 1 ) whilst the predominant portion of the head had not been heated beyond the 1 temperature. (orig.) [de

  15. Experimental investigation of biomimetic self-pumping and self-adaptive transpiration cooling.

    Science.gov (United States)

    Jiang, Pei-Xue; Huang, Gan; Zhu, Yinhai; Xu, Ruina; Liao, Zhiyuan; Lu, Taojie

    2017-09-01

    Transpiration cooling is an effective way to protect high heat flux walls. However, the pumps for the transpiration cooling system make the system more complex and increase the load, which is a huge challenge for practical applications. A biomimetic self-pumping transpiration cooling system was developed inspired by the process of trees transpiration that has no pumps. An experimental investigation showed that the water coolant automatically flowed from the water tank to the hot surface with a height difference of 80 mm without any pumps. A self-adaptive transpiration cooling system was then developed based on this mechanism. The system effectively cooled the hot surface with the surface temperature kept to about 373 K when the heating flame temperature was 1639 K and the heat flux was about 0.42 MW m -2 . The cooling efficiency reached 94.5%. The coolant mass flow rate adaptively increased with increasing flame heat flux from 0.24 MW m -2 to 0.42 MW m -2 while the cooled surface temperature stayed around 373 K. Schlieren pictures showed a protective steam layer on the hot surface which blocked the flame heat flux to the hot surface. The protective steam layer thickness also increased with increasing heat flux.

  16. Numerical Study of Condensation Heat Exchanger Design in a Cooling jacket: Correlation Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myoung Jun; Lee, Hee Joon [Kookmin Univ., Seoul (Korea, Republic of); Kang, Han Ok; Lee, Tae Ho; Park, Cheon Tae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, condensing heat transfer correlation of TSCON is evaluated with the existing experimental data set to design condensation heat exchanger without noncondensable gas effect (pure steam condensation) in a cooling jacket. From the investigation of the existing condensation heat transfer correlation to the existing experimental data, the improved Shah's correlation showed most satisfactory result for the condensation heat transfer coefficient with experimental data of Khun in a cooling jacket, whereas the Shah's correlation with experimental data of Lee. Lee et al. reported the improved Shah correlation gave us the best predictor for the condensation heat transfer data of Kim and Henderson in a subcooled and saturated water pool. They suggested the improved Shah correlation should be adopted as condensation heat transfer module in TSCON(Thermal Sizing of CONdenser) to design condensation heat exchanger in secondary passive cooling system of nuclear plant.

  17. Analysis of the steady state hydraulic behaviour of the ITER blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A.; Vallone, E. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • Nominal steady state hydraulic behaviour of ITER blanket standard sector cooling system has been investigated. • Numerical simulations have been run adopting a qualified thermal-hydraulic system code. • Hydraulic characteristic functions and coolant mass flow rates, velocities and pressure drops have been assessed. • Most of the considered circuits are able to effectively cool blanket modules, meeting ITER requirements. - Abstract: The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised.

  18. An Investigation on In-Vessel Composting of Pistachio Residuals with Different Additions

    OpenAIRE

    M Jalili; M Mokhtari; AA Ebrahimi; F Boghri

    2016-01-01

    Background and Objective: About 1.35×105 tons of pistachio waste are produced in annually Iran that can result in environmental problems if managed improperly. . The purpose of this study was to investigate in-vessel composting of pistachio residuals with addition of cow manure and dewatered sludge as a recycling alternative. Materials and Methods: Pistachios wastes were combined with weight ratio of 5.5:10 (dewatered sludge: pistachio waste) and weight ratio of 1:10 (Cow manure: pi...

  19. Experimental investigation of the performance characteristics of a counterflow wet cooling tower

    International Nuclear Information System (INIS)

    Lemouari, M.; Boumaza, M.

    2010-01-01

    An experimental investigation of the performance characteristics of a counter flow wet cooling tower represented by the heat rejected by the tower and its thermal effectiveness is presented in this paper. The tower is filled with a 'VGA.' (Vertical Grid Apparatus) type packing which is 0.42 m high and contains four (04) galvanized sheets having a zigzag form, between which are disposed three (03) metallic vertical grids in parallel with a cross-sectional test area of 0.15 m - 0.148 m. The investigation is concerned mainly on the effect of the air, water flow rates and the inlet water temperatures on the thermal effectiveness of the cooling tower as well as the heat rejected by this tower from water to be cooled to the air stream discharged into the atmosphere. The two operating regimes which were observed during the air/water contact inside the tower, a Pellicular Regime (PR) and a Bubble and Dispersion Regime (BDR) appear to be important, as The BDR regime enables to cool larger amount of water flow rates, while the Pellicular regime results with higher thermal effectiveness. (authors)

  20. Experimental investigation of the hydraulic characteristics of a counter flow wet cooling tower

    International Nuclear Information System (INIS)

    Lemouari, M.; Boumaza, M.; Kaabi, A.

    2011-01-01

    Thermal and nuclear electric power plants as well as several industrial processes invariably discharge considerable energy to their surrounding by heat transfer. Although water drawn from a nearby river or lake can be employed to carry away this energy, cooling towers offer an excellent alternative particularly in locations where sufficient cooling water cannot be easily obtained from natural sources or where concern for the environment imposes some limits on the temperature at which cooling water can be returned to the surrounding. This paper concerns an experimental investigation of the hydraulic characteristics of a counter flow wet cooling tower. The tower contains a 'VGA.' (Vertical Grid Apparatus) type packing which is 0.42 m high and consists of four (04) galvanised sheets having a zigzag form, between which are disposed three (03) metallic vertical grids in parallel with a cross sectional test area of 0.15 m x 0.148 m. The present investigation is focused mainly on the effect of the air and water flow rates on the hydraulic characteristics of the cooling tower, for different inlet water temperatures. The two hydrodynamic operating regimes which were observed during the air/water contact operation within the tower, namely the Pellicular Regime (PR) and the Bubble and Dispersion Regime (BDR) have enabled to distinguish two different states of pressure drop characteristics. The first regime is characterized by low pressure drop values, while in the second regime, the pressure drop values are relatively much higher than those observed in the first one. The dependence between the pressure drop characteristics and the combined heat and mass transport (air-water) through the packing inside the cooling tower is also highlighted. The obtained results indicate that this type of tower possesses relatively good hydraulic characteristics. This leads to the saving of energy. -- Highlights: → Cooling towers are widely used to reject waste heat from thermal and nuclear

  1. Corrosion of vessel steel during its interaction with molten corium

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Lopukh, D.B.; Gusarov, V.V.; Martinov, A.P.; Martinov, V.V.; Fieg, G.; Tromm, W.; Bottomley, D.; Tuomisto, H.

    2006-01-01

    This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheres during an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities and oxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium-specimen ingot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction

  2. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS's heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis

  3. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  4. Auxiliary cooling device for power plant

    International Nuclear Information System (INIS)

    Yamanoi, Kozo.

    1996-01-01

    An auxiliary cooling sea water pipeline for pumping up cooling sea water, an auxiliary cooling sea water pipeline and a primary side of an auxiliary cooling heat exchanger are connected between a sea water taking vessel and a sea water discharge pit. An auxiliary cooling water pump is connected to an auxiliary water cooling pipeline on the second side of the auxiliary cooling heat exchanger. The auxiliary cooling water pipeline is connected with each of auxiliary equipments of a reactor system and each of auxiliary equipments of the turbine system connected to a turbine auxiliary cooling water pipeline in parallel. During ordinary operation of the reactor, heat exchange for each of the auxiliary equipments of the reactor and heat exchange for each of the equipments of the turbine system are conducted simultaneously. Since most portions of the cooling devices of each of the auxiliary equipments of the reactor system and each of the auxiliary equipments of the turbine system can be used in common, the operation efficiency of the cooling device is improved. In addition, the space for the pipelines and the cost for the equipments can be reduced. (I.N.)

  5. Investigating the influence of photocatalytic cool wall adoption on meteorology and air quality in the Los Angeles basin

    Science.gov (United States)

    Zhang, J.; Tang, X.; Levinson, R.; Destaillats, H.; Mohegh, A.; Li, Y.; Tao, W.; Liu, J.; Ban-Weiss, G. A.

    2017-12-01

    Solar reflective "cool materials" can be used to lower urban temperatures, useful for mitigating the urban heat island effect and adapting to the local impacts of climate change. While numerous past studies have investigated the climate impacts of cool surfaces, few studies have investigated their effects on air pollution. Meteorological changes from increases in surface albedo can lead to temperature and transport induced modifications in air pollutant concentrations. In an effort to maintain high albedos in polluted environments, cool surfaces can also be made using photocatalytic "self-cleaning" materials. These photocatalytic materials can also remove NOx from ambient air, with possible consequences on ambient gas and particle phase pollutant concentrations. In this research, we investigate the impact of widespread deployment of cool walls on urban meteorology and air pollutant concentrations in the Los Angeles basin. Both photocatalytic and standard (not photocatalytic) high albedo wall materials are investigated. Simulations using a coupled meteorology-chemistry model (WRF-Chem) show that cool walls could effectively decrease urban temperatures in the Los Angeles basin. Preliminary results indicate that meteorology-induced changes from adopting standard cool walls could lead to ozone concentration reductions of up to 0.5 ppb. NOx removal induced by photocatalytic materials was modeled by modifying the WRF-Chem dry deposition scheme, with deposition rates informed by laboratory measurements of various commercially available materials. Simulation results indicate that increased deposition of NOx by photocatalytic materials could increase ozone concentrations, analogous to the ozone "weekend effect" in which reduced weekend NOx emissions can lead to increases in ozone. The impacts of cool walls on particulate matter concentrations are also discussed. Changes in particulate matter concentrations are found to be driven by albedo-induced changes in air pollutant

  6. Investigation of Horizontal Velocity Fields in Stirred Vessels with Helical Coils by PIV

    Directory of Open Access Journals (Sweden)

    Volker Bliem

    2014-01-01

    Full Text Available Horizontal velocity flow fields were measured by particle image velocimetry for a stirred vessel with baffles and two helical coils for enlargement of heat transfer area. The investigation was carried out in a cylindrical vessel with flat base and two different stirrers (radial-flow Rushton turbine and axial-flow propeller stirrer. Combined velocity plots for flow fields at different locations are presented. It was found that helical coils change the flow pattern significantly. Measurements for the axial-flow Rushton turbine showed a strong deflection by the coils, leading to a mainly tangential flow pattern. Behind baffles large areas of unused heat transfer area were found. First results for the axial-flow propeller reveal an extensive absence of fluid movement in the horizontal plane. Improved design considerations for enhanced heat transfer by more compatible equipment compilation are proposed.

  7. Unconventional liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.

    1989-06-01

    This report describes the rationale for, design of and analytical studies on an unconventional sodium-cooled power reactor, called the Trench Reactor. It derives its name from the long, narrow sodium pool in which the reactor is placed. Unconventional features include: pool shape; reactor shape (also long and narrow); reflector control; low power density; hot-leg primary pumping; absence of a cold sodium pool; large core boxes rather than a large number of subassemblies; large diameter metal fuel; vessel suspension from cables; and vessel cooling by natural circulation of building atmosphere (nitrogen) at all times. These features all seem feasible. They result in a system that is capable of at least a ten year reload interval and shows good safety through direct physical response to loss-of-heat-sink, loss-of-flow and limited-reactivity nuclear transients. 43 figs., 43 tabs

  8. ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2010-02-01

    A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

  9. Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 2. Comprehending the divertor structure

    International Nuclear Information System (INIS)

    Suzuki, Satoshi; Akiba, Masato; Saito, Masakatsu

    2006-01-01

    Divertor is given the largest heat load in the in-vessel components of fusion machine. The functions and conditions of divertor are stated from the point of view of thermal and structural dynamics. The way of thinking of structure design of divertor of JT-60 and the ITER (International Thermonuclear Experimental Reactor) is explained. As the conditions of divertor, the materials for large heat load, heat removal, pressure boundary, control of damage, and thermal stress/strain are considered. The divertor has to be changed periodically. The materials are required the heat removal function for high heat load. CuCrZr will be used to cooling tube and heat sink, and CFC materials for the surface. The cross section of ITER, a part of divertor, heat load of divertor and other components, the thermal conductivity of CFC and metal materials, conditions of cooling water for divertor of BWR, PWR and ITER, the thermal stress produced on rod, vertical target of ITER, structure of cooling tube, distribution of temperature and critical heart flux of inner wall of cooling tube, and fatigue clack of cooling tube are shown. (S.Y.)

  10. ITER vacuum vessel design and electromagnetic analysis on in-vessel components

    International Nuclear Information System (INIS)

    Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.; Iizuka, T.

    1995-01-01

    Major functional requirements for the vacuum vessel are to provide the first safety barrier and to support electromagnetic loads due to plasma disruptions and vertical displacement events, and to withstand plausible accidents without losing confinement. A double wall structure concept has been developed for the vacuum vessel due to its beneficial characteristics from the viewpoints of structural integrity and electrical continuity. An electromagnetic analysis of the blanket modules and the vacuum vessel has been performed to investigate force distributions on in-vessel components. According to the vertical displacement events (VDE) scenario, which assumes a critical q-value of 1.5, the total downward vertical force, induced by coupling between the eddy current and external fields, is about 110 MN. We have performed a stress analysis for the vacuum vessel using the VDE disruption forces acting on the blankets, and a maximum stress intensity of 112 MPa was obtained in the vicinity of the lower support of the vessel. (orig.)

  11. Challenges and innovative technologies on fuel handling systems for future sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

    2011-01-01

    The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short-cooled fuel and criteria to ensure safety during handling. In addition, the handling and elimination of residual sodium must be investigated; it implies specific cleaning treatment to prevent chemical risks such as corrosion or excess hydrogen production. The objective of this study is to identify the challenges of a SFR fuel handling system. It will then present the range of technical options incorporating innovative technologies under development to answer the GENERATION IV SFR requirements. (author)

  12. On a computational study for investigating acoustic streaming and heating during focused ultrasound ablation of liver tumor

    International Nuclear Information System (INIS)

    Solovchuk, Maxim A.; Sheu, Tony W.H.; Thiriet, Marc; Lin, Win-Li

    2013-01-01

    The influences of blood vessels and focused location on temperature distribution during high-intensity focused ultrasound (HIFU) ablation of liver tumors are studied numerically. A three-dimensional acoustics-thermal-fluid coupling model is employed to compute the temperature field in the hepatic cancerous region. The model construction is based on the linear Westervelt and bioheat equations as well as the nonlinear Navier–Stokes equations for the liver parenchyma and blood vessels. The effect of acoustic streaming is also taken into account in the present HIFU simulation study. Different blood vessel diameters and focal point locations were investigated. We found from this three-dimensional numerical study that in large blood vessels both the convective cooling and acoustic streaming can considerably change the temperature field and the thermal lesion near blood vessels. If the blood vessel is located within the beam width, both acoustic streaming and blood flow cooling effects should be addressed. The temperature rise on the blood vessel wall generated by a 1.0 MHz focused ultrasound transducer with the focal intensity 327 W/cm 2 was 54% lower when acoustic streaming effect was taken into account. Subject to the applied acoustic power the streaming velocity in a 3 mm blood vessel is 12 cm/s. Thirty percent of the necrosed volume can be reduced, when taking into account the acoustic streaming effect. -- Highlights: • 3D three-field coupling physical model for focused ultrasound tumor ablation is presented. • Acoustic streaming and blood flow cooling effects on ultrasound heating are investigated. • Acoustic streaming can considerably affect the temperature distribution. • The lesion can be reduced by 30% due to the acoustic streaming effect. • Temperature on the blood vessel wall is reduced by 54% due to the acoustic streaming effect

  13. Burst pressure investigation of filament wound type IV composite pressure vessel

    Science.gov (United States)

    Farhood, Naseer H.; Karuppanan, Saravanan; Ya, H. H.; Baharom, Mohamad Ariff

    2017-12-01

    Currently, composite pressure vessels (PVs) are employed in many industries such as aerospace, transportations, medical etc. Basically, the use of PVs in automotive application as a compressed natural gas (CNG) storage cylinder has been growing rapidly. Burst failure due to the laminate failure is the most critical failure mechanism for composite pressure vessels. It is predominantly caused by excessive internal pressure due to an overfilling or an overheating. In order to reduce fabrication difficulties and increase the structural efficiency, researches and studies are conducted continuously towards the proper selection of vessel design parameters. Hence, this paper is focused on the prediction of first ply failure pressure for such vessels utilizing finite element simulation based on Tsai-Wu and maximum stress failure criterions. The effects of laminate stacking sequence and orientation angle on the burst pressure were investigated in this work for a constant layered thickness PV. Two types of winding design, A [90°2/∓θ16/90°2] and B [90°2/∓θ]ns with different orientations of helical winding reinforcement were analyzed for carbon/epoxy composite material. It was found that laminate A sustained a maximum burst pressure of 55 MPa for a sequence of [90°2/∓15°16/90°2] while the laminate B returned a maximum burst pressure of 45 MPa corresponding to a stacking sequence of [90°2/±15°/90°2/±15°/90°2/±15° ....] up to 20 layers for a constant vessel thickness. For verification, a comparison was done with the literature under similar conditions of analysis and good agreement was achieved with a maximum difference of 4% and 10% for symmetrical and unsymmetrical layout, respectively.

  14. Vascular patterns in the heads of crocodilians: blood vessels and sites of thermal exchange.

    Science.gov (United States)

    Porter, William Ruger; Sedlmayr, Jayc C; Witmer, Lawrence M

    2016-12-01

    Extant crocodilians are a highly apomorphic archosaur clade that is ectothermic, yet often achieve large body sizes that can be subject to higher heat loads. Therefore, the anatomical and physiological roles that blood vessels play in crocodilian thermoregulation need further investigation to better understand how crocodilians establish and maintain cephalic temperatures and regulate neurosensory tissue temperatures during basking and normal activities. The cephalic vascular anatomy of extant crocodilians, particularly American alligator (Alligator mississippiensis) was investigated using a differential-contrast, dual-vascular injection technique and high resolution X-ray micro-computed tomography (μCT). Blood vessels were digitally isolated to create representations of vascular pathways. The specimens were then dissected to confirm CT results. Sites of thermal exchange, consisting of the oral, nasal, and orbital regions, were given special attention due to their role in evaporative cooling and cephalic thermoregulation in other diapsids. Blood vessels to and from sites of thermal exchange were studied to detect conserved vascular patterns and to assess their ability to deliver cooled blood to neurosensory tissues. Within the orbital region, both the arteries and veins demonstrated consistent branching patterns, with the supraorbital, infraorbital, and ophthalmotemporal vessels supplying and draining the orbit. The venous drainage of the orbital region showed connections to the dural sinuses via the orbital veins and cavernous sinus. The palatal region demonstrated a vast plexus that comprised both arteries and veins. The most direct route of venous drainage of the palatal plexus was through the palatomaxillary veins, essentially bypassing neurosensory tissues. Anastomotic connections with the nasal region, however, may provide an alternative route for palatal venous blood to reach neurosensory tissues. The nasal region in crocodilians is probably the most

  15. APFIM investigation of clustering in neutron-irradiated Fe-Cu alloys and pressure vessel steels

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Blavette, D.

    1996-01-01

    Pressure vessel steels used in PWRs are known to be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are commonly supposed to result from the formation of point defects, dislocation loops, voids and copper-rich precipitates. However, the real nature of the irradiation induced damage, in these particularly low copper steels (>0,1 wt%), has not been clearly identify yet. A new experimental work has been carried out thanks to atom probe and field ion microscopy (APFIM) facilities and, more particularly with a new generation of atom probe recently developed, namely the tomographic atom probe (TAP), in order to improve: the understanding of the complex behavior of copper precipitation which occurs when low-alloyed Fe-Cu model alloys are irradiated with neutrons; the microstructural characterization of the pressure vessel steel of the CHOOZ A reactor under various fluences (French Surveillance Programme). The investigations clearly reveal the precipitation of copper-rich clusters in irradiated Fe-Cu alloys while more complicated Si, Ni, Mn and Cu-solute 'clouds' were observed to develop in the low-copper ferritic solid solution of the pressure vessel steel. (authors)

  16. Americium behaviour in plastic vessels

    Energy Technology Data Exchange (ETDEWEB)

    Legarda, F.; Herranz, M. [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain); Idoeta, R., E-mail: raquel.idoeta@ehu.e [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain); Abelairas, A. [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain)

    2010-07-15

    The adsorption of {sup 241}Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of {sup 241}Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of {sup 241}Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification.

  17. Americium behaviour in plastic vessels

    International Nuclear Information System (INIS)

    Legarda, F.; Herranz, M.; Idoeta, R.; Abelairas, A.

    2010-01-01

    The adsorption of 241 Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of 241 Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of 241 Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification.

  18. Determinants of injuries in passenger vessel accidents.

    Science.gov (United States)

    Yip, Tsz Leung; Jin, Di; Talley, Wayne K

    2015-09-01

    This paper investigates determinants of crew and passenger injuries in passenger vessel accidents. Crew and passenger injury equations are estimated for ferry, ocean cruise, and river cruise vessel accidents, utilizing detailed data of individual vessel accidents that were investigated by the U.S. Coast Guard during the time period 2001-2008. The estimation results provide empirical evidence (for the first time in the literature) that crew injuries are determinants of passenger injuries in passenger vessel accidents. Copyright © 2015 Elsevier Ltd. All rights reserved.

  19. Numerical investigations on pressurized AL-composite vessel response to hypervelocity impacts: Comparison between experimental works and a numerical code

    Directory of Open Access Journals (Sweden)

    Mespoulet Jérôme

    2015-01-01

    Full Text Available Response of pressurized composite-Al vessels to hypervelocity impact of aluminum spheres have been numerically investigated to evaluate the influence of initial pressure on the vulnerability of these vessels. Investigated tanks are carbon-fiber overwrapped prestressed Al vessels. Explored internal air pressure ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from experiments (Xray radiographies, particle velocity measurement and post-mortem vessels have been compared to numerical results given from LS-DYNA ALE-Lagrange-SPH full coupling models. Simulations exhibit an under estimation in term of debris cloud evolution and shock wave propagation in pressurized air but main modes of damage/rupture on the vessels given by simulations are coherent with post-mortem recovered vessels from experiments. First results of this numerical work are promising and further simulation investigations with additional experimental data will be done to increase the reliability of the simulation model. The final aim of this crossed work is to numerically explore a wide range of impact conditions (impact angle, projectile weight, impact velocity, initial pressure that cannot be explore experimentally. Those whole results will define a rule of thumbs for the definition of a vulnerability analytical model for a given pressurized vessel.

  20. Experimental and numerical CHT-investigations of cooling structures formed by lost cores in cast housings for optimal heat transfer

    Science.gov (United States)

    Kohlstädt, S.; Vynnycky, M.; Gebauer-Teichmann, A.

    2018-05-01

    This paper investigates the cooling performance of six different lost core designs for automotive cast houses with regard to their cooling efficiency. For this purpose, the conjugate heat transfer (CHT) solver, chtMultiregion, of the freely available CFD-toolbox OpenFOAM in its implementation of version 2.3.1 is used. The turbulence contribution to the Navier-Stokes equations is accounted for by using the RANS Menter SST k - ω model. The results are validated for one of the geometries by comparing with experimental data. Of the six investigated cooling structures, the one that forces the fluid flow to change its direction the most produces the lowest temperatures on the surface of the cast housing. This good cooling performance comes at the price of the highest pressure loss in the cooling fluid and hence increased pump power. It is also found that the relationship between performance and pressure drop is by no means generally linear. Slight changes in the design can lead to a structure which cools almost as well, but at much decreased pressure loss. Regarding the absolute values, the simulations showed that the designed cooling structures are suitable for handling the cooling requirements in the particular applications and that the maximum temperature stays below the critical limits of the electronic components.

  1. Investigation of a weld defect, reactor vessel head Ringhals 2

    International Nuclear Information System (INIS)

    Embring, G.; Pers-Anderson, E.B.

    1994-01-01

    During the summer-outage 1993 Ringhals unit 2 vessel head was inspected at weld-area of Alloy 182. One major defect was found Two plus two ''boat-samples'' were taken out from the zone between the weld and the stainless cladding. All samples were sent to Studsviks laboratories for detailed investigations. The metallographic and fractographic investigations revealed that the major weld-defect had been there from manufacturing. The defect was located between the Alloy 182-buttering and the pressure vessel steel SA 533 grB cl 1. No indications of PWSCC or IDSCC were found. An inspection programme was defined. Different types of reference blocks were provided by Ringhals in cooperation with ABB TRC. Reference reflectors of type flat bottom hole (FBH) and eroded notches (EDM), with different sizes and separation were manufactured. One weld sample with manufacturing defects -lack of fusion and slag was inclusions- was present. ABB TRC performed UT inspection in the gap between the penetration and the thermal sleeve. Inspection results like defect identification, defect separation and defect sizing accuracy were compared with result from the destructive inspection. No relevant additional defects were found. An analysing and repair program was performed. A special designed disc sealed off the defect area. (authors). 5 figs., 3 refs

  2. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Dinh, T.N. [Royal Institute of Technology (Sweden)

    2007-04-15

    The report summarizes activities conducted at the Division of Nuclear Power Safety, Royal Institute of Technology-Sweden (KTH-NPS) within the ExCoolSe project during the year 2005, which is a transition year for the KTH-NPS program. The ExCoolSe project supported by NKS contributes to the severe accident research at KTH-NPS concurrently supported by APRI, HSK and EU SARNET. The main objective in ExCoolSe project is to scrutinize research on risk-significant safety issues related to severe accident management (SAM) strategy adopted for Nordic BWR plants, namely the Ex-vessel Coolability and Energetic Steam explosion. The work aims to pave way toward building a tangible research framework to tackle these long-standing safety issues. Chapter 1 describes the project objectives and work description. Chapter 2 provides a critical assessment of research results obtained from several past programs at KTH. This includes review of key data, insights and implications from POMECO (Porous Media Coolability) program, COMECO (Corium Melt Coolability) program, SIMECO (Study of In-Vessel Melt Coolability) program, and MISTEE (Micro-Interactions in Steam Explosion Experiments) program. Chapter 3 discusses the rationale of the new research program focusing on the SAM issue resolution. The program emphasizes identification and qualification of physics-based limiting mechanisms for both in-vessel phenomena (melt progression and debris coolability in the lower head, vessel failure), and ex-vessel phenomena. Chapter 4 introduces research results from the newly established DEFOR (Debris Formation) program and the ongoing MISTEE program. The focus of DEFOR is fulfill an apparent gap in the contemporary knowledge of severe accidents, namely mechanisms which govern the debris bed formation and bed characteristics. The later control the debris bed coolability. In the MISTEE program, methods for image synchronization and data processing were developed and tested, which enable processing of

  3. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors

    International Nuclear Information System (INIS)

    Park, H.S.; Dinh, T.N.

    2007-04-01

    The report summarizes activities conducted at the Division of Nuclear Power Safety, Royal Institute of Technology-Sweden (KTH-NPS) within the ExCoolSe project during the year 2005, which is a transition year for the KTH-NPS program. The ExCoolSe project supported by NKS contributes to the severe accident research at KTH-NPS concurrently supported by APRI, HSK and EU SARNET. The main objective in ExCoolSe project is to scrutinize research on risk-significant safety issues related to severe accident management (SAM) strategy adopted for Nordic BWR plants, namely the Ex-vessel Coolability and Energetic Steam explosion. The work aims to pave way toward building a tangible research framework to tackle these long-standing safety issues. Chapter 1 describes the project objectives and work description. Chapter 2 provides a critical assessment of research results obtained from several past programs at KTH. This includes review of key data, insights and implications from POMECO (Porous Media Coolability) program, COMECO (Corium Melt Coolability) program, SIMECO (Study of In-Vessel Melt Coolability) program, and MISTEE (Micro-Interactions in Steam Explosion Experiments) program. Chapter 3 discusses the rationale of the new research program focusing on the SAM issue resolution. The program emphasizes identification and qualification of physics-based limiting mechanisms for both in-vessel phenomena (melt progression and debris coolability in the lower head, vessel failure), and ex-vessel phenomena. Chapter 4 introduces research results from the newly established DEFOR (Debris Formation) program and the ongoing MISTEE program. The focus of DEFOR is fulfill an apparent gap in the contemporary knowledge of severe accidents, namely mechanisms which govern the debris bed formation and bed characteristics. The later control the debris bed coolability. In the MISTEE program, methods for image synchronization and data processing were developed and tested, which enable processing of

  4. Thermodynamic Alloy Design of High Strength and Toughness in 300 mm Thick Pressure Vessel Wall of 1.25Cr-0.5Mo Steel

    Directory of Open Access Journals (Sweden)

    Hye-sung Na

    2018-01-01

    Full Text Available In the 21st century, there is an increasing need for high-capacity, high-efficiency, and environmentally friendly power generation systems. The environmentally friendly integrated gasification combined-cycle (IGCC technology has received particular attention. IGCC pressure vessels require a high-temperature strength and creep strength exceeding those of existing pressure vessels because the operating temperature of the reactor is increased for improved capacity and efficiency. Therefore, high-pressure vessels with thicker walls than those in existing pressure vessels (≤200 mm must be designed. The primary focus of this research is the development of an IGCC pressure vessel with a fully bainitic structure in the middle portion of the 300 mm thick Cr-Mo steel walls. For this purpose, the effects of the alloy content and cooling rates on the ferrite precipitation and phase transformation behaviors were investigated using JMatPro modeling and thermodynamic calculation; the results were then optimized. Candidate alloys from the simulated results were tested experimentally.

  5. Simulation for temperature changing investigation at RSG-GAS cooling system

    International Nuclear Information System (INIS)

    Utaja

    2002-01-01

    The RSG-GAS cooling system considers of primary and secondary system, is used for heat rejection from reactor core to the atmosphere. For temperature changing investigation cause by atmospherics condition changing or coolant flow rate changing, is more safe done by simulation. This paper describes the simulation for determine the RSG-GAS coolant temperature changing base on heat exchange and cooling tower characteristic. The simulation is done by computer programme running under WINDOWS 95 or higher. The temperature changing is based on heat transfer process on heat exchanger and cooling tower. The simulation will show the water tank temperature changing caused by the temperature and humidity of the atmosphere or by coolant flow rate changing. For example the humidity changing from 60% to 80% atmospherics temperature 30 oC and 32400 k Watt power will change the tank temperature from 37,97 oC to 40,03 oC

  6. Experimental investigation of temperature rise in bone drilling with cooling: A comparison between modes of without cooling, internal gas cooling, and external liquid cooling.

    Science.gov (United States)

    Shakouri, Ehsan; Haghighi Hassanalideh, Hossein; Gholampour, Seifollah

    2018-01-01

    Bone fracture occurs due to accident, aging, and disease. For the treatment of bone fractures, it is essential that the bones are kept fixed in the right place. In complex fractures, internal fixation or external methods are used to fix the fracture position. In order to immobilize the fracture position and connect the holder equipment to it, bone drilling is required. During the drilling of the bone, the required forces to chip formation could cause an increase in the temperature. If the resulting temperature increases to 47 °C, it causes thermal necrosis of the bone. Thermal necrosis decreases bone strength in the hole and, subsequently, due to incomplete immobilization of bone, fracture repair is not performed correctly. In this study, attempts have been made to compare local temperature increases in different processes of bone drilling. This comparison has been done between drilling without cooling, drilling with gas cooling, and liquid cooling on bovine femur. Drilling tests with gas coolant using direct injection of CO 2 and N 2 gases were carried out by internal coolant drill bit. The results showed that with the use of gas coolant, the elevation of temperature has limited to 6 °C and the thermal necrosis is prevented. Maximum temperature rise reached in drilling without cooling was 56 °C, using gas and liquid coolant, a maximum temperature elevation of 43 °C and 42 °C have been obtained, respectively. This resulted in decreased possibility of thermal necrosis of bone in drilling with gas and liquid cooling. However, the results showed that the values obtained with the drilling method with direct gas cooling are independent of the rotational speed of drill.

  7. Investigation of heating and cooling in a stand-alone high temperature PEM fuel cell system

    International Nuclear Information System (INIS)

    Zhang, Caizhi; Yu, Tao; Yi, Jun; Liu, Zhitao; Raj, Kamal Abdul Rasheedj; Xia, Lingchao; Tu, Zhengkai; Chan, Siew Hwa

    2016-01-01

    Highlights: • Heating-up and cooling-down processes of HT-PEMFC are the mainly interested topics. • Dynamic behaviours, power and energy demand of the heating and cooling was studied. • Hybrid system based on LiFeYPO_4 battery for heating and cooling is built and tested. • The concept of combining different heating sources together is recommended. - Abstract: One key issue pertaining to the cold-start of High temperature PEM fuel cell (HT-PEMFC) is the requirement of high amount of thermal energy for heating up the stack to a temperature of 120 °C or above before it can generate electricity. Furthermore, cooling down the stack to a certain temperature (e.g. 50 °C) is necessary before stopping. In this study, the dynamic behaviours, power and energy demand of a 6 kW liquid cooled HT-PEMFC stack during heating-up, operation and cooling-down were investigated experimentally. The dynamic behaviours of fuel cell under heating-up and cooling-down processes are the mainly interested topics. Then a hybridisation of HT-PEMFC with Li-ion battery to demonstrate the synergistic effect on dynamic behaviour was conducted and validated for its feasibility. At last, the concept of combining different heating sources together is analysed to reduce the heating time of the HT-PEMFC as well.

  8. Assessment of In-vessel corium retention for VVER-440/V213

    International Nuclear Information System (INIS)

    Matejovic, P.; Barnak, M.; Bachraty, M.; Berky, R.

    2011-01-01

    In-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) has been recognised as a feasible and promising severe accident management strategy for VVER-440/V213 reactors. In general, the avoiding of boiling crisis on outer (cooled) RPV (reactor pressure vessel) surface is sufficient condition for preserving the RPV integrity. The crucial point of the proposed IVR concept for VVER-440/V213 is the narrow gap between elliptical lower head and thermal and biological shield. In the cold conditions the width of this gap is only about 2 cm and would be even lower in hot IVR conditions, when the reactor wall is subjected to large thermal gradients due to temperature difference between the hot inner surface (loaded by corium) and cold outer surface (which is cooled by water in flooded cavity). Sufficient gap should remain free for coolant flow for the success of the proposed IVR concept. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width are of primarily importance. Two different approaches were used for the estimation of the thermal load: a conservative approach and a transient approach, both were computed with the ASTEC code. The structural analysis of RPV subjected to IVR load was performed using the finite element method (FEM) code ANSYS release 10.0. From the results obtained it follows, that even when the RPV is subjected to limiting loading conditions during severe accident, there should be sufficient gap width (∼ 1 cm) between RPV wall and thermal/biological shield for the coolant flow in natural circulation regime alongside the outer surface of the RPV wall

  9. Heat transfer calculations on the KNK II emergency cooling system

    International Nuclear Information System (INIS)

    Vossebrecker, H.; Groenefeld, G.

    1976-12-01

    The Licensing Authority had demanded that in case of the change of the KNK thermal core into a fast core the decay heat removal system must be improved by a diverse and spatially separated emergency cooling system. In order to meet this requirement an existing nitrogen system of the facility is extended in such a manner that the decay heat will be removed by a nitrogen flow passing through the gap between reactor vessel and guard vessel. The heat transport from the core to the vessel is accomplished by natural convection flow rates which are generated by density differences between the hot core subassemblies, the reflector subassemblies and other passages between the upper and the lower plenum. The calculations show that the maximum temperatures in the core do not reach the sodium boiling-point. The maximum vessel temperature is 673 deg. C. In this report the function of the emergency cooling system and the methods of calculation are described, the input data and the results are stated and it is shown that the calculated temperatures are conservative [de

  10. Analyses on ex-vessel debris formation and coolability in SARNET frame

    International Nuclear Information System (INIS)

    Pohlner, G.; Buck, M.; Meignen, R.; Kudinov, P.; Ma, W.; Polidoro, F.; Takasuo, E.

    2014-01-01

    Highlights: • Melt outflow varies from dripping melt outflow to molten corium jets of variable size. • Experiments show clear trend of producing particles in size range 2-4 mm. • Code calculations show complete solidification of particles, yielding formation of fragmented debris beds. • Limits of debris bed cooling and coolability margins are analysed. - Abstract: The major aim of work in the SARNET2 European project on ex-vessel debris formation and coolability was to get an overall perspective on coolability of melt released from a failed reactor pressure vessel and falling into a water-filled cavity. Especially, accident management concepts for BWRs, dealing with deep water pools below the reactor vessel, are addressed, but also shallower pools in existing PWRs, with questions about partial cooling and time delay of molten corium concrete interaction. The subject can be divided into three main topics: (i) Debris bed formation by breakup of melt, (ii) Coolability of debris and (iii) Coupled treatment of the processes. Accompanied by joint collaborations of the partners, the performed work comprises theoretical, experimental and modelling activities. Theoretical work was done by KTH on the melt outflow conditions from a RPV and on the quantification of the probability of yielding a non-coolable ex-vessel bed by use of probabilistic assessment. IKE introduced a theoretical concept to improve debris bed coolability. A large amount of experimental work was done by partners (KTH, VTT, IKE) on the coolability of debris beds using different bed geometries, particles, heating methods and water feeds, yielding a valuable base for code validation. Modelling work was mainly done by IKE, IRSN, RSE and VTT concerning jet breakup and/or debris bed formation and cooling in 2D and 3D geometries. A benchmark for the DEFOR-A experiment of KTH was performed. Important progress was reached for several tasks and aspects and important insights are given, enabling to focus the

  11. Melt cooling by bottom flooding: The experiment CometPC-H3. Ex-vessel core melt stabilization research

    International Nuclear Information System (INIS)

    Alsmeyer, H.; Cron, T.; Merkel, G.; Schmidt-Stiefel, S.; Tromm, W.; Wenz, T.

    2003-03-01

    The CometPC-H3 experiment was performed to investigate melt cooling by water addition to the bottom of the melt. The experiment was performed with a melt mass of 800 kg, 50% metal and 50% oxide, and 300 kW typical decay heat were simulated in the melt. As this was the first experiment after repair of the induction coil, attention was given to avoid overload of the induction coil and to keep the inductor voltage below critical values. Therefore, the height of the sacrificial concrete layer was reduced to 5 cm only, and the height of the porous concrete layers was also minimized to have a small distance and good coupling between heated melt and induction coil. After quite homogeneous erosion of the upper sacrificial concrete layer, passive bottom flooding started from the porous concrete after 220 s with 1.3 liter water/s. The melt was safely stopped, arrested and cooled. The porous, water filled concrete was only slightly attacked by the hot melt in the upper 25 mm of one sector of the coolant device. The peak cooling rate in the early contact phase of coolant water and melt was 4 MW/m 2 , and exceeded the decay heat by one order of magnitude. The cooling rate remarkably dropped, when the melt was covered by the penetrating water and a surface crust was formed. Volcanic eruptions from the melt during the solidification process were observed from 360 - 510 s and created a volcanic dome some 25 cm high, but had only minor effect on the generation of a porous structure, as the expelled melt solidified mostly with low porosity. Unfortunately, decay heat simulation in the melt was interrupted at 720 s by an incorrect safety signal, which excluded further investigation of the long term cooling processes. At that time, the melt was massively flooded by a layer of water, about 80 cm thick, and coolant water inflow was still 1 l/s. The melt had reached a stable situation: Downward erosion was stopped by the cooling process from the water filled, porous concrete layer. Top

  12. Thermal and Fluid Mechanical Investigation of an Internally Cooled Piston Rod

    Science.gov (United States)

    Klotsche, K.; Thomas, C.; Hesse, U.

    2017-08-01

    The Internal Cooling of Reciprocating Compressor Parts (ICRC) is a promising technology to reduce the temperature of the thermally stressed piston and piston rod of process gas compressors. The underlying heat transport is based on the flow of a two-phase cooling medium that is contained in the hollow reciprocating assembly. The reciprocating motion forces the phases to mix, enabling an enhanced heat transfer. In order to investigate this heat transfer, experimental results from a vertically reciprocating hollow rod are presented that show the influence of different liquid charges for different working temperatures. In addition, pressure sensors are used for a crank angle dependent analysis of the fluid mechanical processes inside the rod. The results serve to investigate the two-phase flow in terms of the velocity and distribution of the liquid and vapour phase for different liquid fractions.

  13. Thermal-hydraulic evaluation study of the effectiveness of emergency core cooling system for light water reactors

    International Nuclear Information System (INIS)

    Sobajima, Makoto

    1985-08-01

    In order to evaluate the core cooling capability of the emergeny core cooling system, which is a safety guard system of light water reactors for a loss-of-coolant accident, a variety of large scale test were performed. Through the results, many phenomena were investigated and the predictabity of analytical codes were examined. The tests conducted were a single-vessel blowdown test, emergency core cooling test in a PWR simulation facility, spray cooling test for a BWR, large scale reflood test and a separate effect test on countercurrent flow. These test results were examined to clarify thermal-hydraulic phenomena and the effect of various test parameters and were utilized to improve predictability of the analytical codes. Some models for flow behavior in the upper core were also developed. By evaluating the effectiveness of various emergency core cooling system configurations, more effective cooling system than the current one was proposed and demonstrated. (author)

  14. Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights

    International Nuclear Information System (INIS)

    Ninokata, H.; Kamide, H.

    2011-01-01

    In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)

  15. Problems in manufacturing and transport of pressure vessels of integral reactors

    International Nuclear Information System (INIS)

    Kralovec, J.

    1997-01-01

    Integral water-cooled reactors are typical with eliminating large-diameter primary pipes and placing primary components, i.e. steam generators and pressurizers in reactor vessels. This arrangement leads to reactor pressure vessels of large dimensions: diameters, heights and thick walls and subsequently to great weights. Thus, even medium power units have pressure vessels which are on the very limit of present manufacturing capabilities. Principal manufacturing and inspection operations as well as pertinent equipment are concerned: welding, cladding, heat treatment, machining, shop-handling, non-destructive testing, hydraulic pressure tests etc. Tile transport of such a large and heavy component makes a problem which effects its design as well as the selection of the plant site. Railway, road and ship are possible ways of transport each of them having its advantages and limitations. Specific features and limits of the manufacture and transport of large pressure vessels are discussed in the paper. (author)

  16. Americium behaviour in plastic vessels.

    Science.gov (United States)

    Legarda, F; Herranz, M; Idoeta, R; Abelairas, A

    2010-01-01

    The adsorption of (241)Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of (241)Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of (241)Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification. Copyright 2009 Elsevier Ltd. All rights reserved.

  17. Elmo Bumpy Torus proof of principle, Phase II: Title 1 report. Volume II. Toroidal vessel

    International Nuclear Information System (INIS)

    1982-01-01

    The Toroidal Vessel provides the vacuum enclosure for containing the high temperature steady state plasma. In addition, the Toroidal Vessel must provide several viewing ports for plasma diagnostics, vacuum pumping ports for both high vacuum and roughing vacuum, feed-through ports for ECRH waveguides, limiter feed throughs for cooling and supporting the limiters, and ports for ion gages. The vessel must operate in an intense environment comprised of x-rays, microwaves, magnetic fields and plasma heat loads as well as the atmosphere pressure and gravity loads and the internal thermal stress loads due to heating and cooling of the torus. A key issue addressed was the choice of vacuum vessel seal and wall materials. In addition, during the course of the study, ORNL requested that horsecollar diagnostic ports be incorporated in the design. A comprehensive trade study was performed considering the vessel material issues in concert with the impact of the horsecollar port design. A change in baseline from an aluminum vessel with elastomer seals and circular diagnostic ports to austenitic stainless steel vessel with metal seals and horsecollar ports was agreed upon by both MDAC and ORNL towards the end of Title I

  18. Computational scheme for transient temperature distribution in PWR vessel wall

    International Nuclear Information System (INIS)

    Dedovic, S.; Ristic, P.

    1980-01-01

    Computer code TEMPNES is a part of joint effort made in Gosa Industries in achieving the technique for structural analysis of heavy pressure vessels. Transient heat conduction problems analysis is based on finite element discretization of structures non-linear transient matrix formulation and time integration scheme as developed by Wilson (step-by-step procedure). Convection boundary conditions and the effect of heat generation due to radioactive radiation are both considered. The computation of transient temperature distributions in reactor vessel wall when the water temperature suddenly drops as a consequence of reactor cooling pump failure is presented. The vessel is treated as as axisymmetric body of revolution. The program has two finite time element options a) fixed predetermined increment and; b) an automatically optimized time increment for each step dependent on the rate of change of the nodal temperatures. (author)

  19. Estimation of the lifetime of resin insulators against baking temperature for JT-60SA in-vessel coils

    Energy Technology Data Exchange (ETDEWEB)

    Sukegawa, Atsuhiko M., E-mail: morioka.atsuhiko@jaea.go.jp; Murakami, Haruyuki; Matsunaga, Go; Sakurai, Shinji; Takechi, Manabu; Yoshida, Kiyoshi; Ikeda, Yoshitaka

    2015-10-15

    Highlights: • The lifetime of resin insulators at about 200 °C was estimated. • We make use of the Arrhenius plot by the Weibull analysis for the estimation. • A suitable temperatures for the in-vessel coils were discussed. - Abstract: In the present study, the thermal endurance of epoxy-based, bismaleimides, and cyanate ester resins for the current design of the in-vessel coils was measured by performing acceleration tests to assess their insulation properties using the thermal endurance defined by the International Electrotechnical Commission (IEC-60216 Part1–Part 6) for a minimum of 5,000 h in the 180–240 °C temperature range. It was found that none of the resin insulators could tolerate the baking conditions of 40,000 h at ∼200 °C in the JT-60SA vacuum vessel. Therefore, the design of the in-vessel coils, including the error field correction coils (EFCC), was changed from the type without water cooling to with water cooling on JT-60SA.

  20. Venus Mobile Explorer with RPS for Active Cooling: A Feasibility Study

    Science.gov (United States)

    Leifer, Stephanie D.; Green, Jacklyn R.; Balint, Tibor S.; Manvi, Ram

    2009-01-01

    We present our findings from a study to evaluate the feasibility of a radioisotope power system (RPS) combined with active cooling to enable a long-duration Venus surface mission. On-board power with active cooling technology featured prominently in both the National Research Council's Decadal Survey and in the 2006 NASA Solar System Exploration Roadmap as mission-enabling for the exploration of Venus. Power and cooling system options were reviewed and the most promising concepts modeled to develop an assessment tool for Venus mission planners considering a variety of future potential missions to Venus, including a Venus Mobile Explorer (either a balloon or rover concept), a long-lived Venus static lander, or a Venus Geophysical Network. The concepts modeled were based on the integration of General Purpose Heat Source (GPHS) modules with different types of Stirling cycle heat engines for power and cooling. Unlike prior investigations which reported on single point design concepts, this assessment tool allows the user to generate either a point design or parametric curves of approximate power and cooling system mass, power level, and number of GPHS modules needed for a "black box" payload housed in a spherical pressure vessel.

  1. Experimental investigations on the cooling of a motorcycle helmet with phase change material (PCM

    Directory of Open Access Journals (Sweden)

    Fok S.C.

    2011-01-01

    Full Text Available The thermal comfort of motorcycle helmet during hot weather is important as it can affect the physiological and psychological condition of the rider. This paper examines the use of phase change material (PCM to cool a motorcycle helmet and presents the experimental investigations on the influences of the simulated solar radiation, wind speed, and heat generation rate on the cooling system. The result shows that the PCM-cooled helmet is able to prolong the thermal comfort period compared to a normal helmet. The findings also indicate that the heat generation from the head is the predominant factor that will affect the PCM melting time. Simulated solar radiation and ram-air due to vehicle motion under adiabatic condition can have very little influences on the PCM melting time. The results suggested that the helmet usage time would be influenced by the amount of heat generated from the head. Some major design considerations based on these findings have been included. Although this investigation focuses on the cooling of a motorcyclist helmet, the findings would also be useful for the development of PCM-cooling systems in other applications.

  2. Size dependence investigations of hot electron cooling dynamics in metal/adsorbates nanoparticles

    International Nuclear Information System (INIS)

    Bauer, Christophe; Abid, Jean-Pierre; Girault, Hubert H.

    2005-01-01

    The size dependence of electron-phonon coupling rate has been investigated by femtosecond transient absorption spectroscopy for gold nanoparticles (NPs) wrapped in a shell of sulfate with diameter varying from 1.7 to 9.2 nm. Broad-band spectroscopy gives an overview of the complex dynamics of nonequilibrium electrons and permits the choice of an appropriate probe wavelength for studying the electron-phonon coupling dynamics. Ultrafast experiments were performed in the weak perturbation regime (less than one photon in average per nanoparticle), which allows the direct extraction of the hot electron cooling rates in order to compare different NPs sizes under the same conditions. Spectroscopic data reveals a decrease of hot electron energy loss rates with metal/adsorbates nanosystem sizes. Electron-phonon coupling time constants obtained for 9.2 nm NPs are similar to gold bulk materials (∼1 ps) whereas an increase of hot electron cooling time up to 1.9 ps is observed for sizes of 1.7 nm. This is rationalized by the domination of surface effects over size (bulk) effects. The slow hot electron cooling is attributed to the adsorbates-induced long-lived nonthermal regime, which significantly reduces the electron-phonon coupling strength (average rate of phonon emission)

  3. Site-specific investigations of aquifer thermal energy storage for space and process cooling

    International Nuclear Information System (INIS)

    Brown, D.R.

    1991-01-01

    This paper reports on the Pacific Northwest Laboratory (PNL) that has completed three preliminary site-specific feasibility studies that investigated aquifer thermal energy storage (ATES) for reducing space and process cooling costs. Chilled water stored in an ATES system could be used to meet all or part of the process and/or space cooling loads at the three facilities investigated. Seasonal or diurnal chill ATES systems could be significantly less expensive than a conventional electrically-driven, load-following chiller system at one of the three sites, depending on the cooling water loop return temperature and presumed future electricity escalation rate. For the other two sites investigated, a chill ATES system would be economically competitive with conventional chillers if onsite aquifer characteristics were improved. Well flow rates at one of the sites were adequate, but the expected thermal recovery efficiency was too low. The reverse of this situation was found at the other site, where the thermal recovery efficiency was expected to be adequate, but well flow rates were too low

  4. Investigations on passive containment cooling

    International Nuclear Information System (INIS)

    Knebel, J.U.; Cheng, X.; Neitzel, H.J.; Erbacher, F.J.; Hofmann, F.

    1997-01-01

    The composite containment design for advanced LWRs that has been examined under the PASCO project is a promising design concept for purely passive decay heat removal after a severe accident. The passive cooling processes applied are natural convection and radiative heat transfer. Heat transfer through the latter process removes at an emission coefficient of 0.9 about 50% of the total heat removed via the steel containment, and thus is an essential factor. The heat transferring surfaces must have a high emission coefficient. The sump cooling concept examined under the SUCO project achieves a steady, natural convection-driven flow from the heat source to the heat sink. (orig./CB) [de

  5. Conceptual design of EAST flexible in-vessel inspection system

    International Nuclear Information System (INIS)

    Peng, X.B.; Song, Y.T.; Li, C.C.; Lei, M.Z.; Li, G.

    2010-01-01

    Remote handling technology, especially the flexible in-vessel inspection system (FIVIS) without breaking the working condition of the vacuum vessel, has been identified as one major challenge on the maintenance for the future tokamak fusion reactor. The FIVIS introduced here is specially developed for EAST superconducting tokamak that has actively cooled plasma facing components (PFCs). It aims flexible close-up inspection of EAST PFCs to help the understanding of operation issues that could occur in the vacuum vessel. This paper resumes the preliminary work of the FIVIS project, including the requirement analysis and the development of the conceptual design. The FIVIS consists out of a long reach multi-articulated manipulator and a process tool. The manipulator has a modular design for its subsystems and can reach all areas of the first wall in the distance of 15 mm and in the range of ±90 o along toroidal direction. It will be folded and hidden in the designated horizontal port during plasma discharge period.

  6. Analytical and experimental investigation of closed-cycle sorption cooling systems

    Science.gov (United States)

    Liu, Lianquan

    1992-01-01

    The first part of the present thesis concerns the Coefficient of Performance (COP) of two types of closed-cycle sorption cooling systems: the Single Effect Liquid (SEL) absorption system and the Regenerative Solid (RS) adsorption system. When specific cycle configurations are considered, the COP is always less than that allowed by the second law. The potential of the two systems to approach the second law limit is considered in this work. The analysis shows that COP of a SEL system using LiBr-H2O is not limited by one, as believed before, and that the COP of a RS cooling system using zeolite-water is considerably larger than that of the SEL system. This is due to recovery of the heat of adsorption which is made possible by capturing the thermal wave in the solid adsorbent. In the second part, a one dimensional model has been developed for a real RS cooling system featured by finite heat transfer coefficients. The problem is solved numerically to yield the temperature and uptake profiles, COP, and cooling capacity and cooling rates. The effects of various design and operating parameters on system performance have been investigated by using the model. The convective heat transfer coefficient at the inner wall of the fluid channel passing through the zeolite columns, the flow rate of the heat transfer fluid, the condenser and evaporator temperature are identified as the most significant factors. A new correlation of adsorption equilibrium has been derived in this thesis. The derivation is based on established thermodynamic relationships and is shown to be able to well represent the data of three adsorption pairs widely used in sorption cooling applications: zeolite-water, silica gel-water and activated carbon-methanol. Finally, in the experimental part of the present work a test set-up of a zeolite-water heat and mass regenerator was designed, instrumented and built. Temperature profiles at various operating conditions were measured. The data of a 'single blow' mode

  7. The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

    International Nuclear Information System (INIS)

    Peacock, A.; Girlinger, A.; Vorkoeper, A.; Boscary, J.; Greuner, H.; Hurd, F.; Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G.

    2011-01-01

    320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m 2 , are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m 2 and a maximum heat load of 200 kW/m 2 with a water velocity of approximately 2 m s -1 . High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m 2 for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

  8. Structure of liquid metal cooled nuclear reactor with loops and steady vessel

    International Nuclear Information System (INIS)

    Costes, D.

    1990-01-01

    This structure comprises, in a vessel containing liquid metal, a nuclear core steadied on an alimentation diagrid and external loops comprising heat exchanger and reinjection pump of sodium in the diagrid. The vessel has the bottom resting on the concrete surround with a thermal stratification of the sodium between the bottom and the diagrid. This disposition has for advantage to allow a vertical connection of the sodium reinjection channel. This channel is contained in a metal sheath with a sliding leak tightness [fr

  9. Investigation of impulsively loaded pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.; Cornwell, R.; Hanner, D.; Leichter, H.; Mohr, P.

    1963-10-15

    Explosion containment vessels for containing from 2,000 to 3,000 five ton nuclear explosions are considered. Analysis methods appear adequate and lowest weights using the most advanced materials available in the next five years are projected.None of these materials can be fabricated today and all require extensive development. Present material technology limits the choice of materials and defines the weight. The addition of safety factors and fixtures (nozzles, etc.) will add to this weight considerably, and may well radically alter the vessel response. Improvements in the strength weight ratios of metals and glasses over those considered in this report do not appear reasonable at this time. Winding schemes to utilize the high strength of steel wires and somehow maintain a reasonable thickness appear to offer the most promise. A `ductile` beryllium would of course offer vast improvement, but no indications that this is being developed have appeared and all presently known beryllium is much too brittle.

  10. Investigation of Condensation Heat Transfer Correlation of Heat Exchanger Design in Secondary Passive Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Yun Jae; Lee, Hee Joon [Kookmin Univ., Seoul (Korea, Republic of); Kang, Hanok; Lee, Taeho; Park, Cheontae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-12-15

    Recently, condensation heat exchangers have been studied for applications to the passive cooling systems of nuclear plants. To design vertical-type condensation heat exchangers in secondary passive cooling systems, TSCON (Thermal Sizing of CONdenser), a thermal sizing program for a condensation heat exchanger, was developed at KAERI (Korea Atomic Energy Research Institute). In this study, the existing condensation heat transfer correlation of TSCON was evaluated using 1,157 collected experimental data points from the heat exchanger of a secondary passive cooling system for the case of pure steam condensation. The investigation showed that the Shah correlation, published in 2009, provided the most satisfactory results for the heat transfer coefficient with a mean absolute error of 34.8%. It is suggested that the Shah correlation is appropriate for designing a condensation heat exchanger in TSCON.

  11. High temperature gas cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hosegood, S.B.; Lockett, G.E.

    1975-01-01

    For high-temperature gas cooled reactors it is considered advantageous to design the core so that the moderator blocks can be removed and replaced by some means of standpipes normally situated in the top of the reactor vessel. An arrangement is here described to facilitate these operations. The blocks have end faces shaped as irregular hexagons with three long sides of equal length and three short sides also of equal length, one short side being located between each pair of adjacent long sides, and the long sides being inclined towards one another at 60 0 . The block defines a number of coolant channels located parallel to its sides. Application of the arrangement to a high temperature gas-cooled reactor with refuelling standpipes is described. The standpipes are located in the top of the reactor vessel above the tops of the columns and are disposed coaxially above the hexagonal channels, with diameters that allow the passage of the blocks. (U.K.)

  12. Ex-vessel debris coolability test during severe accident (COTELS project)

    International Nuclear Information System (INIS)

    Ogasawara, H.

    1998-01-01

    The objectives of the COTELS project are for severe accident management, to investigate phenomena of ex-vessel fuel-coolant interactions after reactor pressure vessel (RPV) failure and to investigate molten core-concrete interaction when coolant is injected onto molten debris. The project has being cooperated with the National Nuclear Center in the Republic of Kazakstan from 1994 to 1997 under the sponsorship of the Ministry of International Trade and Industry of Japan. Total programs are composed with the following tests. (1) Test 01 was meant to observe flow mode of falling debris. (2) Test A was meant to investigate phenomena of fuel-coolant interactions when molten debris falls into a coolant pool. (3) Test B/C investigated fuel coolant interactions and molten core-concrete interaction when coolant is injected onto debris. Detail data evaluation is underway. The following results were thus for obtained: (1) It was confirmed in Test 01 series that about 60 kg of UO 2 mixture was completely melted and fallen as a continuous jet. (2) No energetic fuel-coolant interaction was observed both in Test A and B series. (3) Debris in which decay heat was simulated was cooled by water injection in Test C series

  13. In vessel core melt progression phenomena

    International Nuclear Information System (INIS)

    Courtaud, M.

    1993-01-01

    For all light water reactor (LWR) accidents, including the so called severe accidents where core melt down can occur, it is necessary to determine the amount and characteristics of fission products released to the environment. For existing reactors this knowledge is used to evaluate the consequences and eventual emergency plans. But for future reactors safety authorities demand decrease risks and reactors designed in such a way that fission products are retained inside the containment, the last protective barrier. This requires improved understanding and knowledge of all accident sequences. In particular it is necessary to be able to describe the very complex phenomena occurring during in vessel core melt progression because they will determine the thermal and mechanical loads on the primary circuit and the timing of its rupture as well as the fission product source term. On the other hand, in case of vessel failure, knowledge of the physical and chemical state of the core melt will provide the initial conditions for analysis of ex-vessel core melt progression and phenomena threatening the containment. Finally a good understanding of in vessel phenomena will help to improve accident management procedures like Emergency Core Cooling System water injection, blowdown and flooding of the vessel well, with their possible adverse effects. Research and Development work on this subject was initiated a long time ago and is still in progress but now it must be intensified in order to meet the safety requirements of the next generation of reactors. Experiments, limited in scale, analysis of the TMI 2 accident which is a unique source of global information and engineering judgment are used to establish and assess physical models that can be implemented in computer codes for reactor accident analysis

  14. System for bearing a nuclear reactor vessel cooled by liquid metal

    International Nuclear Information System (INIS)

    Mahe, A.; Jullien, G.

    1976-01-01

    The invention relates to a bearing system for supporting a nuclear reactor vessel of the kind which is suspended from the reactor closure slab. The bearing system comprises a ring connected at one end to a collar and at the other end to two collars. The collar connected to the bottom end of the ring forms the top part of the vessel to be supported while the other two collars fit into the slab at two separate places. The ring and collars are disposed in an annular space formed in the slab and dividing it into two parts, i.e., a central part and a peripheral part surrounding the central part of the slab

  15. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H

    2001-11-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable.

  16. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H.

    2001-01-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable

  17. Multiphase flow in ex-vessel coolability: development of an innovative concept

    International Nuclear Information System (INIS)

    Corradini, Michael L.

    2006-01-01

    The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific Advanced Light Water Reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability

  18. The integral analysis of 40 mm diameter pipe rupture in cooling system of fusion facility W7-X with ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Kačegavičius, Tomas, E-mail: Tomas.Kacegavicius@lei.lt; Povilaitis, Mantas, E-mail: Mantas.Povilaitis@lei.lt

    2015-12-15

    Highlights: • The analysis of loss-of-coolant accident (LOCA) in W7-X facility. • Burst disc is sufficient to prevent pressure inside the plasma vessel exceeding 110 kPa. • Developed model of the cooling system adequately represents the expected phenomena. - Abstract: Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental facility of stellarator type, which is currently being built at the Max-Planck-Institute for Plasmaphysics located in Greifswald, Germany. W7-X shall demonstrate that in future the energy could be produced in such type of fusion reactors. The safety analysis is required before the operation of the facility could be started. A rupture of 40 mm diameter pipe, which is connected to the divertor unit (module for plasma cooling) to ensure heat removal from the vacuum vessel in case of no-plasma operation mode “baking” is one of the design basis accidents to be investigated. During “baking” mode the vacuum vessel structures and working fluid – water are heated to the temperature 160 °C. This accident was selected for the detailed analysis using integral code ASTEC, which is developed by IRSN (France) and GRS mbH (Germany). This paper presents the integral analysis of W7-X response to a selected accident scenario. The model of the main cooling circuit and “baking” circuit was developed for ASTEC code. There were analysed two cases: (1) rupture of a pipe connected to the upper divertor unit and (2) rupture of a pipe connected to the lower divertor unit. The results of analysis showed that in both cases the water is almost completely released from the units into the plasma vessel. In both cases the pressure in the plasma vessel rapidly increases and in 28 s the set point for burst disc opening is reached preventing further pressurisation.

  19. UK regulatory aspects of prestressed concrete pressure vessels for gas-cooled reactor nuclear power stations

    International Nuclear Information System (INIS)

    Watson, P.S.

    1990-01-01

    Safety assessment principles for nuclear power plants and for nuclear chemical plants demand application of best proven techniques, recognised standards, adequacy margins, inspection and maintenance of all the components including prestressed concrete pressure vessels. In service inspection of prestressed concrete pressure vessels includes: concrete surface examination; anchorage inspection; tendon load check; tendon material examination; foundation settlement and tilt; log-term deformation; vessel temperature excursions; coolant loss; top cap deflection. Hartlepool and Heysham 1 power plants prestress shortfall problem is discussed. Main recommendations can be summarised as follows: at all pressure vessel stations prestress systems should be calibrated in a manner which results in all load bearing components being loaded in a representative manner; at all pressure vessel stations load measurements during calibration should be verified by a redundant and diverse system

  20. Experimental studies of oxidic molten corium-vessel steel interaction

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V.

    2001-01-01

    The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere

  1. Experimental studies of oxidic molten corium-vessel steel interaction

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. E-mail: niti-npc@sbor.net; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V

    2001-12-01

    The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere.

  2. Parametric Investigation of Brayton Cycle for High Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Chang Oh

    2004-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. In this project, we are investigating helium Brayton cycles for the secondary side of an indirect energy conversion system. Ultimately we will investigate the improvement of the Brayton cycle using other fluids, such as supercritical carbon dioxide. Prior to the cycle improvement study, we established a number of baseline cases for the helium indirect Brayton cycle. These cases look at both single-shaft and multiple-shaft turbomachinery. The baseline cases are based on a 250 MW thermal pebble bed HTGR. The results from this study are applicable to other reactor concepts such as a very high temperature gas-cooled reactor (VHTR), fast gas-cooled reactor (FGR), supercritical water reactor (SWR), and others. In this study, we are using the HYSYS computer code for optimization of the helium Brayton cycle. Besides the HYSYS process optimization, we performed parametric study to see the effect of important parameters on the cycle efficiency. For these parametric calculations, we use a cycle efficiency model that was developed based on the Visual Basic computer language. As a part of this study we are currently investigated single-shaft vs. multiple shaft arrangement for cycle efficiency and comparison, which will be published in the next paper. The ultimate goal of this study is to use supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency to values great than that of the helium Brayton cycle. This paper includes preliminary calculations of the steady state overall Brayton cycle efficiency based on the pebble bed reactor reference design (helium used as the working fluid) and compares those results with an initial calculation of a CO2 Brayton cycle

  3. Conceptual Design of Electrical Propulsion System for Nuclear Operated Vessel Adventurer

    International Nuclear Information System (INIS)

    Halimi, B.; Suh, K. Y.

    2009-01-01

    A design concept of the electric propulsion system for the Nuclear Operated Vessel Adventure (NOVA) is presented. NOVA employs Battery Omnibus Reactor Integral System (BORIS), a liquid metal cooled small fast integral reactor, and Modular Optimized Brayton Integral System (MOBIS), a supercritical CO 2 (SCO 2 ) Brayton cycle as power converter to Naval Application Vessel Integral System (NAVIS)

  4. Investigation and analysis on ITER in-vessel coils’ raw-materials

    International Nuclear Information System (INIS)

    Jin, Huan; Wu, Yu; Long, Feng; Yu, Min; Han, Qiyang; Liu, Huajun

    2013-01-01

    Highlights: • The R and D works for the ITER in-vessel coils (IVC) are now being conducted in Institute of Plasma Physics, and the analysis work are being done by Princeton Plasma Physics Laboratory. • There is little published paper about the raw materials for ITER IVC coils. • This manuscript points out the progress of the selected materials for ITER IVC coils. -- Abstract: The ITER in-vessel coils (IVCs) consist of 27 coils edge localized modes (ELM) and 2 coils vertical stabilization (VS) which are all mounted on the vacuum vessel wall behind the shield modules. The IVCs design and manufacturing work is being conducted in between Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and Princeton Plasma Physics Laboratory (PPPL). Because the position of ELM and VS coils is close and face to the plasma, the IVCs must undergo a severe environment, such as the high dose of radiation and high operation temperature, thus the conventional electrical insulation materials cannot be used. And the technology of “Stainless Steel Jacketed Mineral Insulated Conductor” (SSMIC) is deemed as the best choice to provide the necessary radiation resistance and compatibility strength in ITER's vacuum vessel. While mineral insulated conductor technology is not new, and is similar to the mineral insulated cable used in industrial. Some difficulties still need to be solved, such as searching for the proper raw-materials to make sure that the conductor have the properties of high current carrying capability, the necessary radiation resistance, the proper strength, at the same time, it must be come true in manufacture technology. This paper described the analysis of the materials for VS and ELM coil conductor

  5. Muscle-Cooling Intervention to Reduce Fatigue and Fatigue-Induced Tremor in Novice and Experienced Surgeons: A Preliminary Investigation.

    Science.gov (United States)

    Jensen, Lauren; Dancisak, Michael; Korndorffer, James

    2016-10-01

    A localized, intermittent muscle-cooling protocol was implemented to determine cooling garment efficacy in reducing upper extremity muscular fatigue and tremor in novice ( n  = 10) and experienced surgeons ( n  = 9). Subjects wore a muscle-cooling garment while performing multiple trials of a forearm exercise and paired suturing task to induce muscular fatigue and exercise-induced tremor. A reduction in tremor amplitude and an extension in time to fatigue were expected with muscle cooling as compared with control trials. Each subject completed an intervention session (5°C cooling condition) and a control session (32°C or thermal neutral condition). A paired samples t test indicated that tremor amplitude was significantly reduced ( t [8] = 1.89458; p  effect was not significant. Time to fatigue and suture time improved in both cohorts with muscle cooling, but the effect did not reach significance. Results from the pilot work suggest muscle cooling as an intervention for reduction of fatigue and tremor is very promising, warranting further investigation. Surgical specialties that require prolonged procedures might benefit more from this intervention.

  6. Control of Non-linear Marine Cooling System

    DEFF Research Database (Denmark)

    Hansen, Michael; Stoustrup, Jakob; Bendtsen, Jan Dimon

    2011-01-01

    We consider the problem of designing control laws for a marine cooling system used for cooling the main engine and auxiliary components aboard several classes of container vessels. We focus on achieving simple set point control for the system and do not consider compensation of the non-linearitie......-linearities, closed circuit flow dynamics or transport delays that are present in the system. Control laws are therefore designed using classical control theory and the performance of the design is illustrated through two simulation examples....

  7. Manufacture of electron beam irradiation vessel and its characteristics

    International Nuclear Information System (INIS)

    Kanazawa, Takao; Haruyama, Yasuyuki; Yotsumoto, Keiichi

    1992-05-01

    Electron beam irradiation vessel, which is used for the irradiation of samples under an inert or a vacuum atmosphere, is made by considering the temperature control during or after irradiation. The vessel was composed of the temperature controlable samples supporting plate, beam slit with water cooling plate and the insert of thermosensor. The four samples supporting plate was produced with the materials made up of aluminium, stainless steel (SUS304), and copper. The stainless steel supporting plate has a heater inside the cooling pipes for the high temperature treatment of samples without exposure to atmosphere after the irradiation. In this report, the temperature distribution and dose characteristics such as dose distribution and effects of backscattered electron were studied by using several supporting plate and the comparison of the experimental results with the simulated results was also carried out. (author)

  8. Nuclear reactor installation with outer shell enclosing a primary pressure vessel

    International Nuclear Information System (INIS)

    1975-01-01

    The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel containing the heat source, an outer shell enclosing the primary pressure vessel and acting as a secondary means of containment for this vessel against outside projectiles. Multiple auxiliary equipment points are arranged outside the outer shell which comprises a part of a lower wall around the primary pressure vessel, an annular part integrated in the lower wall and extending outwards as from this wall and an upper part integrated in the annular part and extending above this annular part and above the primary pressure vessel. The annular part and the primary pressure vessel are formed with vertical penetrations which can be closed communicating respectively with the auxiliary equipment points and with inside the pressure vessel whilst handling gear is provided in the upper part for vertically raising reactor components through these penetrations and for transporting them over the annular part and over the primary pressure vessel [fr

  9. Cool colored coating and phase change materials as complementary cooling strategies for building cooling load reduction in tropics

    International Nuclear Information System (INIS)

    Lei, Jiawei; Kumarasamy, Karthikeyan; Zingre, Kishor T.; Yang, Jinglei; Wan, Man Pun; Yang, En-Hua

    2017-01-01

    Highlights: • Cool colored coating and PCM are two complementary passive cooling strategies. • A PCM cool colored coating system is developed. • The coating reduces cooling energy by 8.5% and is effective yearly in tropical Singapore. - Abstract: Cool colored coating and phase change materials (PCM) are two passive cooling strategies often used separately in many studies and applications. This paper investigated the integration of cool colored coating and PCM for building cooling through experimental and numerical studies. Results showed that cool colored coating and PCM are two complementary passive cooling strategies that could be used concurrently in tropical climate where cool colored coating in the form of paint serves as the “first protection” to reflect solar radiation and a thin layer of PCM forms the “second protection” to absorb the conductive heat that cannot be handled by cool paint. Unlike other climate zones where PCM is only seasonally effective and cool paint is only beneficial during summer, the application of the proposed PCM cool colored coating in building envelope could be effective throughout the entire year with a monthly cooling energy saving ranging from 5 to 12% due to the uniform climatic condition all year round in tropical Singapore.

  10. Experimental Investigation of Mechanical Properties of PVC Polymer under Different Heating and Cooling Conditions

    Directory of Open Access Journals (Sweden)

    Sarkawt Rostam

    2016-01-01

    Full Text Available Due to a widely increasing usage of polymers in various industrial applications, there should be a continuous need in doing research investigations for better understanding of their properties. These applications require the usage of the polymer in different working environments subjecting the material to various temperature ranges. In this paper, an experimental investigation of mechanical properties of polyvinyl chloride (PVC polymer under heating and cooling conditions is presented. For this purpose standard samples are prepared and tested in laboratory using universal material testing apparatus. The samples are tested under different conditions including the room temperature environment, cooling in a refrigerator, and heating at different heating temperatures. It is observed that the strength of the tested samples decreases with the increasing of heating temperature and accordingly the material becomes softer. Meanwhile the cooling environments give a clear increasing to the strength of the material.

  11. A method for increasing the homogeneity of the temperature distribution during magnetic fluid hyperthermia with a Fe-Cr-Nb-B alloy in the presence of blood vessels

    Science.gov (United States)

    Tang, Yundong; Flesch, Rodolfo C. C.; Jin, Tao

    2017-06-01

    Magnetic hyperthermia ablates tumor cells by absorbing the thermal energy from magnetic nanoparticles (MNPs) under an external alternating magnetic field. The blood vessels (BVs) within tumor region can generally reduce treatment effectiveness due to the cooling effect of blood flow. This paper aims to investigate the cooling effect of BVs on the temperature field of malignant tumor regions using a complex geometric model and numerical simulation. For deriving the model, the Navier-Stokes equation for blood flow is combined with Pennes bio-heat transfer equation for human tissue. The effects on treatment temperature caused by two different BV distributions inside a mammary tumor are analyzed through numerical simulation under different conditions of flow rate considering a Fe-Cr-Nb-B alloy, which has low Curie temperature ranging from 42 °C to 45 °C. Numerical results show that the multi-vessel system has more obvious cooling effects than the single vessel one on the temperature field distribution for hyperthermia. Besides, simulation results show that the temperature field within tumor area can also be influenced by the velocity and diameter of BVs. To minimize the cooling effect, this article proposes a treatment method based on the increase of the thermal energy provided to MNPs associated with the adoption of low Curie temperature particles recently reported in literature. Results demonstrate that this approach noticeably improves the uniformity of the temperature field, and shortens the treatment time in a Fe-Cr-Nb-B system, thus reducing the side effects to the patient.

  12. Overview of experimental results obtained under the Prestressed Concrete Nuclear Pressure Vessel Development Program at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Naus, D.J.

    1978-01-01

    Under the Prestressed Concrete Nuclear Pressure Vessel Development Program at the Oak Ridge National Laboratory, various aspects of Prestressed Concrete Pressure Vessels (PCPVs) are investigated and evaluated with respect to reliability, structural performance, constructability, and economy. Based upon identified needs, analytical and experimental investigations are conducted. Areas of interest include finite-element analysis development, materials and structural behavior tests, instrumentation evaluation and development, and structural model tests. Studies have been recently completed in the following areas: concrete embedment instrumentation systems for PCPVs, grouted-nongrouted prestressing systems, acoustic emission as a technique for structural integrity monitoring, and model tests of steam-generator cavity closure plugs for a Gas-Cooled Fast Reactor (GCFR). An overview of results is presented

  13. Visual function and retinal vessel diameters during hyperthermia in man

    DEFF Research Database (Denmark)

    Jensen, Bettina Hagström; Bram, Thue; Kappelgaard, Per

    2017-01-01

    .01), a 10.6-mmHg mean reduction in diastolic blood pressure (p vein...... laser ophthalmoscopy was used to measure retinal trunk vessel diameters. Assessment was made at baseline, during hyperthermia and after cooling. RESULTS: The induction of a mean increase in core body temperature of 1.02°C was associated with a 7.15-mmHg mean reduction in systolic blood pressure (p

  14. Investigation on the Energy Saving Potential of Using a Novel Dew Point Cooling System in Data Centres

    Directory of Open Access Journals (Sweden)

    Yin Bi

    2017-10-01

    Full Text Available Abstract: Information technology (IT has brought significant changes in people’s lives. As an important part of the IT industry, data centres (DCs have been rapidly growing in both the number and size over the past 40 years. Around 30% to 40% of electricity consumption in DCs is used for space cooling, thus leading to very inefficient DC operation. To identify ways to reduce the energy consumption for space cooling and increase the energy efficiency of DCs’ operation, a dedicated investigation into the energy usage in DCs has been undertaken and a novel high performance dew point cooling system was introduced into a DC operational scheme. Based on the cooling load in DCs, a case study was carried out to evaluate the energy consumptions and energy usage effectiveness when using the novel dew point cooling system in different scales of DCs in various climates. It was found that by using the novel dew point cooling system, for 10 typical climates a DC can have a much lower power usage effectiveness (PUE of 1.10 to 1.22 compared to that of 1.7 to 3.7 by using existing traditional cooling systems, leading to significantly increased energy efficiency of the DC operation. In addition, the energy performance by managing the cooling air supply at the different levels in DCs, i.e., room, row and rack level, was simulated by using a dynamic computer model. It was found that cooling air supply at rack level can provide a higher energy efficiency in DCs. Based on the above work, the energy saving potential in DCs was conducted by comparing DCs using an the novel dew point cooling system and the optimum management scheme for the cooling air supply to that using traditional air cooling systems and the same supply air management. Annual electricity consumptions for the two cases were given. It was found that by using the novel dew point cooling system and optimum management system for the cooling air supply, an 87.7~91.6% electricity consumption saving for

  15. Lay-out and construction of a pressure vessel built-up of cast steel segments for a pebble-bed high temperature reactor with a thermal power of 3000 MW

    International Nuclear Information System (INIS)

    Voigt, J.

    1978-03-01

    The prestressed cast vessel is an alternative to the prestressed concrete vessel for big high temperature reactors. In this report different cast steel vessel concepts for an HTR for generation of current with 3000 MW(th) are compared concerning their realization and economy. The most favourable variant serves as a base for the lay-out of the single vessel components as cast steel segments, bracing, cooling and outer sealing. Hereby the actual available possibilities of production and transport are considered. For the concept worked out possibilities of inspection and repair are suggested. A comparison of costs with adequate proposititons of the industry for a prestressed concrete and a cast iron pressure vessel investigates the economical competition. (orig.) [de

  16. Heavy liquid metal cooled FBR. Results 2003

    International Nuclear Information System (INIS)

    Hayahune, Hiroki; Enuma, Yasuhiro; Soman, Yoshindo; Konomura, Mamoru; Mizuno, Tomoyasu

    2004-08-01

    Concepts of the reactor, SG and main coolant pump have been studied considering maintainability and aseismic capability, which is a medium size pool type lead-bismuth cooled reactor. The results are following. (1) Reconsideration of reactor design concepts concerning maintainability: In pursuit of good reactor maintainability, the structural concepts of SG, UIS and core support structures have been changed to be drawn up above the upper area of the reactor system. After a few decade of interval, lead-bismuth inventory in the reactor vessel shall be fully drained for easy ISI operation of in-vessel main components such as core support structures. From the viewpoint of the reactor aseismic capability, the axial length of reactor vessel was reduced and the reactor vessel support location was changed from the top handing to the circumference of the vessel. (2) SG concept selection in conjunction with a compact reactor vessel: The concept of SG consisting of a once through type with helical coil tube is selected. 6 units of a small scale SG are arranged on a reactor roof deck along the peripheral direction, in addition to 3 units of a centrifugal mechanical pump. (3) Aseismic structural integrity of the reactor components: Aseismic structural integrity of the reactor vessel, core support structures, UIS, FHM, SG and the main pumps has been vigorously examined respectively. These components besides FHM could keep the aseismic structural integrity for strong S2 earthquake under the design condition. FHM could also keep the integrity for S1 earthquake. (4) Safety evaluation: Thermal transients following loss of flow type accident due to plant total blackout and typical manual reactor trip incident, have been evaluated to assure the pant safety design, by analyzing thermal hydraulic behavior of transients concerning core flow rate and temperatures of the plant cooling system. Loss of flow accident due to plant total blackout: The reactor coolant pumps shall be tripped and

  17. Decay heat removal analyses in heavy-liquid-metal-cooled fast breeding reactors. Development of the thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, Takaaki; Enuma, Yasuhiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwasaki, Takashi [Nuclear Energy System Inc., Tokyo (Japan); Ohyama, Kazuhiro [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

    2001-05-01

    The feasibility study on future commercial fast breeder reactors in Japan has been conducted at JNC, in which various plant design options with all the possible coolant and fuel types are investigated to determine the conditions for the future detailed study. Lead-bismuth eutectic coolant has been selected as one of the possible coolant options. During the phase-I activity of the feasibility study in FY1999 and FY2000, several plant concepts, which were cooled by the heavy liquid metal coolant, were examined to evaluate the feasibility mainly with respect to economical competitiveness with other coolant reactors. A medium-scale (300 - 550 MWe) plant, cooled by a lead-bismuth natural circulation flow in a pool type vessel, was selected as the most possible plant concept for the heavy liquid metal coolant. Thus, a conceptual design study for a lead-bismuth-cooled, natural-circulation reactor of 400 MWe has been performed at JNC to identify remaining difficulties in technological aspect and its construction cost evaluation. In this report, a thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors is described. A Multi-dimensional Steam Generator analysis code (MSG) was applied to evaluate the natural circulation plant by combination with a flow-network-type, plant dynamics code (Super-COPD). By using this combined multi-dimensional plant dynamics code, decay heat removals, ULOHS and UTOP accidents were evaluated for the 100 MWe STAR-LM concept designed by ANL. In addition, decay heat removal by the Primary Reactor Auxiliary Cooling System (PRACS) in the 400 MWe lead-bismuth-cooled, natural-circulation reactor, being studied at JNC, was analyzed. In conclusion, it becomes clear that the combined multi-dimensional plant dynamics code is suitably applicable to analyses of lead-bismuth-cooled, natural-circulation reactors to evaluate thermal-hydraulic phenomena during steady-state and transient conditions. (author)

  18. A fracture mechanics method of evaluating structural integrity of a reactor vessel due to thermal shock effects following LOCA condition

    International Nuclear Information System (INIS)

    Ramani, D.T.

    1977-01-01

    The importance of knowledge of structural integrity of a reactor vessel due to thermal shock effects, is related to safety and operational requirements in assessing the adequacy and flawless functioing of the nuclear power systems. Followig a loss-of-coolant accident (LOCA) condition the integrity of the reactor vessel due to a sudden thermal shock induced by actuation of emergency core cooling system (ECCS), must be maintained to ensure safe and orderly shutdown of the reactor and its components. The paper encompasses criteria underlaying a fracture mechanics method of analysis to evaluate structural integrity of a typical 950 MWe PWR vessel as a result of very drastic changes in thermal and mechanical stress levels in the reactor vessel wall. The main object of this investigation therefore consists in assessing the capability of a PWR vessel to withstand the most critical thermal shock without inpairing its ability to conserve vital coolant owing to probable crack propagation. (Auth.)

  19. Thermal investigation of lithium-ion battery module with different cell arrangement structures and forced air-cooling strategies

    International Nuclear Information System (INIS)

    Wang, Tao; Tseng, K.J.; Zhao, Jiyun; Wei, Zhongbao

    2014-01-01

    Highlights: • Three-dimensional CFD model with forced air cooling are developed for battery modules. • Impact of different air cooling strategies on module thermal characteristics are investigated. • Impact of different model structures on module thermal responses are investigated. • Effect of inter-cell spacing on cell thermal characteristics are also studied. • The optimal battery module structure and air cooling strategy is recommended. - Abstract: Thermal management needs to be carefully considered in the lithium-ion battery module design to guarantee the temperature of batteries in operation within a narrow optimal range. This article firstly explores the thermal performance of battery module under different cell arrangement structures, which includes: 1 × 24, 3 × 8 and 5 × 5 arrays rectangular arrangement, 19 cells hexagonal arrangement and 28 cells circular arrangement. In addition, air-cooling strategies are also investigated by installing the fans in the different locations of the battery module to improve the temperature uniformity. Factors that influence the cooling capability of forced air cooling are discussed based on the simulations. The three-dimensional computational fluid dynamics (CFD) method and lumped model of single cell have been applied in the simulation. The temperature distributions of batteries are quantitatively described based on different module patterns, fan locations as well as inter-cell distance, and the conclusions are arrived as follows: when the fan locates on top of the module, the best cooling performance is achieved; the most desired structure with forced air cooling is cubic arrangement concerning the cooling effect and cost, while hexagonal structure is optimal when focus on the space utilization of battery module. Besides, the optimized inter-cell distance in battery module structure has been recommended

  20. Design and analysis of multicavity prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Goodpasture, D.W.; Burdette, E.G.; Callahan, J.P.

    1977-01-01

    During the past 25 years, a rather rapid evolution has taken place in the design and use of prestressed concrete reactor vessels (PCRVs). Initially the concrete vessel served as a one-to-one replacement for its steel counterpart. This was followed by the development of the integral design which led eventually to the more recent multicavity vessel concept. Although this evolution has seen problems in construction and operation, a state-of-the-art review which was recently conducted by the Oak Ridge National Laboratory indicated that the PCRV has proven to be a satisfactory and inherently safe type of vessel for containment of gas-cooled reactors from a purely functional standpoint. However, functionalism is not the only consideration in a demanding and highly competitive industry. A summary is presented of the important considerations in the design and analysis of multicavity PCRVs together with overall conclusions concerning the state of the art of these vessels

  1. Design and preliminary analysis of in-vessel core catcher made of high-temperature ceramics material in PWR

    International Nuclear Information System (INIS)

    Xu Hong; Ma Li; Wang Junrong; Zhou Zhiwei

    2011-01-01

    In order to protect the interior wall of pressure vessel from melting, as an additional way to external reactor vessel cooling (ERVC), a kind of in-vessel core catcher (IVCC) made of high-temperature ceramics material was designed. Through the high-temperature and thermal-resistance characteristic of IVCC, the distributing of heat flux was optimized. The results show that the downward average heat flux from melt in ceramic layer reduces obviously and the interior wall of pressure vessel doesn't melt, keeping its integrity perfectly. Increasing of upward heat flux from metallic layer makes the upper plenum structure's temperature ascend, but the temperature doesn't exceed its melting point. In conclusion, the results indicate the potential feasibility of IVCC made of high-temperature ceramics material. (authors)

  2. Emergency core cooling system

    International Nuclear Information System (INIS)

    Ando, Masaki.

    1987-01-01

    Purpose: To actuate an automatic pressure down system (ADS) and a low pressure emergency core cooling system (ECCS) upon water level reduction of a nuclear reactor other than loss of coolant accidents (LOCA). Constitution: ADS in a BWR type reactor is disposed for reducing the pressure in a reactor container thereby enabling coolant injection from a low pressure ECCS upon LOCA. That is, ADS has been actuated by AND signal for a reactor water level low signal and a dry well pressure high signal. In the present invention, ADS can be actuated further also by AND signal of the reactor water level low signal, the high pressure ECCS and not-operation signal of reactor isolation cooling system. In such an emergency core cooling system thus constituted, ADS operates in the same manner as usual upon LOCA and, further, ADS is operated also upon loss of feedwater accident in the reactor pressure vessel in the case where there is a necessity for actuating the low pressure ECCS, although other high pressure ECCS and reactor isolation cooling system are not operated. Accordingly, it is possible to improve the reliability upon reactor core accident and mitigate the operator burden. (Horiuchi, T.)

  3. The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, A., E-mail: alan.peacock@ipp.mpg.de [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Girlinger, A. [MAN Diesel and Turbo SE D-94469 Deggendorf (Germany); Vorkoeper, A. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 17491 Greifswald (Germany); Boscary, J.; Greuner, H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Hurd, F. [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany)

    2011-10-15

    320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m{sup 2}, are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m{sup 2} and a maximum heat load of 200 kW/m{sup 2} with a water velocity of approximately 2 m s{sup -1}. High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m{sup 2} for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

  4. Theoretical investigations on improving performance of cooling systems for fuel cell vehicles; Theoretische Untersuchungen zur Kuehlleistungssteigerung durch innovative Kuehlsysteme fuer Brennstoffzellen-Elektrofahrzeuge

    Energy Technology Data Exchange (ETDEWEB)

    Reichler, Mark

    2008-04-01

    In this work theoretical investigations are carried out for cooling systems, which are used in fuel cell vehicles. This work focuses mainly on the capability of increasing the heat rejection rate by using new alternative cooling systems and by improving the conventional cooling system. Fuel cell vehicles have a higher demand of heat rejection to the ambient than comparable vehicles with combustion engine. The performance of conventional liquid cooling systems, especially at high loads and high ambient temperatures, is often not sufficient anymore. Hence, cooling systems with improved performance are necessary for fuel cell vehicles. The investigations in this work are based on DaimlerChrysler's ''A-Class'' having a PEM-Fuel Cell system integrated. Specific computational models are developed for radiators and condensers to evaluate the performance of different cooling concepts. The models are validated with experimental data. Based on an intensive investigation in the open literature the state of the art of cooling systems for fuel cell vehicles is depicted. Furthermore new cooling concepts as an alternative to the liquid cooling system are presented. The method of cooling the fuel cell by using two-phase transition shows the greatest capability to increase the cooling performance. Hence, this concept is investigated in detail. Two different concepts with three different refrigerants (R113, R245fa und R236fa) are analyzed. Cooling performance of this concept shows improvement of 18.2 up to 32.6 % compared to the conventional liquid cooling system. Thus, a two phase cooling system represents an alternative cooling system for fuel cell vehicles, which should be closer investigated by experiments. (orig.)

  5. Lead-cooled flexible conversion ratio fast reactor

    International Nuclear Information System (INIS)

    Nikiforova, Anna; Hejzlar, Pavel; Todreas, Neil E.

    2009-01-01

    Lead-cooled reactor systems capable of accepting either zero or unity conversion ratio cores depending on the need to burn actinides or operate in a sustained cycle are presented. This flexible conversion ratio reactor is a pool-type 2400 MWt reactor coupled to four 600 MWt supercritical CO 2 (S-CO 2 ) power conversion system (PCS) trains through intermediate heat exchangers. The cores which achieve a power density of 112 kW/l adopt transuranic metallic fuel and reactivity feedbacks to achieve inherent shutdown in anticipated transients without scram, and lead coolant in a pool vessel arrangement. Decay heat removal is accomplished using a reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary auxiliary cooling system (PSACS). The transient simulation of station blackout (SBO) using the RELAP5-3D/ATHENA code shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and a minimum (two) number of PSACS trains. The design of the passive safety systems also prevents coolant freezing in case all four of the PSACS trains are in operation. Both cores are also shown able to accommodate unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents using the S-CO 2 PCS.

  6. The outline of investigation on integrity of JMTR concrete structures, cooling system and utility facilities

    International Nuclear Information System (INIS)

    Ebisawa, Hiroyuki; Hanakawa, Hiroki; Asano, Norikazu; Kusunoki, Hidehiko; Yanai, Tomohiro; Sato, Shinichi; Miyauchi, Masaru; Ohto, Tsutomu; Kimura, Tadashi; Kawamata, Takanori; Nemoto, Nobuaki; Watahiki, Shunsuke; Hanawa, Yoshio; Tsuboi, Kazuaki; Ogasawara, Yasushi; Nemoto, Hiroyoshi; Echigoya, Shinichi; Ohtsuka, Kaoru; Onoue, Ryuji; Koike, Sumio; Gorai, Shigeru; Nishiyama, Yutaka; Kurosawa, Akihiko; Hanawa, Nobuhiro; Tobita, Kenji; Tabata, Shuzo; Fukasaku, Akitomi; Isozaki, Takanori; Akashi, Kazutomo; Takahashi, Kunihiro; Tsuji, Tomoyuki

    2009-07-01

    The condition of facilities and machinery used continuously were investigated before the renewal work of JMTR on FY 2007. The subjects of investigation were reactor building, primary cooling system tanks, secondary cooling system piping and tower, emergency generator and so on. As the result, it was confirmed that some facilities and machinery were necessary to repair and others were used continuously for long term by maintaining on the long-term maintenance plan. JMTR is planed to renew by the result of this investigation. (author)

  7. Research to sustain cases for Magnox-reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Graham, W.J.

    1997-01-01

    Britain's Magnox Electric plc owns and operates six power stations, each of which has twin gas-cooled reactors of the Magnox-fuel type. The older group of four power stations has steel pressure-circuits. The reactor cores are housed within spherical, steel vessels. This article describes some of the research which is undertaken to sustain the safety cases for these steel vessels which have now been in operation for just over 30 years. (author) 2 figs., 4 refs

  8. Converging coolness and investigating its relation to user experience

    DEFF Research Database (Denmark)

    Raptis, Dimitrios; Bruun, Anders; Kjeldskov, Jesper

    2017-01-01

    Recently a number of studies appeared that operationalised coolness and explored its relation to digital products. Literature suggests that perceived coolness is another factor of user experience, and this adds to an existing explosion of dimensions related to aesthetics, hedonic quality, pragmatic...... quality, attractiveness, etc. A critical challenge highlighted in prior research is to study the relationships among those factors and so far, no studies have empirically examined the relationship between coolness and other established user experience factors. In this paper, we address this challenge...... cool and UX factors converge into 5 for the case of mobile devices. Our findings are important for researchers, as we demonstrate through a validated model that coolness is part of UX research, as well as for practitioners, by developing a questionnaire that can reliably measure both perceived inner...

  9. Modeling the Thermal Mechanical Behavior of a 300 K Vacuum Vessel that is Cooled by Liquid Hydrogen in Film Boiling

    International Nuclear Information System (INIS)

    Yang, S.Q.; Green, M.A.; Lau, W.

    2004-01-01

    This report discusses the results from the rupture of a thin window that is part of a 20-liter liquid hydrogen vessel. This rupture will spill liquid hydrogen onto the walls and bottom of a 300 K cylindrical vacuum vessel. The spilled hydrogen goes into film boiling, which removes the thermal energy from the vacuum vessel wall. This report analyzes the transient heat transfer in the vessel and calculates the thermal deflection and stress that will result from the boiling liquid in contact with the vessel walls. This analysis was applied to aluminum and stainless steel vessels

  10. Investigation on heat transfer enhancement and pressure loss of double swirl chambers cooling

    Directory of Open Access Journals (Sweden)

    Gang Lin

    2013-09-01

    Full Text Available By merging two standard swirl chambers, an alternative cooling configuration named double swirl chambers (DSC has been developed. In the DSC cooling configuration, the main physical phenomena of the swirl flow in swirl chamber and the advantages of swirl flow in heat transfer augmentation are maintained. Additionally, three new physical phenomena can be found in DSC cooling configuration, which result in a further improvement of the heat transfer: (1 impingement effect has been observed, (2 internal heat exchange has been enhanced between fluids in two swirls, and (3 “∞” shape swirl has been generated because of cross effect between two chambers, which improves the mixing of the fluids. Because of all these improvements, the DSC cooling configuration leads to a higher globally-averaged thermal performance parameter (Nu¯¯/Nu∞/(f/f01/3 than standard swirl chamber. In particular, at the inlet region, the augmentation of the heat transfer is nearly 7.5 times larger than the fully developed non-swirl turbulent flow and the circumferentially averaged Nusselt number coefficient is 41% larger than the standard swirl chamber. Within the present work, a further investigation on the DSC cooling configuration has been focused on the influence of geometry parameters e.g. merging ratio of chambers and aspect ratio of inlet duct on the cooling performance. The results show a very large influence of these geometry parameters in heat transfer enhancement and pressure drop ratio. Compared with the basic configuration of DSC cooling, the improved configuration with 20% to 23% merging ratio shows the highest globally-averaged thermal performance parameter. With the same cross section area in tangential inlet ducts, the DSC cooling channel with larger aspect ratio shows larger heat transfer enhancement and at the same time reduced pressure drop ratio, which results in a better globally-averaged thermal performance parameter.

  11. A novel personal cooling system (PCS) incorporated with phase change materials (PCMs) and ventilation fans: An investigation on its cooling efficiency.

    Science.gov (United States)

    Lu, Yehu; Wei, Fanru; Lai, Dandan; Shi, Wen; Wang, Faming; Gao, Chuansi; Song, Guowen

    2015-08-01

    Personal cooling systems (PCS) have been developed to mitigate the impact of severe heat stress for humans working in hot environments. It is still a great challenge to develop PCSs that are portable, inexpensive, and effective. We studied the performance of a new hybrid PCS incorporating both ventilation fans and phase change materials (PCMs). The cooling efficiency of the newly developed PCS was investigated on a sweating manikin in two hot conditions: hot humid (HH, 34°C, 75% RH) and hot dry (HD, 34°C, 28% RH). Four test scenarios were selected: fans off with no PCMs (i.e., Fan-off, the CONTROL), fans on with no PCMs (i.e., Fan-on), fans off with fully solidified PCMs (i.e., PCM+Fan-off), and fans on with fully solidified PCMs (i.e., PCM+Fan-on). It was found that the addition of PCMs provided a 54∼78min cooling in HH condition. In contrast, the PCMs only offered a 19-39min cooling in HD condition. In both conditions, the ventilation fans greatly enhanced the evaporative heat loss compared with Fan-off. The hybrid PCS (i.e., PCM+Fan-on) provided a continuous cooling effect during the three-hour test and the average cooling rate for the whole body was around 111 and 315W in HH and HD conditions, respectively. Overall, the new hybrid PCS may be an effective means of ameliorating symptoms of heat stress in both hot-humid and hot-dry environments. Copyright © 2015 Elsevier Ltd. All rights reserved.

  12. In-vessel retention modeling capabilities in MAAP5

    International Nuclear Information System (INIS)

    Paik, Chan Y.; Lee, Sung Jin; Zhou, Quan; Luangdilok, W.; Reeves, R.W.; Henry, R.E.; Plys, M.; Scobel, J.H.

    2012-01-01

    Modular Accident Analysis Program (MAAP) is an integrated severe accident analysis code for both light water and heavy water reactors. New and improved models to address the complex phenomena associated with in-vessel retention (IVR) were incorporated into MAAP5.01. They include: -a) time-dependent volatile and non-volatile decay heat, -b) material properties at high temperatures, -c) finer vessel wall nodalization, -d) new correlations for natural convection heat transfer in the oxidic pool, -e) refined metal layer heat transfer to the reactor vessel wall and surroundings, -f) formation of a heavy metal layer, and -g) insulation cooling channel model and associated ex-vessel heat transfer and critical heat flux correlations. In this paper, the new and improved models in MAAP5.01 are described and sample calculation results are presented for the AP1000 passive plant. For the IVR evaluation, a transient calculation is useful because the timing of corium relocation, decaying heat load, and formation of separate layers in the lower plenum all affect integrity of the lower head. The key parameters affecting the IVR success are the metal layer emissivity and thickness of the top metal layer, which depends on the amount of steel in the oxidic pool and in the heavy metal layer. With the best estimate inputs for the debris mixing parameters in a conservative IVR scenario, the AP1000 plant results show that the maximum ex-vessel heat flux to CHF ratio is about 0.7, which occurs before 10.000 seconds when the decay heat is high. The AP1000 plant results demonstrate how MAAP5.01 can be used to evaluate IVR and to gain insight into responses of the lower head during a severe accident

  13. Gas cooled HTR

    International Nuclear Information System (INIS)

    Schweiger, F.

    1985-01-01

    In the He-cooled, graphite-moderated HTR with spherical fuel elements, the steam generator is fixed outside the pressure vessel. The heat exchangers are above the reactor level. The hot gases stream from the reactor bottom over the heat exchanger, through an annular space around the heat exchanger and through feed lines in the side reflector of the reactor back to its top part. This way, in case of shutdown there is a supplementary natural draught that helps the inner natural circulation (chimney draught effect). (orig./PW)

  14. Gas-liquid flow filed in agitated vessels

    International Nuclear Information System (INIS)

    Hormazi, F.; Alaie, M.; Dabir, B.; Ashjaie, M.

    2001-01-01

    Agitated vessels in form of sti reed tank reactors and mixed ferment ors are being used in large numbers of industry. It is more important to develop good, and theoretically sound models for scaling up and design of agitated vessels. In this article, two phase flow (gas-liquid) in a agitated vessel has been investigated numerically. A two-dimensional computational fluid dynamics model, is used to predict the gas-liquid flow. The effects of gas phase, varying gas flow rates and variation of bubbles shape on flow filed of liquid phase are investigated. The numerical results are verified against the experimental data

  15. An introduction to the analysis of multi-cavity prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Silva, M.C.A.T. da.

    1986-01-01

    The present work is a study of multi-cavity prestressed concrete pressure vessels (PCRV) for nuclear reactors. A review is made of the designs, analises and models of multi-cavity concrete pressure vessels. A preliminary evaluation of the NONSAP program for applications in complex three-dimensional structures such as a multi-cavity pressure vessel is also made. A model of a PCRV of a 1000 MW(e) high-temperature gas cooled reactor was selected for a three-dimensional analysis with the NONSAP program. The results obtained are compared with experimental data. (Author) [pt

  16. Large-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

    Chu, T.Y.; Slezak, S.E.; Bentz, J.H.; Pasedag, W.F.

    1994-01-01

    This paper presents results of ex-vessel boiling experiments performed in the CYBL (CYlindrical BoiLing) facility. CYBL is a reactor-scale facility for confirmatory research of the flooded cavity concept for accident management. CYBL has a tank-within-a-tank design; the inner tank simulates the reactor vessel and the outer tank simulates the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm 2 across the vessel bottom were performed. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that under prototypic heat load and heat flux distributions, the flooded cavity in a passive pressurized water reactor like the AP-600 should be capable of cooling the reactor pressure vessel in the central region of the lower head that is addressed by these tests

  17. Safety technology qualification of the prestressed cast iron pressure vessel (PCIV) and of the primary cell of the HTR-modul for the passive removal of decay heat, phase 1 (INHR)

    International Nuclear Information System (INIS)

    Warnke, E.P.

    1990-02-01

    During this development program the thermodynamic behaviour of a system was investigated, consisting of a hot working Prestressed Cast Iron Pressure Vessel and an inactive heat sink in the surrounding cavern cell. It could be shown, that the inactive heat removal system designed as a natural circuit can remove the maximum amount of heat of 890 kW during emergency conditions via a natural-draught air cooling tower even under very conservative assumptions and for a 50% loss of cooling pipes. Further it could be shown, that the hot working Prestressed Cast Iron Pressure Vessel has a very safe load carrying behaviour during all normal and upset conditions. (orig.) With 10 tabs., 38 figs., 43 refs [de

  18. The effect of water vapor in the reactor cavity in a MHTGR [Modular High Temperature Gas Cooled Reactor] on the radiation heat transfer

    International Nuclear Information System (INIS)

    Cappiello, M.W.

    1991-01-01

    Analyses have been completed to determine the effect of the presence of water vapor in the reactor cavity in a modular high temperature gas cooled reactor on the predicted radiation heat transfer from the vessel wall to the reactor cavity cooling system. The analysis involves the radiation heat transfer between two parallel plates with an absorbing and emitting medium present. Because the absorption in the water vapor is spectrally dependent, the solution is difficult even for simple geometries. A computer code was written to solve the problem using the Monte Carlo method. The code was validated against closed form solutions, and shows excellent agreement. In the analysis of the reactor problem, the results show that the reduction in heat transfer, and the consequent increase in the vessel wall temperature, can be significant. This effect can be cast in terms of a reduction in the wall surface emissivities from 0.8 to 0.59. Because of the insulating effect of the water vapor, increasing the gap distance between the vessel wall and the cooling system will cause the vessel wall temperature to increase further. Care should be taken in the design of the facility to minimize the gap distance and keep temperature increase within allowable limits. 3 refs., 6 figs., 4 tabs

  19. Liquid metal cooled nuclear reactor constructions

    International Nuclear Information System (INIS)

    Aspden, G.J.; Allbeson, K.F.

    1984-01-01

    In a liquid metal cooled nuclear reactor with a nuclear fuel assembly in a coolant-containing primary vessel housed within a concrete containment vault, there is thermal insulation to protect the concrete, the insulation being disposed between vessel and concrete and being hung from metal structure secured to and projecting from the concrete, the insulation consisting of a plurality of adjoining units each unit incorporating a pack of thermal insulating material and defining a contained void co-extensive with said pack and situated between pack and concrete, the void of each unit being connected to the voids of adjoining units so as to form continuous ducting for a fluid coolant. (author)

  20. An investigation into the suitability of additive manufacturing techniques for neutron moderator vessels

    International Nuclear Information System (INIS)

    Gallimore, S.

    2016-01-01

    Additive manufacturing (also known as rapid prototyping or 3D printing) techniques are increasing in popularity for several key reasons; greater freedom in possible geometry, reduced time of manufacture and connected to these are potential cost savings. ISIS has begun an investigation into the suitability of the various available techniques for the manufacture of neutron moderator vessels, in order to see if it can exploit these advantages. It is however understood that additive manufacturing is by no means a perfect technique and part of the investigations will be to try and better understand how some of the disadvantages of the technique affect its potential application within the spallation neutron environment. Some of the main disadvantages commonly listed are; the grades of materials available/suitable for the process are limited, virtually no pre-existing material data from radiation environments, lower quality surface finish (directly from the manufacturing process), less familiarity with residual stresses in the material and questions over whether tight tolerances and consistent material thicknesses be achieved? The work has been divided into two streams; one which utilises small samples to evaluate and compare different manufacturing and post-treatment techniques, the other that performs tests on a full-size representative moderator vessel. The complete programme of testing shall include the following tests; fundamental 'neutronic transparency', room temperature vacuum leak test, cold shock (using LN_2) and subsequent room temperature leak test, pressure cycling, a burst test, welding suitability and material data testing. The investigations being conducted at ISIS are very much in the early stages and looking at fairly fundamental questions. Answering these will clearly guide the decision whether is it worth continuing with further investigation and development or if the currently available techniques do not produce materials that are suitable for use as

  1. Investigation of the design of a metal-lined fully wrapped composite vessel under high internal pressure

    Science.gov (United States)

    Kalaycıoğlu, Barış; Husnu Dirikolu, M.

    2010-09-01

    In this study, a Type III composite pressure vessel (ISO 11439:2000) loaded with high internal pressure is investigated in terms of the effect of the orientation of the element coordinate system while simulating the continuous variation of the fibre angle, the effect of symmetric and non-symmetric composite wall stacking sequences, and lastly, a stacking sequence evaluation for reducing the cylindrical section-end cap transition region stress concentration. The research was performed using an Ansys® model with 2.9 l volume, 6061 T6 aluminium liner/Kevlar® 49-Epoxy vessel material, and a service internal pressure loading of 22 MPa. The results show that symmetric stacking sequences give higher burst pressures by up to 15%. Stacking sequence evaluations provided a further 7% pressure-carrying capacity as well as reduced stress concentration in the transition region. Finally, the Type III vessel under consideration provides a 45% lighter construction as compared with an all metal (Type I) vessel.

  2. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

  3. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  4. Simulant melt experiments on performance of the in-vessel core catcher

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Sang-Baik; Suh, K.Y.; Cheung, F.B.; Rempe, J.L.

    2007-01-01

    In order to enhance the feasibility of in-vessel retention (IVR) of molten core material during a severe accident for high-power reactors, an in-vessel core catcher (IVCC) was designed and evaluated as part of a joint United States-Korean International Nuclear Energy Research Initiative (INERI). The proposed IVCC is expected to increase the thermal margin for success of IVR by providing an 'engineered gap' for heat transfer from materials that relocate during a severe accident and potentially serving as a sacrificial material under a severe accident. In this study, LAVA-GAP experiments were performed to investigate the thermal and mechanical performance of the IVCC using the alumina melt as simulant. The LAVA-GAP experiments aim to examine the feasibility and sustainability of the IVCC under the various test conditions using 1/8th scale hemispherical test sections. As a feasibility test of the proposed IVCC in this INERI project, the effects of IVCC base steel materials, internal coating materials, and gap size between the IVCC and the vessel lower head were examined. The test results indicated that the internally coated IVCC has high thermal performance compared with the uncoated IVCC. In terms of integrity of the base steel, carbon steel is superior to stainless steel and the effect of bond coat is found to be trivial for the tests performed in this study. The thermal load is mitigated via boiling heat removal in the gap between the IVCC and the vessel lower head. The current test results imply that gaps less than 10 mm are not enough to guarantee effective cooling induced by water ingression and steam venting there through. Selection of endurable material and pertinent gap size is needed to implement the proposed IVCC concept into advanced reactor designs

  5. Development of the Simulation Program for the In-Vessel Fuel Handling System of Double Rotating Plug Type

    International Nuclear Information System (INIS)

    Kim, S. H.; Kim, J. B.

    2011-01-01

    In-vessel fuel handling machines are the main equipment of the in-vessel fuel handling system, which can move the core assembly inside the reactor vessel along with the rotating plug during refueling. The in vessel fuel handling machines for an advanced sodium cooled fast reactor(SFR) demonstration plant are composed of a direct lift machine(DM) and a fixed arm machine(FM). These machines should be able to access all areas above the reactor core by means of the rotating combination of double rotating plugs. Thus, in the in vessel fuel handling system of the double rotating plug type, it is necessary to decide the rotating plug size and evaluate the accessibility of in-vessel fuel handling machines in given core configuration. In this study, the simulation program based on LABVIEW which can effectively perform the arrangement design of the in vessel fuel handling system and simulate the rotating plug motion was developed. Fig. 1 shows the flow chart of the simulation program

  6. Design and implementation of visual inspection system handed in tokamak flexible in-vessel robot

    International Nuclear Information System (INIS)

    Wang, Hesheng; Xu, Lifei; Chen, Weidong

    2016-01-01

    In-vessel viewing system (IVVS) is a fundamental tool among the remote handling systems for ITER, which is used to providing information on the status of the in-vessel components. The basic functional requirement of in-vessel visual inspection system is to perform a fast intervention with adequate optical resolution. In this paper, we present the software and hardware solution, which is designed and implemented for tokamak in-vessel viewing system that installed on end-effector of flexible in-vessel robot working under vacuum and high temperature. The characteristic of our in-vessel viewing system consists of two parts: binocular heterogeneous vision inspection tool and first wall scene emersion based augment virtuality. The former protected with water-cooled shield is designed to satisfy the basic functional requirement of visual inspection system, which has the capacity of large field of view and high-resolution for detection precision. The latter, achieved by overlaying first wall tiles images onto virtual first wall scene model in 3D virtual reality simulation system, is designed for convenient, intuitive and realistic-looking visual inspection instead of viewing the status of first wall only by real-time monitoring or off-line images sequences. We present the modular division of system, each of them in smaller detail, and go through some of the design choices according to requirements of in-vessel visual inspection task.

  7. Design and implementation of visual inspection system handed in tokamak flexible in-vessel robot

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hesheng; Xu, Lifei [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China)

    2016-05-15

    In-vessel viewing system (IVVS) is a fundamental tool among the remote handling systems for ITER, which is used to providing information on the status of the in-vessel components. The basic functional requirement of in-vessel visual inspection system is to perform a fast intervention with adequate optical resolution. In this paper, we present the software and hardware solution, which is designed and implemented for tokamak in-vessel viewing system that installed on end-effector of flexible in-vessel robot working under vacuum and high temperature. The characteristic of our in-vessel viewing system consists of two parts: binocular heterogeneous vision inspection tool and first wall scene emersion based augment virtuality. The former protected with water-cooled shield is designed to satisfy the basic functional requirement of visual inspection system, which has the capacity of large field of view and high-resolution for detection precision. The latter, achieved by overlaying first wall tiles images onto virtual first wall scene model in 3D virtual reality simulation system, is designed for convenient, intuitive and realistic-looking visual inspection instead of viewing the status of first wall only by real-time monitoring or off-line images sequences. We present the modular division of system, each of them in smaller detail, and go through some of the design choices according to requirements of in-vessel visual inspection task.

  8. Design, fabrication and operating experience of Monju ex-vessel fuel storage tank

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Yamagishi, Yoshiaki; Kuroha, Mitsuo; Inoue, Tatsuya

    1995-01-01

    In FBRs there are two methods of storing and cooling the spent fuel - the in-vessel storage and the ex-vessel storage. Because of the sodium leaks through the tank at the beginning of pre-operation, the utilization of the ex-vessel fuel storage tank (EVST) of some FBR plant has been changed from the ex-vessel fuel storage to the interim fuel transfer tank. This led to reactor designers focusing on the material, structure and fabrication of the carbon steel sodium storage tanks worldwide. The Monju EVST was at the final stage of the design, when the leaks occurred. The lesson learned from that experience and the domestic fabrication technology are reflected to the design and fabrication of the Monju EVST. This paper describes the design, fabrication and R and D results for the tank, and operating experience in functional test. The items to be examined are as follows: (1) Overall structure of the tank and design philosophy on the function, (2) Structure of the cover shielding plug and its design philosophy, (3) Structures of the rotating rack and its bearings, and their design philosophy, (4) Cooling method and its design philosophy, (5) Structure and fabrication of the cooling coil support inside EVST with comparison of leaked case, (6) R and D effort for items above. The fabrication of the Monju EVST started in August 1986 and it was shipped to the site in March 1990. Installation was completed in November 1990, and sodium fill after pre-heating started in 1991. The operation has been continued since September 1992. In 1996 when the first spent fuel is stored, its total functions will be examined. (author)

  9. Experimental and theoretical investigation on the depressurization of a vessel with internals

    International Nuclear Information System (INIS)

    Vigni, P.; Oriolo, F.; Rosa, U.

    1978-01-01

    This paper is about some blow-down experiments performed at the Scalbatraio Center of the University of Pisa. The blow-down tests have been made to investigate the depressurization of a vessel with internal structures, reproducing the geometry of a BWR. The experimental data have been compared with calculations performed by the RELAP program, in order to evaluate the scaling effects related to their application to large scale units. (author)

  10. Plants for passive cooling. A preliminary investigation of the use of plants for passive cooling in temperate humid climates

    Energy Technology Data Exchange (ETDEWEB)

    Spirn, A W; Santos, A N; Johnson, D A; Harder, L B; Rios, M W

    1981-04-01

    The potential of vegetation for cooling small, detached residential and commercial structures in temperate, humid climates is discussed. The results of the research are documented, a critical review of the literature is given, and a brief review of energy transfer processes is presented. A checklist of design objectives for passive cooling, a demonstration of design applications, and a palette of selected plant species suitable for passive cooling are included.

  11. Transient temperature and stress distributions in the pressure vessel's wall of a nuclear reactor

    International Nuclear Information System (INIS)

    Silva, G.A. da

    1979-01-01

    In order to calculate the temperature distribution in a reactor vessel wall which is under the effect of gamma radiation originated in the reactor core, a numerical solution is proposed. This problem may arise from a reactor cooling pump failure .The thermal stresses are also calculated. (Author) [pt

  12. Reactor pressure vessel behaviour with a small crack in the cladding

    International Nuclear Information System (INIS)

    Fayolle, P.; Churier-Bossennec, H.; Faidy, C.

    1990-01-01

    This paper reports on fracture mechanic analysis of a PWR reactor pressure vessel with a 3.5 mm embedded circumferential crack in the cladding under a small lost of cooling accident transient. Different RTNDT level and effect of irradiation on material properties are considered. The study compares simplified one-dimensional and two-dimensional elastic approach and complete elastoplastic approach using J-parameter. The results show: good correlation between the different elastic approaches, important conservatism of the elastic approach compared to elastoplastic approach, no influence of irradiated material properties. The behavior of a vessel with this type of crack is acceptable for RTNDT less than 135 deg and safety injection temperature of 60 deg

  13. Investigation of aluminium state in some popular food, which are cooked in aluminium vessels, using spectroscopic analysis methods

    International Nuclear Information System (INIS)

    Al-Shahneh, M.; Saheune, Kh.

    2009-01-01

    Aluminium and lead elements were determined in drinking water and salt solution from chick-pea and faba-bean cooked in aluminium vessels and others from teflon for comparison using atomic absorption spectroscopy by graphite furnace. The relationship between heating time and aluminium quantities transferred to these food solutions was investigated. The lead element was determined taking into consideration the fact that this element may enter in these vessels somehow during the manufacturing process. Results show that the highest value of aluminium quantities was recorded in salt solution ( 17.022 μg/ml) without heating , followed by chick-pea solution (9.95 μg/ml), then faba-bean solution (2.81 μg/ml) when the heated period was 120 minutes. (author)

  14. Reactor Pressure Vessel (RPV) Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2008-04-01

    The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

  15. Nuclear power station with a water-cooled reactor pressure vessel

    International Nuclear Information System (INIS)

    Hoffmann, R.; Brunner, G.; Jost, N.

    1987-01-01

    Nuclear radiation produces radiolysis gases, which are undesirable for corrosion and oxyhydrogen gas reasons. To limit the proportion of this radiolysis gas, the invention provides that catalytic surfaces should be introduced into the primary circuit, to produce recombination of hydrogen and oxygen. These surfaces can be accommodated in the upper part of the reactor pressure vessel. The live steam screen can also have a catalytic surface. (orig./HP) [de

  16. Study on enhancement of heat transfer of reactor vessel auxiliary cooling system of fast breeder reactor

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Kinoshita, Izumi; Ueda, Nobuyuki; Furuya, Masahiro

    1996-01-01

    A reactor vessel auxiliary cooling system (RVACS), which is one of the decay heat removal systems of the fast breeder reactor (FBR), has passive safety as well as high reliability. However, the heat removal capability is relatively small, because its heat exchange is dependent on the natural convection of the air. The objectives of this report are to propose a heat transfer medium to enhance the heat transfer and to confirm the heat transfer performance of this system by experimental and analytical studies. From these studies, the following main results were obtained. (1) A porous plate with 5 mm thickness, 5 mm pore diameter, 92% porosity, was found to have the highest enhancement of heat transfer. (2) The heat transfer enhancement was demonstrated by large scale heat transfer experiments. Also, the heat transfer correlations, which can be used in the plant transient analyses, were derived from the experimental results. (3) Analysing the transient conditions of conventional pool-type FBR by means of the system analysis code, the applicable range of this system was assumed from the capability of the RVACS with porous plates. As a result, this type of RVACS was found to be applicable to conventional pool-type FBRs with capacity of about 500 MWe or less. (author)

  17. Investigation of residual stresses in thick-walled vessels with combination of autofrettage and wire-winding

    International Nuclear Information System (INIS)

    Sedighi, M.; Jabbari, A.H.

    2013-01-01

    Wire-winding and autofrettage processes can be used to introduce beneficial residual stress in the cylinder of thick-walled pressure vessels. In both techniques, internal residual compressive stress will increase internal pressure capacity, improve fatigue life and reduce fatigue crack initiation. The purpose of this paper is to analyze the effects of wire-winding on an autofrettaged thick-walled vessel. Direct method which is a modified Variable Material Properties (VMP) method has been used in order to calculate residual stresses in an autofrettaged vessel. Since wire-winding is done after autofrettage process, the tangent and/or Young's modulus could be changed. For this reason, a new wire-winding method based on Direct Method is introduced. The obtained results for wire-wound autofrettaged vessels are validated by finite element method. The results show that by using this approach, the residual hoop stresses in a wire-wound autofrettaged vessel have a more desirable distribution in the cylinder. -- Highlights: • Combination of autofrettage and wire-winding in pressure vessels has been presented. • A new method based on Direct method is presented for wire-winding process. • Residual hoop stresses are compared in vessels cylinders for different cases. • The residual hoop stress has a more desirable stress distribution. • The benefits of the combined vessel are highlighted in comparison with single cases

  18. Safety design/analysis and scenario for prevention of CDA with ECCS in lead-bismuth-cooled fast reactor

    International Nuclear Information System (INIS)

    Minoru, Takahashi; Vaclav, Dostal; Abu Khalid, Rivai; Novitrian; Yumi, Yamada

    2007-01-01

    Safety design has been developed to show safety feature of Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR). The core is designed to have negative void reactivity even if the entire core and upper plenum are voided by steam intrusion from above. In-vessel type control rod driving mechanisms are used to prevent control rods from accidental ejection due to high pressure in the reactor vessel. In cases of coolant leakage from reactor vessel and feed water pipes, Pb-Bi coolant level in the reactor vessel is kept at the required level for decay heat removal by means of closed type guard vessel. Dual pipes are adopted to avoid leak of water in the feedwater system. Pump trip in feedwater systems initiates loss of coolant flow (LOF) event, although there is no concern of loss of flow accident due to primary pump trip. Injection of high pressure water slows down the flow-coast-down of feedwater at the LOF event. It has been evaluated that the fuel temperature is kept lower than safety limits at the unprotected loss of flow and heat sink (ATWS). A scenario for prevention of the core disruptive accident (CDA) with the emergency core cooling system (ECCS) is examined. The reactor becomes super-critical when the reactor vessel is filled with water. It is necessary to use water with boric acid for the ECC system, and additional backup rods for sub-critical core in water injection. (authors)

  19. General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel

    International Nuclear Information System (INIS)

    Roche, R.; Gaudez, J.C.

    1964-01-01

    In the framework of research carried out on a CO 2 -cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [fr

  20. Experimental investigation of gas turbine airfoil aerodynamic performance without and with film cooling in an annular sector cascade

    Energy Technology Data Exchange (ETDEWEB)

    Wiers, S.H.

    2002-02-01

    The steady growing of industrialization, the densification of the anthroposphere, the increasing concern over the effects of gas turbine cruise emissions on the atmosphere threaten the growth of air transportation, and the perception about the possible climatic impact of CO{sub 2} emissions causes a public distinctive sense of responsibility. The conventional energy production techniques, which are based on fossil fuel, will keep its central importance within the global energy production. Forecasts about the increasing air transportation give duplication in the next 10-15 years. The optimization of the specific fuel consumption is necessary to decrease the running costs and the pollution emissions in the atmosphere, which makes an increased process efficiency of stationary turbines as well as of jet engines essential. This leads to the necessity of an increased thermodynamic efficiency of the overall process and the optimization of the aerodynamic components. Due to the necessity of more detailed three-dimensional data on the behavior of film cooled blades an annular sector cascade turbine test facility has gone into service. The annular sector cascade facility is a relative cost efficient solution compared to a full annular facility to investigate three-dimensional effects on a non cooled and cooled turbine blade. The aerodynamic investigations on the annular sector cascade facility are part of a broad perspective where experimental data from a hot annular sector cascade facility and the cold annular sector facility are used to verify, calibrate and understand the physics for both internal and external calculation methods for flow and heat transfer prediction. The objective of the present study is the design and validation of a cold flow annular sector cascade facility, which meets the flow conditions in a modem turbine as close as possible, with emphasis on achieving periodic flow conditions. The first part of this study gives the necessary background on this

  1. Investigation of neutron irradiated reactor vessel steels using post-irradiation annealing techniques

    Energy Technology Data Exchange (ETDEWEB)

    Nakata, Hayato; Fukuya, Koji [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    The matrix damage is known to be a major factor that contributes to embrittlement and hardening of irradiated reactor vessel steels, and is assumed to be composed of the point defect clusters. However field emission gun scanning transmission electron microscopy (FEGSTEM) and atom probe (AP) could not detect any evidence of the matrix damage. In this study, post irradiation annealing experiments combining positron annihilation lineshape analysis (PALA) and hardness experiments were applied to an actual surveillance test specimen and a sample of reactor vessel steel irradiated in a material test reactor (MTR), in order to investigate the matrix damage recovery behavior and its contribution to hardening. It was confirmed that higher fluence increased the hardness and the volume fraction of open volume defects and that higher flux decreased the thermal stability of matrix damage and the effect on hardening. The contribution of matrix damage to hardening could be estimated to be below 30%. (author)

  2. Experimental investigation of cooling performance of a novel HVAC system combining natural ventilation with diffuse ceiling inlet and TABS

    DEFF Research Database (Denmark)

    Yu, Tao; Heiselberg, Per Kvols; Lei, Bo

    2015-01-01

    Highlights •An experimental investigation of cooling performance of a combined HVAC system is carried out. •Cooling performance of TABS with and without the influence of diffuse ceiling is analyzed. •Radiant and convective heat transfer coefficients of TABS cooling are studied. •Cooling components...

  3. Performance investigation of solid desiccant evaporative cooling system configurations in different climatic zones

    International Nuclear Information System (INIS)

    Ali, Muzaffar; Vukovic, Vladimir; Sheikh, Nadeem Ahmed; Ali, Hafiz M.

    2015-01-01

    Highlights: • Five configurations of a DEC system are analyzed in five climate zones. • DEC system model configurations are developed in Dymola/Modelica. • Performance analysis predicted a suitable DEC system configuration for each climate zone. • Results show that climate of Vienna, Sao Paulo, and Adelaide favors the ventilated-dunkle cycle. • While ventilation cycle configuration suits the climate of Karachi and Shanghai. - Abstract: Performance of desiccant evaporative cooling (DEC) system configurations is strongly influenced by the climate conditions and varies widely in different climate zones. Finding the optimal configuration of DEC systems for a specific climatic zone is tedious and time consuming. This investigation conducts performance analysis of five DEC system configurations under climatic conditions of five cities from different zones: Vienna, Karachi, Sao Paulo, Shanghai, and Adelaide. On the basis of operating cycle, three standard and two modified system configurations (ventilation, recirculation, dunkle cycles; ventilated-recirculation and ventilated-dunkle cycles) are analyzed in these five climate zones. Using an advance equation-based object-oriented (EOO) modeling and simulation approach, optimal configurations of a DEC system are determined for each climate zone. Based on the hourly climate data of each zone for its respective design cooling day, performance of each system configuration is estimated using three performance parameters: cooling capacity, COP, and cooling energy delivered. The results revealed that the continental/micro-thermal climate of Vienna, temperate/mesothermal climate of Sao Paulo, and dry-summer subtropical climate of Adelaide favor the use of ventilated-dunkle cycle configuration with average COP of 0.405, 0.89 and 1.01 respectively. While ventilation cycle based DEC configuration suits arid and semiarid climate of Karachi and another category of temperate/mesothermal climate of Shanghai with average COP of

  4. Engineering analysis of ITER In-Vessel Viewing System guide tube

    Energy Technology Data Exchange (ETDEWEB)

    Casal, Natalia, E-mail: natalia.casal@iter.org [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Bates, Philip [Fusion for Energy, Barcelona (Spain); Bede, Ottó [Oxford Technologies Ltd., Abingdon (United Kingdom); Damiani, Carlo; Dubus, Gregory [Fusion for Energy, Barcelona (Spain); Omran, Hassan [Oxford Technologies Ltd., Abingdon (United Kingdom); Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Puiu, Adrian [Fusion for Energy, Barcelona (Spain); Reichle, Roger; Suárez, Alejandro; Walker, Christopher; Walsh, Michael [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France)

    2015-10-15

    Highlights: • Conceptual design of IVVS Loads action on IVVS Dominant loads. • Seismic and baking conditions. • No active cooling needed for IVVS. • IVVS requires additional support points to avoid excessive deformation. - Abstract: The In Vessel Viewing System (IVVS) will be one of the essential machine diagnostic systems at ITER to provide information about the status of in-vessel and plasma facing components and to evaluate the dust inside the Vacuum Vessel. The current design consists of six scanning probes and their deployment systems, which are placed in dedicated ports at the divertor level. These units are located in resident guiding tubes 10 m long, which allow the IVVS probes to go from their storage location to the scanning position by means of a simple straight translation. Moreover, each resident tube is supported inside the corresponding Vacuum Vessel and Cryostat port extensions, which are part of the primary confinement barrier. As the Vacuum Vessel and the Cryostat will move with respect to each other during operation (especially during baking) and during incidents and accidents (disruptions, vertical displacement events, seismic events), the structural integrity of the resident tube and the surrounding vacuum boundaries would be compromised if the required flexibility and supports are not appropriately assured. This paper focuses on the integration of the present design of the IVVS into the Vacuum Vessel and Cryostat environment. It presents the adopted strategy to withstand all the main interfacing loads without damaging the confinement barriers and the corresponding analysis supporting it.

  5. Selected bibliography on pressure vessels for light-water-cooled power reactors (LWRs)

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1975-01-01

    Abstracts on LWR pressure vessels are arranged in the following categories: general, design, materials technology, fabrication techniques, inspection and testing, and failures. Author, keyword, and KWIC (keyword-in-content) indices are provided. (U.S.)

  6. Determination of the failure probability in the weld region of ap-600 vessel for transient condition

    International Nuclear Information System (INIS)

    Wahyono, I.P.

    1997-01-01

    Failure probability in the weld region of AP-600 vessel was determined for transient condition scenario. The type of transient is increase of the heat removal from primary cooling system due to sudden opening of safety valves or steam relief valves on the secondary cooling system or the steam generator. Temperature and pressure in the vessel was considered as the base of deterministic calculation of the stress intensity factor. Calculation of film coefficient of the convective heat transfers is a function of the transient time and water parameter. Pressure, material temperature, flaw depth and transient time are variables for the stress intensity factor. Failure probability consideration was done by using the above information in regard with the flaw and probability distributions of Octavia II and Marshall. Calculation of the failure probability by probability fracture mechanic simulation is applied on the weld region. Failure of the vessel is assumed as a failure of the weld material with one crack which stress intensity factor applied is higher than the critical stress intensity factor. VISA II code (Vessel Integrity Simulation Analysis II) was used for deterministic calculation and simulation. Failure probability of the material is 1.E-5 for Octavia II distribution and 4E-6 for marshall distribution for each transient event postulated. The failure occurred at the 1.7th menit of the initial transient under 12.53 ksi of the pressure

  7. Investigation of a double oscillating-fan cooling device using electromagnetic force

    International Nuclear Information System (INIS)

    Su, Hsien-Chin; Xu, Han Yang

    2016-01-01

    Highlights: • The characteristics of a double oscillating-fan cooling device using electromagnetic force was investigated. • The driving current can be either DC PWM or AC within 3–12 V. • The comparison between a double blower pair, the model and a synjet were examined. • A 50 mm ∗ 50 mm ∗ 15 mm model can provide the flow rate of 154.89 l/min while consuming 0.65 W. • The flow rate, sound pressure, power consumption and two thermal tests have been done. - Abstract: This study proposes a double oscillating-fan cooling device using electromagnetic force. The device consists of two oscillating-fans. It requires only one electromagnet and two fan sheets with one magnet on each of them. The electromagnet and fan sheets are situated on a base and arranged accordingly. The electromagnetic force generated by the electromagnet can actuate the fan sheets. The main advantage of the device is its simple structure because there is no bearing and motor in the device. The driving current can be either DC PWM (Pulse width modulation) or AC (Alternating current) within 3–12 V so it is compatible with most electronic devices. The dimensions of the proposed model are 50 mm ∗ 50 mm ∗ 15 mm during operation. Concerning flow rate, sound pressure, power consumption and resonant frequency tests, a comparison between the proposed model and different type of cooling devices has been completed. The result shows that the model can provide cooling ability similar to a rotary fan while consuming 40% of the power of the rotary fan. It shows not only a good cooling ability but also a great potential for structural reliability and design flexibility.

  8. Nuclear reactor installation with outer shell enclosing a primary pressure vessel

    International Nuclear Information System (INIS)

    1975-01-01

    The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel and outer shell around the primary pressure vessel and acting as a protection for it against outside projectiles. A floor is provided internally dividing the outside shell into two upper and lower sections and an inside wall dividing the lower section into one part containing the primary pressure vessel and a second part, both made pressure tight with respect to each other and with the outside shell and forming with the latter a secondary means of containment [fr

  9. Experimental investigation of natural convection heat transfer in volumetrically heated spherical segments. Final report

    International Nuclear Information System (INIS)

    Asfia, F.; Dhir, V.

    1998-03-01

    One strategy for preventing the failure of lower head of a nuclear reactor vessel is to flood the concrete cavity with subcooled water in accidents in which relocation of core material into the vessel lower head occurs. After the core material relocates into the vessel, a crust of solid material forms on the inner wall of the vessel, however, most of the pool remains molten and natural convection exists in the pool. At present, uncertainty exists with respect to natural convection heat transfer coefficients between the pool of molten core material and the reactor vessel wall. In the present work, experiments were conducted to examine natural convection heat transfer in internally heated partially filled spherical pools with external cooling. In the experiments, Freon-113 contained in a Pyrex bell jar was used as a test liquid. The pool was bounded with a spherical segment at the bottom, and was heated with magnetrons taken from a conventional microwave oven. The vessel was cooled from the outside with natural convection of water or with nucleate boiling of liquid nitrogen

  10. Feasibility of a Miniature Esophageal Heat Exchange Device for Rapid Therapeutic Cooling in Newborns: Preliminary Investigations in a Piglet Model.

    Science.gov (United States)

    Dingley, John; Okano, Satomi; Planas, Silvia; Chakkarapani, Elavazhagan

    2018-03-01

    Therapeutic hypothermia (TH) after neonatal encephalopathy, commonly provided by 72 hours of whole-body cooling using a wrap, limits parents' physical contact with their infants affecting bonding and may not be suitable for encephalopathic preterm infants with fragile skin. Alternative cooling methods are unavailable for this population. We investigated in a neonatal pig model the feasibility of achieving a 3.5°C reduction in rectal temperature (T rectal ) similar to clinical TH protocols from 38.5°C (normothermia for pigs) to a target of 35°C ± 0.2°C, using a novel neonatal esophageal heat exchanger (NEHE), compared its efficacy to passive cooling, and investigated its ability to maintain target T rectal . Ventilated and anesthetized Landrace/Large white newborn pigs had the NEHE inserted. Water at adjustable temperatures and rates flowed down a central tube, returning up a surrounding distensible blind ending latex tube in a continuous loop. An initial experiment guided four subsequent cycles of passive cooling (30 minutes), rewarming to 38.5°C, active esophageal cooling to 35°C ± 0.2°C, active maintenance of target T rectal (30 minutes), and rewarming. We compared surface, rectal temperature, and hemodynamic changes among passive, active, and maintenance phases, and esophageal histopathology against control. Compared with passive cooling, esophageal cooling achieved target T rectal significantly earlier (71.3 minutes vs. 17.25 minutes, p = 0.003) with significantly greater rates of reduction in rectal (p = 0.0002) and surface (p = 0.005) temperatures and heart rate (p = 0.04). A water temperature of 39.1°C-40.2°C at a flow of 108-120 mL/min maintained T rectal around 35°C ± 0.2°C. The higher peak heart rate and blood pressure within 8 minutes of the maintenance phase (p = 0.04) subsequently stabilized. Histopathology showed congestion, edema, and neutrophil infiltration with increasing cycles. Esophageal cooling is

  11. FOREVER Experiments on Thermal and Mechanical Behavior of a Reactor Pressure Vessel during a Severe Accident

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A.; Green, J.A.; Bui, V.A.

    1999-01-01

    This paper describes the FOREVER (Failure Of Reactor Vessel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The facility employs 1/10.-scale carbon steel vessels of 0.4 m diameter, 15 mm thickness and 600 mm height. Up to 20 liters of binary-oxide melts with 100-300 K superheat are employed, as a simulant for the prototypic corium melt, and internal heating is provided by electrical heaters of up to 20 kW power in order to maintain the vessel wall temperatures at 1100-1200 K. Auxiliary systems are designed to provide an overpressure up to 4 MPa in the test vessel. Thus, severe accident scenarios with RCS depressurization are modeled. Creep behavior of the three-dimensional vessel, formation of the gap between the melt pool crust and the creeping vessel, and mechanisms of the gap cooling by water ingression will be the subjects of study and measurements in the FOREVER experimental program. Scaling rationale as well as pre-test analyses of the thermal and mechanical behavior of the FOREVER test vessels are presented. (authors)

  12. A scaling study of the natural circulation flow of the ex-vessel core catcher cooling system of a 1400MW PWR for designing a scale-down test facility

    International Nuclear Information System (INIS)

    Rhee, Bo. W.; Ha, K. S.; Park, R. J.; Song, J. H.

    2012-01-01

    A scaling study on the steady state natural circulation flow along the flow path of the ex-vessel core catcher cooling system of 1400MWe PWR is described. The scaling criteria for reproducing the same thermalhydraulic characteristics of the natural circulation flow as the prototype core catcher cooling system in the scale-down test facility is derived and the resulting natural circulation flow characteristics of the prototype and scale-down facility analyzed and compared. The purpose of this study is to apply the similarity law to the prototype EU-APR1400 core catcher cooling system and the model test facility of this prototype system and derive a relationship between the heating channel characteristics and the down-comer piping characteristics so as to determine the down-comer pipe size and the orifice size of the model test facility. As the geometry and the heating wall heat flux of the heating channel of the model test facility will be the same as those of the prototype core catcher cooling system except the width of the heating channel is reduced, the axial distribution of the coolant quality (or void fraction) is expected to resemble each other between the prototype and model facility. Thus using this fact, the down-comer piping design characteristics of the model facility can be determined from the relationship derived from the similarity law

  13. Numerical investigation of unsteady mixing mechanism in plate film cooling

    Directory of Open Access Journals (Sweden)

    Shuai Li

    2016-09-01

    Full Text Available A large-scale large eddy simulation in high performance personal computer clusters is carried out to present unsteady mixing mechanism of film cooling and the development of films. Simulation cases include a single-hole plate with the inclined angle of 30° and blowing ratio of 0.5, and a single-row plate with hole-spacing of 1.5D and 2D (diameters of the hole. According to the massive simulation results, some new unsteady phenomena of gas films are found. The vortex system is changed in different position with the development of film cooling with the time marching the process of a single-row plate film cooling. Due to the mutual interference effects including mutual exclusion, a certain periodic sloshing and mutual fusion, and the structures of a variety of vortices change between parallel gas films. Macroscopic flow structures and heat transfer behaviors are obtained based on 20 million grids and Reynolds number of 28600.

  14. Numerical investigations of cooling holes system role in the protection of the walls of a gas turbine combustion chamber

    Energy Technology Data Exchange (ETDEWEB)

    Ben Sik Ali, Ahlem; Kriaa, Wassim; Mhiri, Hatem [Ecole Nationale D' Ingenieurs de Monastir, Unite de Thermique et Thermodynamique des Procedes industriels, Monastir (Tunisia); Bournot, Philippe [IUSTI, UMR CNRS 6595, Marseille (France)

    2012-05-15

    Numerical simulations in a gas turbine Swirl stabilized combustor were conducted to investigate the effectiveness of a cooling system in the protection of combustor walls. The studied combustion chamber has a high degree of geometrical complexity related to the injection system as well as the cooling system based on a big distribution of small holes (about 3,390 holes) bored on the flame tube walls. Two cases were considered respectively the flame tube without and with its cooling system. The calculations were carried out using the industrial CFD code FLUENT 6.2. The various simulations made it possible to highlight the role of cooling holes in the protection of the flame tube walls against the high temperatures of the combustion products. In fact, the comparison between the results of the two studied cases demonstrated that the walls temperature can be reduced by about 800 C by the mean of cooling holes technique. (orig.)

  15. Conjugate heat transfer investigation on the cooling performance of air cooled turbine blade with thermal barrier coating

    Science.gov (United States)

    Ji, Yongbin; Ma, Chao; Ge, Bing; Zang, Shusheng

    2016-08-01

    A hot wind tunnel of annular cascade test rig is established for measuring temperature distribution on a real gas turbine blade surface with infrared camera. Besides, conjugate heat transfer numerical simulation is performed to obtain cooling efficiency distribution on both blade substrate surface and coating surface for comparison. The effect of thermal barrier coating on the overall cooling performance for blades is compared under varied mass flow rate of coolant, and spatial difference is also discussed. Results indicate that the cooling efficiency in the leading edge and trailing edge areas of the blade is the lowest. The cooling performance is not only influenced by the internal cooling structures layout inside the blade but also by the flow condition of the mainstream in the external cascade path. Thermal barrier effects of the coating vary at different regions of the blade surface, where higher internal cooling performance exists, more effective the thermal barrier will be, which means the thermal protection effect of coatings is remarkable in these regions. At the designed mass flow ratio condition, the cooling efficiency on the pressure side varies by 0.13 for the coating surface and substrate surface, while this value is 0.09 on the suction side.

  16. Experimental Investigation of Air Conditioner using the Desiccant Cooling System in Equatorial Climates

    Directory of Open Access Journals (Sweden)

    Abdullah Kamaruddin

    2018-01-01

    Full Text Available Indonesia lies in the tropical climate which requires air conditioning to increase working productivity of the people. Up to now people are still using the compressive cooling system which uses Freon as the refrigerant, which have been known to have a negative environmental impact. Therefore, new cooling system which is environmentally friendly is now needed. Desiccant cooling system manipulates the humidity condition of outside air in such a way so that the final temperature should become at 25 °C and RH of 65 %. Since it does not require refrigerant, a desiccant cooling has the potential to be developed in a tropical country like Indonesia. In this study an experimental desiccant cooling system has been designed and constructed and tested under laboratory condition. Experimental results have shown that the resulting air temperature was 26.1 °C with RH of 55.6 %, and average cooling capacity was 0.425 kW. The COP was found to be 0.44.

  17. Evaluation of HFIR vessel surveillance data and hydro-test conditions

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Nanstad, R.K.

    1994-01-01

    Surveillance specimens for the High Flux Isotope Reactor (HFIR) pressure vessel were removed and tested during 1993, after the vessel had accumulated 701,469 MWd of operation. The data agree well with HFIR surveillance data obtained in previous years. In conjunction with this effort, the vessel hydro-test conditions were reevaluated and found to be more than adequate. In view of this result, and because there are economic incentives for reducing the frequency of hydro testing, an analysis was performed to determine the minimum permissible frequency. The value obtained is substantially less than that presently specified. It was also determined that a somewhat lower cooling-tower-basin temperature is acceptable (improves operational flexibility). In 1986, after ∼20 years of reactor operation, it was discovered that the vessel embrittlement rate was substantially greater than expected. Possible reasons for the accelerated rate are reviewed in this report

  18. Development of Ultrasonic Visual Inspection Program for In-Vessel Structures of SFR

    International Nuclear Information System (INIS)

    Joo, Y. S.; Park, C. G.; Lee, J. H.

    2009-02-01

    As the liquid sodium of a sodium-cooled fast reactor (SFR) is opaque to light, a conventional visual inspection is unavailable for the evaluation of the in-vessel structures under a sodium level. ASME Section XI Division 3 provides rules and guidelines for an in-service inspection (ISI) and testing of the components of SFR. For the ISI of in-vessel structures, the ASME code specifies visual examinations. An ultrasonic wave should be applied for an under-sodium visual inspection of the in-vessel structures. The plate-type waveguide sensor has been developed and the feasibility of the waveguide sensor technique has been successfully demonstrated for an ultrasonic visual inspection of the in-vessel structures of SFR. In this study, the C-scan image mapping program (Under-Sodium MultiView) is developed to apply this waveguide sensor technology to an under-sodium visual inspection of in-vessel structures in SFR by using a LabVIEW graphical programming language. The Under-Sodium MultiVIEW program has the functions of a double rotating scanner motion control, a high power pulser receiver control, a image mapping and a signal processing. The performance of Under-Sodium MultiVIEW program was verified by a C-scanning test

  19. Learning from EDF investigations on SG divider plates and vessel head nozzles. Evidence of prior deformation effect on stress corrosion cracking

    International Nuclear Information System (INIS)

    Deforge, D.; Duisabeau, L.; Miloudi, S.; Thebault, Y.; Couvant, T.; Vaillant, F.; Lemaire, E.

    2011-01-01

    Nickel Based alloys Stress Corrosion Cracking (SCC) has been a major concern for all the Nuclear Power Plants (NPP) utilities since the beginning of the seventies. At EDF, the nineties were marked by the occurrence of cracks on vessel head nozzles. These cracks were responsible for a leak at Bugey 3 vessel head, which was the precursor leading to the replacement of all vessel heads. From 2002, new cases of Stress Corrosion Cracking were reported on Steam Generator (SG) Divider Plates (SGDP) welded junctions. These cracks are periodically inspected inservice and reparations could be performed in case of a significant evolution of the phenomenon even if the safety issue is less relevant than for the vessel head nozzles. Both issues have led to an important non-destructive testing (NDT) program and to destructive investigations campaigns. NDT were performed on an exhaustive basis for all vessel head nozzles and for all the divider plates of 900 MWe plants. Destructive investigations were performed on more than 30 vessel head nozzles and on 6 divider plates. The last investigations were performed on samples from two decommissioned Steam Generators of Chinon B1 which present SCC cracks. In this paper, the main conclusions driven from the analysis of both NDT and destructive investigation results are reported and a comparison of the behaviours of divider plates and vessel head nozzles is given. Results give evidence that prior plastic deformation of the components before operation is fundamental for the further environmental behaviour of the material. Analysis of field experience based on parameters characteristics of prior deformation and parameters characteristics of material microstructure can be used to account for the components which are the most sensitive to SCC cracking. Some perspectives on SCC predictive models are also presented. (authors)

  20. Preliminary Analysis on Decay Heat Removal Capability of Helium Cooled Solid Breeder Test Blanket Module

    International Nuclear Information System (INIS)

    Ahn, Mu Young; Cho, Seung Yon; Kim, Duck Hoi; Lee, Eun Seok; Kim, Hyung Seok; Suh, Jae Seung; Yun, Sung Hwan; Cho, Nam Zin

    2007-01-01

    One of the main ITER goals is to test and validate design concepts of tritium breeding blankets relevant to DEMO or fusion power plants. Korea Helium-Cooled Solid Breeder (HCSB) Test Blanket Module (TBM) has been developed with overall objectives of achieving this goal. The TBM employs high pressure helium to cool down the First Wall (FW), Side Wall (SW) and Breeding Zone (BZ). Therefore, safety consideration is a part of the design process. Each ITER Party performing the TBM program is requested to reach a similar level of confidence in the TBM safety analysis. To meet ITER's request, Failure Mode and Effects Analysis (FMEA) studies have been performed on the TBM to identify the Postulated Initial Event (PIE). Although FMEA on the KO TBM has not been completed, in-vessel, in-box and ex-vessel Loss Of Coolant Accident (LOCA) are considered as enveloping cases of PIE in general. In this paper, accidental analyses for the three selected LOCA were performed to investigate the decay heat removal capability of the TBM. To simulate transient thermo-hydraulic behavior of the TBM for the selected scenarios, RELAP5/MOD3.2 code was used

  1. Prediction of the Long Term Cooling Performance for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

    2005-12-15

    In the long term cooling phase that the emergency cooling water injection ends, the performance of the residual heat removal for the 3-pin fuel test loop has been predicted by a simplified heat transfer model. In the long term cooling phase the residual heat is 1323W for PWR fuel test mode and 1449W for CANDU fuel test mode. The each residual heat is assumed as 2% of the fission power of the test fuel used in the anticipated operational occurrence and design basis accident analyses. The each fission power used for the analyses is 105% of the rated fission power in the normal operation. In the long term cooling phase the residual heat is removed to the HANARO pool through the double pressure vessels of the in-pile test section. Saturate pooling boiling is assumed on the test fuel and condensation heat transfer is expected on the inner wall of the fuel carrier and the flow divider. Natural convection heat transfer on a heated vertical wall is also assumed on the outer wall of the outer pressure vessel. The conduction heat transfer is only considered in the gap between the double pressure vessels charged with neon gas and in the downcomer filled with coolant. The heat transfer rate between the coolant temperature of 152 .deg. C in the in-pile test section and the water temperature of 45 .deg. C in the HANARO pool is predicted as about 1666W. The 152 .deg. C is the saturate temperature of the coolant pressure predicted from the MARS code. The cooling capacity of 1666W is greater than the residual heats of 1323W and 1449W. Consequently the long term cooling performance of the 3-pin fuel test loop is sufficient for the anticipated operational occurrences and design basis accidents.

  2. A three-temperature model of selective photothermolysis for laser treatment of port wine stain containing large malformed blood vessels

    International Nuclear Information System (INIS)

    Li, D.; Wang, G.X.; He, Y.L.; Wu, W.J.; Chen, B.

    2014-01-01

    As congenital vascular malformations, port wine stain (PWS) is composed of ectatic venular capillary blood vessels buried within healthy dermis. In clinic, pulsed dye laser (PDL) in visible band (e.g. 585 nm) together with cryogen spray cooling (CSC) have become the golden standard for treatment of PWS. However, due to the limited energy deposition of the PDL in blood, large blood vessels are likely to survive from the laser irradiation. As a result, complete clearance of the lesions is rarely achieved. Assuming the local thermal non-equilibrium in skin tissue during the laser surgery, a three-temperature model is proposed to treat the PWS tissue as a porous media composed of a non-absorbing dermal matrix buried with the blood as well as the large malformed blood vessels. Three energy equations are constructed and solved coupling for the temperature of the blood in average-sized PWS vessels, non-absorbing dermal tissues and large malformed blood vessels, respectively. Subsequently, the thermal responses of human skin to visible (585 nm) and near-infrared (1064 nm) laser irradiations with various pulse durations in conjunction with cryogen spray cooling are investigated by the new model, and Arrhenius integral is used to analyze the thermal damage. The simulations show that the short pulse duration of 1.5 ms results in a higher selective heating of blood over epidermis, which will lead to a desired clinic outcome than the longer pulse duration. Due to a much deeper light penetration depth, laser irradiation with 1064 nm in wavelength is superior to that with 585 nm in treating patients with cutaneous hyper-vascular malformation. Complete coagulations are predicted in large-sized and deeply extending blood vessels by 1064 nm laser. - Highlights: •A three-temperature model is proposed for the laser treatment of port wine stain (PWS). •Average sized and large malformed blood vessels in porous medium (tissue) are considered. •Thermal responses of PWS to

  3. Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-03-01

    System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

  4. Melt cooling by bottom flooding. The COMET core-catcher concept

    International Nuclear Information System (INIS)

    Foit, Jerzy Jan; Alsmeyer, Hans; Tromm, Walter; Buerger, Manfred; Journeau, Christophe

    2009-01-01

    The COMET concept has been developed to cool an ex-vessel corium melt in case of a hypothetical severe accident leading to vessel melt-through. After erosion of a sacrificial concrete layer the melt is passively flooded by bottom injection of coolant water. The open porosities and large surface that are generated during melt solidification form a porous permeable structure that is permanently filled with the evaporating water and thus allows an efficient short-term as well as long-term removal of the decay heat. The advantages of this concept are the fast cool-down and complete solidification of the melt within less than one hour typically. This stops further release of fission products from the corium. A drawback may be the fast release of steam during the quenching process. Several experimental series have been performed by FZK (Germany) to test and optimise the functionality of the different variants of the COMET concept. Thermite generated melts of iron and aluminium oxide were used. The large scale COMET-H test series with sustained inductive heating includes nine experiments performed with an array of water injection channels embedded in a sacrificial concrete layer. Variation of the water inlet pressure and melt height showed that melts up to 50 cm height can be safely cooled with an overpressure of the coolant water of 0.2 bar. The CometPC concept is based on cooling by flooding the melt from the bottom through layers of porous, water filled concrete. The third variant of the COMET design, CometPCA, uses a layer of porous, water filled concrete CometPCA from which flow channels protrude into the layer of sacrificial concrete. This modified concept combines the advantages of the original COMET concept with flow channels and the high resistance of a water-filled porous concrete layer against downward melt attack. Four large scale CometPCA experiments (FZK, Germany) have demonstrated an efficient cooling of melts up to 50 cm height using the recommended water

  5. Study on the nuclear heat application system with a high temperature gas-cooled reactor and its safety evaluation (Thesis)

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo

    2008-03-01

    Aiming at the realization of the nuclear heat application system with a High Temperature Gas-cooled Reactor (HTGR), research and development on the whole evaluation of the system, the connection technology between the HTGR and a chemical plant such as the safety evaluation against the fire and explosion and the control technology, and the vessel cooling system of the HTGR were carried out. In the whole evaluation of the nuclear heat application system, an ammonia production system using nuclear heat was examined, and the technical subjects caused by the connection of the chemical plant to the HTGR were distilled. After distilling the subjects, the safety evaluation method against the fire and explosion to the reactor, the mitigation technology of thermal disturbance to the reactor, and the reactor core cooling by the vessel cooling system were discussed. These subjects are very important in terms of safety. About the fire and explosion, the safety evaluation method was established by developing the process and the numerical analysis code system. About the mitigation technology of the thermal disturbance, it was demonstrated that the steam generator, which was installed at the downstream of the chemical reactor in the chemical plant, could mitigate the thermal disturbance to the reactor. In order to enhance the safety of the reactor in accidents, the heat transfer characteristic of the passive indirect core cooling system was investigated, and the heat transfer equation considering both thermal radiation and natural convection was developed for the system design. As a result, some technical subjects related to safety in the nuclear heat application system were solved. (author)

  6. Review of in-service thermal annealing of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Server, W.L.

    1984-01-01

    Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper-shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. A test reactor pressure vessel has been wet annealed at less than 343 0 C (650 0 F), and annealing of the Belgian BR-3 reactor vessel has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place is feasible, but solvable engineering problems do exist. The materials with highest radiation sensitivity in the older reactor vessels are submerged-arc weld metals with high copper and nickel concentrations. The limited Charpy V-notch and fracture toughness data available for five such welds were reviewed. The review suggested that significant recovery results from annealing at 454 0 C (850 0 F) for one week. Two of the main concerns with a localized heat treatment at 454 0 C (850 0 F) are the degree of distortion that may occur after the annealing cycle and the extent of residual stresses. A thermal and structural analysis of a reactor vessel for distortions and residual stresses found no problems with the reactor vessel itself but did indicate a rotation at the nozzle region of the vessel that would plastically deform the attached primary piping. Further analytical studies are needed. An American Society for Testing and Materials (ASTM) task group is upgrading and revising the ASTM Recommended Guide for In-Service Annealing of WaterCooled Nuclear Reactor Vessels (E 509-74) with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (for example, the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

  7. Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.1. Design of the multi-cavity prestressed concrete reactor vessel with warm liner

    International Nuclear Information System (INIS)

    Lafitte, R.; Marchand, J.D.

    1979-01-01

    The design studies and tests described in this paper were undertaken as part of ''PROJECT HHT'', a German-Swiss joint effort for the development of high-temperature helium cooled reactors with direct-cycle turbine. The prestressed concrete reactor pressure vessel encloses the core of the reactor itself, the heat exchangers (coolers and recuperators), the helium turbine, the main helium circuit, all nuclear and thermal equipment, and auxiliary reactor cooling equipment. In order to make the liner accessible for inspection, no thermal insulation is provided between the coolant and the liner. The temperature of the helium in contact with the liner is limited to 200 0 C, under all normal operation conditions of the reactor. In the HHT reactor pressure vessel, the resisting structure is protected thermally by a layer of warm concrete between the liner and the structural prestressed concrete. The main features of this pressure vessel are the marked pressure differences in the cavities during normal operation, and the use of warm liner. The objectives of the reference design were chiefly related to the sizing up of the main structure, taking into account the modifications to be expected in the material characteristics as a result of the high temperatures developed

  8. Distribution of normal superficial ocular vessels in digital images.

    Science.gov (United States)

    Banaee, Touka; Ehsaei, Asieh; Pourreza, Hamidreza; Khajedaluee, Mohammad; Abrishami, Mojtaba; Basiri, Mohsen; Daneshvar Kakhki, Ramin; Pourreza, Reza

    2014-02-01

    To investigate the distribution of different-sized vessels in the digital images of the ocular surface, an endeavor which may provide useful information for future studies. This study included 295 healthy individuals. From each participant, four digital photographs of the superior and inferior conjunctivae of both eyes, with a fixed succession of photography (right upper, right lower, left upper, left lower), were taken with a slit lamp mounted camera. Photographs were then analyzed by a previously described algorithm for vessel detection in the digital images. The area (of the image) occupied by vessels (AOV) of different sizes was measured. Height, weight, fasting blood sugar (FBS) and hemoglobin levels were also measured and the relationship between these parameters and the AOV was investigated. These findings indicated a statistically significant difference in the distribution of the AOV among the four conjunctival areas. No significant correlations were noted between the AOV of each conjunctival area and the different demographic and biometric factors. Medium-sized vessels were the most abundant vessels in the photographs of the four investigated conjunctival areas. The AOV of the different sizes of vessels follows a normal distribution curve in the four areas of the conjunctiva. The distribution of the vessels in successive photographs changes in a specific manner, with the mean AOV becoming larger as the photos were taken from the right upper to the left lower area. The AOV of vessel sizes has a normal distribution curve and medium-sized vessels occupy the largest area of the photograph. Copyright © 2013 British Contact Lens Association. Published by Elsevier Ltd. All rights reserved.

  9. In-Vessel Coolability. Workshop Proceedings, in collaboration with EC-SARNET

    International Nuclear Information System (INIS)

    2011-01-01

    Severe Accident Management Guidelines increase focus on containment integrity after some progression in the course of a severe accident. This change in priorities is made according to criteria that vary depending on reactor type and specific procedures. Once a water source has been recovered, different accident management strategies can be used: send water into the core and/or cool the reactor pressure vessel (RPV) externally. It should be noticed that, depending on the amount of water available, these strategies might conflict with other uses of water such as for instance activating spray systems in the containment or may have deleterious effects as for instance an increase in the production of hydrogen. Generally, for in-vessel reflooding, the models used for evaluation of accident management measures suffer from a lack of validation. Given this background, the objectives of the workshop were: -) to exchange information on different Severe Accident Management strategies used or contemplated for the in-vessel coolability issue; -) to review recent, ongoing and planned experimental programmes on reflooding; -) to review models used for reflooding in severe accident calculation tools, either simplified or sophisticated; -) to exchange information on the treatment of reflooding in different safety studies such as Probabilistic Safety Assessment; and -) to provide recommendations for future work, as necessary

  10. Basic Boiling Experiments with An Inclined Narrow Gap Associated With In-Vessel Retention

    International Nuclear Information System (INIS)

    Terazu, Kuninobu; Watanabe, Fukashi; Iwaki, Chikako; Yokobori, Seiichi; Akinaga, Makoto; Hamazaki, Ryoichi; SATO, Ken-ichi

    2002-01-01

    In the case of a severe accident with relocation of the molten corium into the lower plenum of reactor pressure vessel (RPV), the successful in-vessel corium retention (IVR) can prevent the progress to ex-vessel events with uncertainties and avoid the containment failure. One of the key phenomena governing the possibility of IVR would be the gap formation and cooling between a corium crust and the RPV wall, and for the achievement of IVR, it would be necessary to supply cooling water to RPV as early as possible. The BWR features relative to IVR behavior are a deep and massive water pool in the lower plenum, and many of control rod drive guide tubes (CRDGT) installed in the lower head of RPV, in which water is injected continuously except in the case of station blackout scenario. The present paper describes the basic boiling experiment conducted in order to investigate the boiling characteristics in an inclined narrow gap simulating a part of the lower head curvature. The boiling experiments were composed of visualization tests and heat transfer tests. In the visualization tests, two types of inclined gap were constructed using the parallel plate and the V-shaped parallel plate with heating from the top plate, and the boiling flow pattern was observed with various gap width and heat flux. These observation results showed that water was easily supplied from the gap bottom of parallel plate even in a very narrow gap with smaller width than 1 mm, and water could flow continuously in the narrow gap by the geometric and thermal imbalance from the experiment results using the V-shaped parallel plate. In the heat transfer tests, the critical heat flux (CHF) data in an inclined narrow channel formed by the parallel plates were measured in terms of the parameters of gap width, heated length and inclined angle of a channel, and the effect of inclination was incorporated into the existing CHF correlation for a narrow gap. The CHF correlation modified for an inclined narrow gap

  11. Design of Hemispherical Downward-Facing Vessel for Critical Heat Flux Experiment

    International Nuclear Information System (INIS)

    Hwang, J. S.; Suh, K. Y.

    2009-01-01

    The in-vessel retention (IVR) is one of major severe accident management strategies adopted by some operating nuclear power plants during a severe accident. The recent Shin-Gori Units 3 and 4 of the Advanced Power Reactor 1400 MWe (APR1400) have adopted the external reactor vessel cooling (ERVC) by reactor cavity flooding as major severe accident management strategy. The ERVC in the APR1400 design resorts to active flooding system using thermal insulator. The Corium Attack Stopper Apparatus Spherical Channel (CASA SC) tests are conducted to measure the critical power and critical heat flux (CHF) on a downward hemispherical vessel scaled down from the APR1400 lower head by 1/10 on a linear scale. CASA is designed through scaling and thermal analysis to simulate the APR1400 vessel and thermal insulator. The heated vessel of CASA SC represents the external surface of a hemisphere submerged vessel in water. The heated vessel plays an important role in the ERVC experiment depending on the configuration of oxide pool and metallic layer. Hand calculation and computational analysis are performed to produce high heat flux from the downward facing hemisphere in excess of 1 MW/m 2

  12. Legionnaires' disease from a cooling tower in a community outbreak in Lidköping, Sweden- epidemiological, environmental and microbiological investigation supported by meteorological modelling.

    Science.gov (United States)

    Ulleryd, Peter; Hugosson, Anna; Allestam, Görel; Bernander, Sverker; Claesson, Berndt E B; Eilertz, Ingrid; Hagaeus, Anne-Christine; Hjorth, Martin; Johansson, Agneta; de Jong, Birgitta; Lindqvist, Anna; Nolskog, Peter; Svensson, Nils

    2012-11-21

    An outbreak of Legionnaires' Disease took place in the Swedish town Lidköping on Lake Vänern in August 2004 and the number of pneumonia cases at the local hospital increased markedly. As soon as the first patients were diagnosed, health care providers were informed and an outbreak investigation was launched. Classical epidemiological investigation, diagnostic tests, environmental analyses, epidemiological typing and meteorological methods. Thirty-two cases were found. The median age was 62 years (range 36 - 88) and 22 (69%) were males. No common indoor exposure was found. Legionella pneumophila serogroup 1 was found at two industries, each with two cooling towers. In one cooling tower exceptionally high concentrations, 1.2 × 109 cfu/L, were found. Smaller amounts were also found in the other tower of the first industry and in one tower of the second plant. Sero- and genotyping of isolated L. pneumophila serogroup 1 from three patients and epidemiologically suspected environmental strains supported the cooling tower with the high concentration as the source. In all, two L. pneumophila strains were isolated from three culture confirmed cases and both these strains were detected in the cooling tower, but one strain in another cooling tower as well. Meteorological modelling demonstrated probable spread from the most suspected cooling tower towards the town centre and the precise location of four cases that were stray visitors to Lidköping. Classical epidemiological, environmental and microbiological investigation of an LD outbreak can be supported by meteorological modelling methods.The broad competence and cooperation capabilities in the investigation team from different authorities were of paramount importance in stopping this outbreak.

  13. Legionnaires’ disease from a cooling tower in a community outbreak in Lidköping, Sweden- epidemiological, environmental and microbiological investigation supported by meteorological modelling

    Science.gov (United States)

    2012-01-01

    Background An outbreak of Legionnaires’ Disease took place in the Swedish town Lidköping on Lake Vänern in August 2004 and the number of pneumonia cases at the local hospital increased markedly. As soon as the first patients were diagnosed, health care providers were informed and an outbreak investigation was launched. Methods Classical epidemiological investigation, diagnostic tests, environmental analyses, epidemiological typing and meteorological methods. Results Thirty-two cases were found. The median age was 62 years (range 36 – 88) and 22 (69%) were males. No common indoor exposure was found. Legionella pneumophila serogroup 1 was found at two industries, each with two cooling towers. In one cooling tower exceptionally high concentrations, 1.2 × 109 cfu/L, were found. Smaller amounts were also found in the other tower of the first industry and in one tower of the second plant. Sero- and genotyping of isolated L. pneumophila serogroup 1 from three patients and epidemiologically suspected environmental strains supported the cooling tower with the high concentration as the source. In all, two L. pneumophila strains were isolated from three culture confirmed cases and both these strains were detected in the cooling tower, but one strain in another cooling tower as well. Meteorological modelling demonstrated probable spread from the most suspected cooling tower towards the town centre and the precise location of four cases that were stray visitors to Lidköping. Conclusions Classical epidemiological, environmental and microbiological investigation of an LD outbreak can be supported by meteorological modelling methods. The broad competence and cooperation capabilities in the investigation team from different authorities were of paramount importance in stopping this outbreak. PMID:23171054

  14. Legionnaires’ disease from a cooling tower in a community outbreak in Lidköping, Sweden- epidemiological, environmental and microbiological investigation supported by meteorological modelling

    Directory of Open Access Journals (Sweden)

    Ulleryd Peter

    2012-11-01

    Full Text Available Abstract Background An outbreak of Legionnaires’ Disease took place in the Swedish town Lidköping on Lake Vänern in August 2004 and the number of pneumonia cases at the local hospital increased markedly. As soon as the first patients were diagnosed, health care providers were informed and an outbreak investigation was launched. Methods Classical epidemiological investigation, diagnostic tests, environmental analyses, epidemiological typing and meteorological methods. Results Thirty-two cases were found. The median age was 62 years (range 36 – 88 and 22 (69% were males. No common indoor exposure was found. Legionella pneumophila serogroup 1 was found at two industries, each with two cooling towers. In one cooling tower exceptionally high concentrations, 1.2 × 109 cfu/L, were found. Smaller amounts were also found in the other tower of the first industry and in one tower of the second plant. Sero- and genotyping of isolated L. pneumophila serogroup 1 from three patients and epidemiologically suspected environmental strains supported the cooling tower with the high concentration as the source. In all, two L. pneumophila strains were isolated from three culture confirmed cases and both these strains were detected in the cooling tower, but one strain in another cooling tower as well. Meteorological modelling demonstrated probable spread from the most suspected cooling tower towards the town centre and the precise location of four cases that were stray visitors to Lidköping. Conclusions Classical epidemiological, environmental and microbiological investigation of an LD outbreak can be supported by meteorological modelling methods. The broad competence and cooperation capabilities in the investigation team from different authorities were of paramount importance in stopping this outbreak.

  15. An improved water cooled nuclear reactor and pressuriser assembly

    International Nuclear Information System (INIS)

    Gardner, F.J.; Strong, R.

    1991-01-01

    A water cooled nuclear reactor is described which comprises a reactor core, a primary water coolant circuit and a pressuriser arranged as an integral unit in a pressure vessel. The pressure vessel is divided into an upper and a lower chamber by a casing. The reactor core and primary water coolant circuit are arranged in the lower chamber and the pressuriser is arranged in the upper chamber. A plurality of spray pipes interconnect a steam space of the pressuriser with the downcomer of the primary water coolant circuit below a heat exchanger. A plurality of surge ports interconnect a water space of the pressuriser with the primary water coolant circuit. The surge ports have hydraulic diodes so that there is a low flow resistance for water from the water space of the pressuriser to the primary water coolant circuit and high flow resistance in the opposite direction. The spray pipes provide a desuperheating spray of cooled water into the pressuriser during positive volume surges of the primary water coolant. The pressuriser arrangement may also be applied to integral water cooled reactors with separate pressurisers and to dispersed pressurised water reactors. The surge ports also allow water to flow by gravity to the core in an emergency. (author)

  16. Prestressed concrete vessels suitable for helium high temperature reactors

    International Nuclear Information System (INIS)

    Lockett, G.E.; Kinkead, A.N.

    1967-02-01

    In considering prestressed concrete vessels for use with helium cooled high temperature reactors, a number of new problems arise and projected designs involve new approaches and new solutions. These reactors, having high coolant outlet temperature from the core and relatively high power densities, can be built into compact designs which permit usefully high working pressures. Consequently, steam generators and circulating units tend to be small. Although circuit activity can be kept quite low with coated particle fuels, designs which involve entry for subsequent repair are not favoured, and coupled with the preferred aim of using fully shop fabricated units within the designs with removable steam generators which involve no tube welding inside the vessel. A particular solution uses a number of slim cylindrical assemblies housed in the wall of the pressure vessel and this vessel design concept is presented. The use of helium requires very high sealing standards and one of the important requirements is a vessel design which permits leak testing during construction, so that a repair seal can be made to any faulty part in a liner seam. Very good demountable joint seals can be made without particular difficulty and Dragon experience is used to provide solutions which are suitable for prestressed concrete vessel penetrations. The concept layout is given of a vessel meeting these requirements; the basis of design is outlined and special features of importance discussed. (author)

  17. Finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Smith, P.D.; Cook, W.A.; Anderson, C.A.

    1977-01-01

    Several present and proposed gas-cooled reactors use concrete pressure vessels. In addition, concrete is almost universally used for the secondary containment structures of water-cooled reactors. Regulatory agencies must have means of assuring that these concrete structures perform their containment functions during normal operation and after extreme conditions of transient overpressure and high temperature. The NONSAP nonlinear structural analysis program has been extensively modified to provide one analytical means of assessing the safety of reinforced concrete pressure vessels and containments. Several structural analysis codes were studied to evaluate their ability to model the nonlinear static and dynamic behavior of three-dimensional structures. The NONSAP code was selected because of its availability and because of the ease with which it can be modified. In particular, the modular structure of this code allows ready addition of specialized material models. Major modifications have been the development of pre- and post-processors for mesh generation and graphics, the addition of an out-of-core solver, and the addition of constitutive models for reinforced concrete subject to either long-term or short-term loads. Emphasis was placed on development of a three-dimensional analysis capability

  18. Investigation of lactose crystallization process during condensed milk cooling using native vacuum-crystallizer

    Directory of Open Access Journals (Sweden)

    E. I. Dobriyan

    2016-01-01

    Full Text Available One of the most general defects of condensed milk with sugar is its consistency heterogeneity – “candying”. The mentioned defect is conditioned by the presence of lactose big crystals in the product. Lactose crystals size up to 10 µm is not organoleptically felt. The bigger crystals impart heterogeneity to the consistency which can be evaluated as “floury”, “sandy”, “crunch on tooth”. Big crystals form crystalline deposit on the can or industrial package bottom in the form of thick layer. Industrial processing of the product with the defective process of crystallization results in the expensive equipment damage of the equipment at the confectionary plant accompanied with heavy losses. One of the factors influencing significantly lactose crystallization is the product cooling rate. Vacuum cooling is the necessary condition for provision of the product consistency homogeneity. For this purpose the vacuum crystallizers of “Vigand” company, Germany, are used. But their production in the last years has been stopped. All-Russian dairy research institute has developed “The references for development of the native vacuum crystallizer” according to which the industrial model has been manufactured. The produced vacuum – crystallizer test on the line for condensed milk with sugar production showed that the product cooling on the native vacuum-crystallizer guarantees production of the finished product with microstructure meeting the requirements of State standard 53436–2009 “Canned Milk. Milk and condensed cream with sugar”. The carried out investigations evidences that the average lactose crystals size in the condensed milk with sugar cooled at the native crystallizer makes up 6,78 µm. The granulometric composition of the product crystalline phase cooled at the newly developed vacuum-crystallizer is completely identical to granulometric composition of the product cooled at “Vigand” vacuum-crystallizer.

  19. Investigation of Heat Sink Efficiency for Electronic Component Cooling Applications

    DEFF Research Database (Denmark)

    Staliulionis, Ž.; Zhang, Zhe; Pittini, Riccardo

    2014-01-01

    Research and optimisation of cooling of electronic components using heat sinks becomes increasingly important in modern industry. Numerical methods with experimental real-world verification are the main tools to evaluate efficiency of heat sinks or heat sink systems. Here the investigation...... of relatively simple heat sink application is performed using modeling based on finite element method, and also the potential of such analysis was demonstrated by real-world measurements and comparing obtained results. Thermal modeling was accomplished using finite element analysis software COMSOL and thermo...

  20. Stress and Thermal Analysis of the In-Vessel RMP Coils in HL-2M

    International Nuclear Information System (INIS)

    Cen Yishun; Li Qiang; Cai Lijun; Jiang Jiaming; Li Guangsheng; Liu Yi; Ding Yonghua

    2013-01-01

    A set of in-vessel resonant magnetic perturbation (RMP) coils for MHD instability suppression is proposed for the design of a HL-2M tokamak. Each coil is to be fed with a current of up to 5 kA, operated in a frequency range from DC to about 1 kHz. Stainless steel (SS) jacketed mineral insulated cables are proposed for the conductor of the coils. In-vessel coils must withstand large electromagnetic (EM) and thermal loads. The support, insulation and vacuum sealing in a very limited space are crucial issues for engineering design. Hence finite element calculations are performed to verify the design, optimize the support by minimizing stress caused by EM forces on the coil conductors and work out the temperature rise occurring on the coil in different working conditions, the corresponding thermal stress caused by the thermal expansion of materials is evaluated to be allowable. The techniques to develop the in-vessel RMP coils, such as support, insulation and cooling, are discussed

  1. Conceptual design tool development for a Pb-Bi cooled reactor

    International Nuclear Information System (INIS)

    Lee, K. G.; Chang, S. H.; No, H. C.; Chunm, M. H.

    2000-01-01

    Conceptual design is generally ill-structured and mysterious problem solving. This leads the experienced experts to be still responsible for the most of synthesis and analysis task, which are not amenable to logical formulations in design problems. Especially because a novel reactor such as a Pb-Bi cooled reactor is going on a conceptual design stage, it will be very meaningful to develop the conceptual design tool. This tool consists of system design module with artificial intelligence, scaling module, and validation module. System design decides the optimal structure and the layout of a Pb-Bi cooled reactor, using design synthesis part and design analysis part. The designed system is scaled to be optimal with desired power level, and then the design basis accidents (Dbase) are analyzed in validation module. Design synthesis part contains the specific data for reactor components and the general data for a Pb-Bi cooled reactor. Design analysis part contains several design constraints for formulation and solution of a design problem. In addition, designer's intention may be externalized through emphasis on design requirements. For the purpose of demonstration, the conceptual design tool is applied to a Pb-Bi cooled reactor with 125 M Wth of power level. The Pb-Bi cooled reactor is a novel reactor concept in which the fission-generated heat is transferred from the primary coolant to the secondary coolant through a reactor vessel wall of a novel design. The Pb-Bi cooled reactor is to deliver 125 M Wth per module for 15 effective full power years without any on-site fuel handling. The conceptual design tool investigated the feasibility of a Pb-Bi cooled reactor. Application of the conceptual design tool will be, in detail, presented in the full paper. (author)

  2. Stress-relieving annealing of Cr-Mo steel for high temperature pressure vessels and the quality change in use

    International Nuclear Information System (INIS)

    Makioka, Minoru; Hirano, Hiromichi

    1976-01-01

    The securing of good mechanical properties is difficult in thick plates for large pressure vessels because cooling rate is insufficient and time is prolonged in heat treatment. Cr-Mo steel plates are usually used in the state of improved notch toughness though somewhat reduced strength by normalizing or accelerated cooling and tempering. If the time for heat treatment is prolonged, the embrittlement occurs. The effects of temperature, holding time, and cooling rate in stress-relieving treatment on the mechanical properties of 1-1/4Cr - 1/2Mo, 2-1/4Cr - 1Mo, 3Cr - 1Mo, and 5Cr - 1/2Mo steels were investigated. The tensile strength lowered almost linearly as the hollomon-Jaffe parameter of heat treatment condition increased in all the steels. The transition temperature shifted continuously to high temperature side in 1-1/4Cr - 1/2Mo steel, but the notch toughness was improved up to certain values and then the tendency turning to brittleness was shown in the other steels, as the H-J parameter increased. When the holding time became longer, the transition temperature shifted to higher temperature side, but the cooling rate showed no effect. The condition for stress relieving treatment must be selected so that the ferrite bands observed in welded metal do not arise. The embrittlement at the operation temperature of 400 - 450 0 C for a long time is evaluated by the comparison with that by stepped cooling method. (Kako, I.)

  3. Corrosion of vessel steel during its interaction with molten corium

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Lopukh, D.B.; Gusarov, V.V.; Martinov, A.P.; Martinov, V.V.; Fieg, G.; Tromm, W.; Bottomley, D.; Tuomisto, H.

    2006-01-01

    An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments

  4. Corrosion of vessel steel during its interaction with molten corium

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation)]. E-mail: bechta@sbor.spb.su; Khabensky, V.B. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Vitol, S.A. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Krushinov, E.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Granovsky, V.S. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Lopukh, D.B. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Gusarov, V.V. [Institute of Silicate Chemistry of Russian Academy of Science (ISC of RAS), Odoevsky str., b. 24/2, 199155 St. Petersburg (Russian Federation); Martinov, A.P. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Martinov, V.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Fieg, G. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Tromm, W. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Bottomley, D. [Europaeische Kommission, General Direktion GFS, Institut fuer Transurane (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Tuomisto, H. [Fortum Engineering Ltd. 00048 FORTUM, Rajatorpantie 8, Vantaa (Finland)

    2006-07-15

    An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments.

  5. Experimental investigation of electron cooling and stacking of lead ions in a low energy accumulation ring

    CERN Document Server

    Bosser, Jacques; Chanel, M; Hill, C; Lombardi, A M; MacCaferri, R; Maury, S; Möhl, D; Molinari, G; Rossi, S; Tanke, E; Tranquille, G; Vretenar, Maurizio

    1999-01-01

    This report gives the results of a programme of experimental investigations, which were carried out to test stacking of lead ions in a storage ring (the former Low Energy Antiproton Ring, LEAR) at 4.2 MeV per nucleon. The motivation was to demonstrate the feasibility of gaining the large factor in the phase-space density required for injection into the LHC. In the first part of the report, the layout of the experiments is described, the choice of the parameters of the electron cooling system used for stacking is reported and the multi-turn injection using horizontal- and longitudinal- (and in the final project also vertical-) phase space is discussed. In the second part the experimental results are presented. Factors of vital importance are the stacking efficiency, the beam life-time and the cooling time of the ions. The beam decay owing to charge exchange with the residual gas and to recombination by the capture of cooling electrons was intensively studied. Beam instabilities and space-charge effects in the ...

  6. An overview of experimental results obtained under the prestressed concrete nuclear pressure vessel development program at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Naus, D.J.

    1979-01-01

    Under the Prestressed Concrete Nuclear Pressure Development Program at the Oak Ridge National Laboratory, various aspects of Prestressed Concrete Pressure Vessels (PCPVs) are investigated with respect to reliability, structural performance, constructability, and economy. These investigations are conducted under the High-Temperature Gas-Cooled Reactor (HTGR) Program and the Gas-Cooled Fast Reactor (GCFR) Program. The objectives are to: (1) provide technical support to ongoing PCPV design activities, (2) contribute to the overall technological data base, and (3) provide independent review and evaluations. Specific areas of interest at present include finite-element analysis development, materials and structural behaviour tests, instrumentation evaluations and development, and structural model tests. The following provides an overview of both the HTGR and GCFR PCPV activities and a summary of recent experimental results

  7. Vessel size measurements in angiograms: Manual measurements

    International Nuclear Information System (INIS)

    Hoffmann, Kenneth R.; Dmochowski, Jacek; Nazareth, Daryl P.; Miskolczi, Laszlo; Nemes, Balazs; Gopal, Anant; Wang Zhou; Rudin, Stephen; Bednarek, Daniel R.

    2003-01-01

    Vessel size measurement is perhaps the most often performed quantitative analysis in diagnostic and interventional angiography. Although automated vessel sizing techniques are generally considered to have good accuracy and precision, we have observed that clinicians rarely use these techniques in standard clinical practice, choosing to indicate the edges of vessels and catheters to determine sizes and calibrate magnifications, i.e., manual measurements. Thus, we undertook an investigation of the accuracy and precision of vessel sizes calculated from manually indicated edges of vessels. Manual measurements were performed by three neuroradiologists and three physicists. Vessel sizes ranged from 0.1-3.0 mm in simulation studies and 0.3-6.4 mm in phantom studies. Simulation resolution functions had full-widths-at-half-maximum (FWHM) ranging from 0.0 to 0.5 mm. Phantom studies were performed with 4.5 in., 6 in., 9 in., and 12 in. image intensifier modes, magnification factor = 1, with and without zooming. The accuracy and reproducibility of the measurements ranged from 0.1 to 0.2 mm, depending on vessel size, resolution, and pixel size, and zoom. These results indicate that manual measurements may have accuracies comparable to automated techniques for vessels with sizes greater than 1 mm, but that automated techniques which take into account the resolution function should be used for vessels with sizes smaller than 1 mm

  8. Shear strength of end slabs of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Cheung, K.C.; Gotschall, H.L.; Liu, T.C.

    1975-01-01

    Prestressed concrete reactor vessels (PCRV's) have been adopted for primary containments in most large high-temperature gas-cooled reactor installations. The most common configuration for PCRVs is a right-vertical cylinder with thick end slabs. In order to assess the integrity of a PCRV it is necessary to predict the ultimate strength of the end slabs. The complexity of the basic mechanism of shear failure in the PCRV end slabs has thus far prohibited the development of a completely analytical solution. However, many experimental investigations of PCRV end slabs have been conducted over the past decade. This information makes it possible to establish empirical formulae for the ultimate strength of PCRV end slabs. The basis and development of an empirical shear-flexure interaction expression is presented. (Auth.)

  9. Cooling unit for a superconducting power cable. Two years successful operation

    Energy Technology Data Exchange (ETDEWEB)

    Herzog, Friedhelm [Messer Group GmbH, Krefeld (Germany); Kutz, Thomas [Messer Industriegase GmbH, Bad Soden (Germany); Stemmle, Mark [Nexans Deutschland GmbH, Hannover (Germany); Kugel, Torsten [Westnetz GmbH, Essen (Germany)

    2016-07-01

    High temperature super conductors (HTS) can efficiently be cooled with liquid nitrogen down to a temperature of 64 K (-209 C). Lower temperatures are not practical, because nitrogen becomes solid at 63 K (-210 C). To achieve this temperature level the coolant has to be vaporized below atmospheric pressure. Messer has developed a cooling unit with an adequate vacuum subcooler, a liquid nitrogen circulation system, and a storage vessel for cooling an HTS power cable. The cooling unit was delivered in 2013 for the German AmpaCity project of RWE Deutschland AG, Nexans and Karlsruhe Institute of Technology. Within this project RWE and Nexans installed the worldwide longest superconducting power cable in the city of Essen, Germany. The cable is in operation since March 10th, 2014.

  10. An experimental investigation of natural circulated air flow in the passive containment cooling system

    International Nuclear Information System (INIS)

    Ryu, S.H.; Oh, S.M.; Park, G.C.

    2004-01-01

    The objective of this study is to investigate the effects of air inlet position and external conditions on the natural circulated air flow rate in a passive containment cooling system of the advanced passive reactor. Experiments have been performed with 1/36 scaled segment type passive containment test facility. The air velocities and temperatures are measured through the air flow path. Also, the experimental results are compared with numerical calculations and show good agreement. (author)

  11. Experimental investigations on vessel-hole ablation during severe accidents

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Dinh, T.N.; Green, J.A.; Paladino, D.

    1997-12-01

    This report presents experimental results, and subsequent analyses, of scaled reactor pressure vessel (RPV) failure site ablation tests conducted at the Royal Institute of Technology, Division of Nuclear Power Safety (RIT/NPS). The goal of the test program is to reduce the uncertainty level associated with the phase-change-ablation process, and, thus, improve the characterization of the melt discharge loading on the containment. In a series of moderate temperature experiments, the corium melt is simulated by the binary oxide CaO-B 2 O 3 or the binary eutectic and non-eutectic salts NaNO 3 -KNO 3 , while the RPV head steel is represented by a Pb, Sn or metal alloys plate. A complementary set of experiments was conducted at lower temperatures, using water as melt and salted ice as plate material. These experiments scale well to the postulated prototypical conditions. The multidimensional code HAMISA, developed at RIT/NPS, is employed to analyze the experiments with good pre- and post-test predictions. The effects of melt viscosity and crust surface roughness, along with failure site entrance and exit frictional losses on the ablation characteristics are investigated. Theoretical concept was proposed to describe physical mechanisms which govern the vessel-hole ablation process during core melt discharge from RPV. Experimental data obtained from hole ablation tests and separate-effect tests performed at RIT/NPS were used to validate component physical models of the HAMISA code. It is believed that the hole ablation phenomenology is quite well understood. Detailed description of experiments and experimental data, as well as results of analyses are provided in the appendixes

  12. The influence of residual stresses on small through-clad cracks in pressure vessels

    International Nuclear Information System (INIS)

    deLorenzi, H.G.; Schumacher, B.I.

    1984-01-01

    The influence of cladding residual stresses on the crack driving force for shallow cracks in the wall of a nuclear pressure vessel is investigated. Thermo-elastic-plastic analyses were carried out on long axial through-clad and sub-clad flaws on the inside of the vessel. The depth of the flaws were one and three times the cladding thickness, respectively. An analysis of a semielliptical axial through-clad flaw was also performed. It was assumed that the residual stresses arise due to the difference in the thermal expansion between the cladding and the base material during the cool down from stress relieving temperature to room temperature and due to the subsequent proof test before the vessel is put into service. The variation of the crack tip opening displacement during these loadings and during a subsequent thermal shock on the inside wall is described. The analyses for the long axial flaws suggest that the crack driving force is smaller for this type of flaw if the residual stresses in the cladding are taken into account than if one assumes that the cladding has no residual stresses. However, the analysis of the semielliptical flaw shows significantly different results. Here the crack driving force is higher than when the residual stresses are not taken into account and is maximum in the cladding at or near the clad/base material interface. This suggests that the crack would propagate along the clad/base material interface before it would penetrate deeper into the wall. The elastic-plastic behavior found in the analyses show that the cladding and the residual stresses in the cladding should be taken into acocunt when evaluating the severity of shallow surface cracks on the inside of a nuclear pressure vessel

  13. MARS vessel safety analysis. LATA report No. 115

    International Nuclear Information System (INIS)

    Rigdon, L.D.; Donham, B.J.; Hughes, P.S.

    1979-08-01

    A previous study was performed to assess the hazards associated with an accidental leakage of cooling water into the crucible of molten 238 U for the MARS laser isotope separation experiment. Since that study found that the probability of such an explosion is extremely low during an accidental cooling system failure, a study was conducted to define a more realistic design basis accident (DBA) for the final MARS configuration. If the vapor-phase explosion is considered to be a significant threat, the design criteria for the vacuum vessel should be a working pressure of 67 psig or 101 psig momentary single pulse equivalent static pressure

  14. Two-Phase Flow Effect on the Ex-Vessel Corium Debris Bed Formation in Severe Accident

    International Nuclear Information System (INIS)

    Kim, Eunho; Park, Jin Ho; Kim, Moo Hwan; Park, Hyun Sun; Ma, Weimin; Bechta, Sevostian V.

    2014-01-01

    In Korean IVR-ERVC(In-Vessel Retention of molten corium through External Reactor Vessel Cooling) strategy, if the situation degenerates into insufficient external vessel cooling, the molten core mixture can directly erupt into the flooded cavity pool from the weakest point of the vessel. Then, FCI (molten Fuel Coolant Interaction) will fragment the corium jet into small particulates settling down to make porous debris bed on the cavity basemat. To secure the containment integrity against the MCCI (Molten Core - Concrete Interaction), cooling of the heat generating porous corium debris bed is essential and it depends on the characteristics of the bed itself. For the characteristics of corium debris bed, many previous experimental studies with simulant melts reported the heap-like shape mostly. There were also following experiments to develop the correlation for the heap-like shaped debris bed. However, recent studies started to consider the effect of the decay heat and reported some noticeable results with the two-phase flow effect on the debris bed formation. The Kyushu University and JAEA group reported the experimental studies on the 'self-leveling' effect which is the flattening effect of the particulate bed by the inside gas generation. The DECOSIM simulation study of RIT (Royal Institute of Technology, Sweden) with Russian researchers showed the 'large cavity pool convection' effect, which is driven by the up-rising gas bubble flow from the pre-settled debris bed, on the particle settling trajectories and ultimately final bed shape. The objective of this study is verification of the two-phase flow effect on the ex-vessel corium debris bed formation in the severe accident. From the analysis on the test movie and resultant particle beds, the two-phase flow effect on the debris bed formation, which has been reported in the previous studies, was verified and the additional findings were also suggested. For the first, in quiescent pool the

  15. Peer review of the Three Mile Island Unit 2 Vessel Investigation Project metallurgical examinations

    Energy Technology Data Exchange (ETDEWEB)

    Bohl, R.W.; Gaydos, R.G.; Vander Voort, G.F.; Diercks, D.R. [Argonne National Lab., IL (United States)

    1994-07-01

    Fifteen samples recovered from the lower head of the Three Mile Island (TMI) Unit 2 nuclear reactor pressure vessel were subjected to detailed metallurgical examinations by the Idaho National Engineering Laboratory (INEL), with supporting work carried out by Argonne National Laboratory (ANL) and several of the European participants. These examinations determined that a portion of the lower head, a so-called elliptical ``hot spot`` measuring {approx}0.8 {times} 1 m, reached temperatures as high as 1100{degrees}C during the accident and cooled from these temperatures at {approx}10--100{degrees}C/min. The remainder of the lower head was found to have remained below the ferrite-toaustenite transformation temperature of 727{degrees}C during the accident. Because of the significance of these results and their importance to the overall analysis of the TMI accident, a panel of three outside peer reviewers, Dr. Robert W. Bohl, Mr. Richard G. Gaydos, and Mr. George F. Vander Voort, was formed to conduct an independent review of the metallurgical analyses. After a thorough review of the previous analyses and examination of photo-micrographs and actual lower head specimens, the panel determined that the conclusions resulting from the INEL study were fundamentally correct. In particular, the panel reaffirmed that four lower head samples attained temperatures as high as 1100{degrees}C, and perhaps as high as 1150--1200{degrees}C in one case, during the accident. They concluded that these samples subsequently cooled at a rate of {approx}50--125{degrees}C/min in the temperature range of 600--400{degrees}C, in good agreement with the original analysis. The reviewers also agreed that the remainder of the lower head samples had not exceeded the ferrite-to-austenite transformation temperature during the accident and suggested several refinements and alternative procedures that could have been employed in the original analysis.

  16. Peer review of the Three Mile Island Unit 2 Vessel Investigation Project metallurgical examinations

    International Nuclear Information System (INIS)

    Bohl, R.W.; Gaydos, R.G.; Vander Voort, G.F.; Diercks, D.R.

    1994-07-01

    Fifteen samples recovered from the lower head of the Three Mile Island (TMI) Unit 2 nuclear reactor pressure vessel were subjected to detailed metallurgical examinations by the Idaho National Engineering Laboratory (INEL), with supporting work carried out by Argonne National Laboratory (ANL) and several of the European participants. These examinations determined that a portion of the lower head, a so-called elliptical ''hot spot'' measuring ∼0.8 x 1 m, reached temperatures as high as 1100 degrees C during the accident and cooled from these temperatures at ∼10--100 degrees C/min. The remainder of the lower head was found to have remained below the ferrite-toaustenite transformation temperature of 727 degrees C during the accident. Because of the significance of these results and their importance to the overall analysis of the TMI accident, a panel of three outside peer reviewers, Dr. Robert W. Bohl, Mr. Richard G. Gaydos, and Mr. George F. Vander Voort, was formed to conduct an independent review of the metallurgical analyses. After a thorough review of the previous analyses and examination of photo-micrographs and actual lower head specimens, the panel determined that the conclusions resulting from the INEL study were fundamentally correct. In particular, the panel reaffirmed that four lower head samples attained temperatures as high as 1100 degrees C, and perhaps as high as 1150--1200 degrees C in one case, during the accident. They concluded that these samples subsequently cooled at a rate of ∼50--125 degrees C/min in the temperature range of 600--400 degrees C, in good agreement with the original analysis. The reviewers also agreed that the remainder of the lower head samples had not exceeded the ferrite-to-austenite transformation temperature during the accident and suggested several refinements and alternative procedures that could have been employed in the original analysis

  17. SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Framatome Advanced Nuclear Power, NDSI, Erlangen (Germany)

    2001-07-01

    Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)

  18. SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

    International Nuclear Information System (INIS)

    Kolev, N.I.

    2001-01-01

    Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)

  19. Emergency core cooling system for a fast reactor

    International Nuclear Information System (INIS)

    Johnson, H.G.; Madsen, R.N.

    1976-01-01

    The main heat transport system for a liquid-metal-cooled nuclear reactor is constructed with elevated piping and guard vessels or pipes around all components of the system below the elevation of the elevated piping so the head developed by the pumps at emergency motor speed will be unsufficient to lift the liquid-metal-coolant over the top of the guard tanks or pipes or out of the elevated piping in the event of a loss-of-coolant accident. In addition, inlet downcomers to the reactor vessel are contained within guard standpipes having a clearance volume as small as practicable. 4 claims, 2 drawing figures

  20. Thermal structural analysis of SST-1 vacuum vessel and cryostat assembly using ANSYS

    International Nuclear Information System (INIS)

    Santra, Prosenjit; Bedakihale, Vijay; Ranganath, Tata

    2009-01-01

    Steady state super-conducting tokamak-1 (SST-1) is a medium sized tokamak, which has been designed to produce a 'D' shaped double null divertor plasma and operate in quasi steady state (1000 s). SST-1 vacuum system comprises of plasma chamber (vacuum vessel, interconnecting rings, baking and cooling channels), and cryostat all made of SS 304L material designed to meet ultra high vacuum requirements for plasma generation and confinement. Prior to plasma shot and operation the vessel assembly is baked to 250/150 deg. C from room temperature and discharge cleaned to remove impurities/trapped gases from wall surfaces. Due to baking the non-uniform temperature pattern on the vessel assembly coupled with atmospheric pressure loading and self-weight give rise to high thermal-structural stresses, which needs to be analyzed in detail. In addition the vessel assembly being a thin shell vessel structure needs to be checked for critical buckling load caused by atmospheric and baking thermal loads. Considering symmetry of SST-1, 1/16th of the geometry is modeled for finite element (FE) analysis using ANSYS for different loading scenarios, e.g. self-weight, pressure loading considering normal operating conditions, and off-normal loads coupled with baking of vacuum vessel from room temperature 250 deg. C to 150 deg. C, buckling and modal analysis for future dynamic analysis. The paper will discuss details about SST-1 vacuum system/cryostat, solid and FE model of SST-1, different loading scenarios, material details and the stress codes used. We will also present the thermal structural results of FE analysis using ANSYS for various load cases being investigated and our observations under different loading conditions.

  1. Safety vessels for explosive fusion reactor

    International Nuclear Information System (INIS)

    Mineev, V.

    1994-01-01

    The failure of several types of geometrically similar cylindrical and spherical steel and glass fibers vessels filled with water or air was investigated when an explosive charge of TNT was detonated in the center. Vessels had radius 50-1000 mm, thickness of walls 2-20%. The detonation on TNT imitated energy release. The parameter: K = M/mf is a measure of the strength of the vessel where M is the mass of the vessel, and mf is the mass of TNT for which the vessel fails. This demanded 2-4 destroyed and nondestroyed shots. It may be showed that: K=A/σ f where σ f is the fracture stress of the material vessel, and A = const = F(energy TNT, characteristic of elasticity of vessel material). The chief results are the following: (1) A similar increase in the geometrical dimensions of steel vessels by a factor of 10 leads to the increase of parameter K in about 5 times and to decrease of failure deformation in 7 times (scale effect). (2) For glass fibers, scale effect is absent. (3) This problem is solved in terms of theory energetic scale effect. (4) The concept of TNT equivalent explosive makes it possible to use these investigations to evaluate the response of safety vessels for explosive fusion reactor

  2. Investigations on sump cooling after core melt down

    Energy Technology Data Exchange (ETDEWEB)

    Knebel, J.U. [Forschungeszentrum Karlsruhe - Technik und Umwelt Institut fuer Angewandte Thermo- und Fluiddynamik, Karlsruhe (Germany)

    1995-09-01

    This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of an optional sump cooling concept for the European Pressurized Water Reactor EPR. This concept is entirely based on passive safety features within the containment. The work is supported by the German utilities and the Siemens dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototypic conditions.

  3. Investigations on sump cooling after core melt down

    International Nuclear Information System (INIS)

    Knebel, J.U.

    1995-01-01

    This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of an optional sump cooling concept for the European Pressurized Water Reactor EPR. This concept is entirely based on passive safety features within the containment. The work is supported by the German utilities and the Siemens dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototypic conditions

  4. A finite element elastic-plastic analysis of residual stresses due to clad welding in reactor vessels

    International Nuclear Information System (INIS)

    Buchalet, C.; Riccardella, P.C.

    1972-01-01

    Residual stresses due to weld deposited cladding on the inside of a typical Westinghouse pressurized water reactor vessel are investigated using an axisymmetric finite element elastic-plastic analysis. At the beginning of the analysis, one head of the weld cladding is assumed to lie on the reactor vessel wall at melting temperature (2600degF), but in the solid phase, while the vessel remains at 300degF (preheat temperature). All material properties used in the calculations are taken as temperature-dependent. Temperature profiles are obtained in the cladding and base metal at several discrete time intervals. These temperatures profiles are used to obtain the stress distribution for the same time intervals. Residual hoop tensile stresses of approximately 25 ksi were found to exist in the cladding. Peak tensile stresses in the hoop direction occur in the base metal near the cladding interface and reach a value of 60 ksi at the end of the transient. The tensile stress decreases very rapidly through the thickness of the base metal and becomes insignificant at about two inches from the inside surface. In order to lower residual stresses, a post-weld heat treatment is performed by uniformly heating the vessel to 1100degF, holding at that temperature for a specified period of time and then cooling slowly. The analysis shows that after this treatment, the peak stresses in the base metal decrease from 60 ksi to 32 ksi, while the stress in the cladding does not change significantly. (author)

  5. Technical potential of evaporative cooling in Danish and European condition

    DEFF Research Database (Denmark)

    Pomianowski, Michal Zbigniew; Andersen, Christian Hede; Heiselberg, Per Kvols

    2015-01-01

    Evaporative cooling is a very interesting high temperature cooling solution that has potential to save energy comparing to refrigerant cooling systems and at the same time provide more cooling reliability than mechanical or natural ventilation system without cooling. Technical cooling potential...... of 5 different evaporative systems integrated in the ventilation system is investigated in this article. Annual analysis is conducted based on hourly weather data for 15 cities located in Denmark and 123 European cities. Investigated systems are direct, indirect, combinations of direct and indirect...

  6. Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

    International Nuclear Information System (INIS)

    Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

    1983-01-01

    The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

  7. Design and Construction of the NSTX Bakeout, Cooling and Vacuum Systems

    International Nuclear Information System (INIS)

    Dudek, L.E.; Kalish, M.; Gernhardt, R.; Parsells, R.F.; Blanchard, W.

    1999-01-01

    This paper will describe the design, construction and initial operation of the NSTX bakeout, water cooling and vacuum systems. The bakeout system is designed for two modes of operation. The first mode allows heating of the first wall components to 350 degrees C while the external vessel is cooled to 150 degrees C. The second mode cools the first wall to 150 degrees C and the external vessel to 50 degrees C. The system uses a low viscosity heat transfer oil which is capable of high temperature low pressure operation. The NSTX Torus Vacuum Pumping System (TVPS) is designed to achieve a base pressure of approximately 1x10 (superscript -8) Torr and to evacuate the plasma fuel gas loads in less than 5 minutes between discharges. The vacuum pumping system is capable of a pumping speed of approximately 3400 l/s for deuterium. The hardware consists of two turbo molecular pumps (TMPs) and a mechanical pump set consisting of a mechanical and a Roots blower pump. A PLC is used as the control system to provide remote monitoring, control and software interlock capability. The NSTX cooling water provides chilled, de ionized water for heat removal in the TF, OH and PF, power supplies, bus bar systems, and various diagnostics. The system provides flow monitoring via a PLC to prevent damage due to loss of flow

  8. In- and ex-vessel flooding as part of the severe accident strategy in the KERENA reactor

    International Nuclear Information System (INIS)

    Levi, P.; Fischer, M.

    2011-01-01

    Currently, AREVA NP is finalizing the basic design of the KERENA reactor, an advanced boiling water reactor with a net electric output of about 1250 MWe. The safety concept in the KERENA reactor is founded on reliable active and passive systems for water supply and heat removal. The passive systems are based on simple physics and do not require operator action. Therefore, a severe accident (SA) with core damage, caused by the subsequent and multiple failures of the safety systems, has an extremely low probability. Despite this, the KERENA design is intended to involve measures that can limit and stop the progression of the severe accident which further reduces the frequency and extent of radioactive releases into the environment. These additional measures include in-vessel and ex-vessel flooding. Flooding is intended to remove the heat from the core or from the reactor pressure vessel (RPV) and transfer it into the containment. There the heat is removed by the active RHR (residual heat removal) system or by the passive CCCs (containment cooling condensers). Both flooding measures are passive and actuated independent of each other by different signals. The study shows that the in-vessel flooding is capable of arresting the core melt progression before a large molten pool can develop. In the unlikely event that the passive in-vessel flooding cannot be actuated or fails, the core will melt and relocate into the lower head of the RPV. In this case, as a further line of defense, decay heat removal can be achieved through the RPV wall into the water in the cavity. In order to assess whether the ex-vessel cooling can ensure RPV wall integrity a dedicated thermodynamics code has been developed which considers heat transfer from the molten corium pool into the RPV wall and the resulting wall ablation. As an input for the code the stratification behavior of the oxidic and metallic phase of the molten pool is examined. In the case of a light metallic phase on top, high heat

  9. Validation of CFD modeling for VGM loss-of-forced-cooling accidents

    International Nuclear Information System (INIS)

    Wysocki, Aaron; Ahmed, Bobby; Charmeau, Anne; Anghaie, Samim

    2009-01-01

    Heat transfer and fluid flow in the VGM reactor cavity cooling system (RCCS) was modeled using Computational Fluid Dynamics (CFD). The VGM is a Russian modular-type high temperature helium-cooled reactor. In the reactor cavity, heat is removed from the pressure vessel wall through natural convection and radiative heat transfer to water-cooled vertical pipes lining the outer cavity concrete. The RCCS heat removal capability under normal operation and accident scenarios needs to be assessed. The purpose of the present study is to validate the use of CFD to model heat transfer in the VGM RCCS. Calculations were based on a benchmark problem which defines a two-dimensional temperature distribution on the pressure vessel outer wall for both Depressurized and Pressurized Loss-of-Forced Cooling events. A two-dimensional axisymmetric model was developed to determine the best numerical modeling approach. A grid sensitivity study for the air region showed that a 20 mm mesh size with a boundary layer giving a maximum y+ of 2.0 was optimal. Sensitivity analyses determined that the discrete ordinates radiative model, the k-omega turbulence model, and the ideal gas law gave the best combination for capturing radiation and natural circulation in the air cavity. A maximum RCCS pipe wall temperature of 62degC located 6 m from the top of the cavity was predicted. The model showed good agreement with previous results for both Pressurized and Depressurized Loss-of-Forced-Cooling accidents based on RCCS coolant outlet temperature, relative contributions of radiative and convective heat transfer, and RCCS heat load profiles. (author)

  10. Refinements of the radiographic cadaver injection technique for investigating minute lymphatic vessels.

    Science.gov (United States)

    Suami, Hiroo; Taylor, G Ian; O'Neill, Jennifer; Pan, Wei-Ren

    2007-07-01

    The authors previously reported a new technique with which to delineate the lymphatic vessels, using hydrogen peroxide to identify them and a lead oxide suspension to demonstrate them on radiographs. This technique provided excellent studies of the lymph vessels in human cadavers, but there was still room for improvement. Lymph collecting vessels run superficially in some regions, where they may be damaged while the surgeon is attempting to find them. Vessels smaller than 0.3 mm in diameter could not be cannulated with a 30-gauge needle, which was the smallest the authors had available, and the lead oxide suspension often blocked this cannula. The authors also encountered problems holding the cannula steady. The authors solved these problems by using a mixture of hydrogen peroxide and ink to better identify the lymphatics, an extruded glass tube instead of a metal needle to cannulate them, an agate pestle and mortar to grind the lead oxide into finer particles, powdered milk to suspend the lead oxide, and a micromanipulator to facilitate accurate and steady cannulation of the vessels. This study developed these modifications to focus on tributaries of the collecting lymphatic channels that are smaller than 0.3 mm in diameter.

  11. COMMIX analysis of AP-600 Passive Containment Cooling System

    International Nuclear Information System (INIS)

    Chang, J.F.C.; Chien, T.H.; Ding, J.; Sun, J.G.; Sha, W.T.

    1992-01-01

    COMMIX modeling and basic concepts that relate components, i.e., containment, water film cooling, and natural draft air flow systems. of the AP-600 Passive Containment Cooling System are discussed. The critical safety issues during a postulated accident have been identified as (1) maintaining the liquid film outside the steel containment vessel, (2) ensuring the natural convection in the air annulus. and (3) quantifying both heat and mass transfer accurately for the system. The lack of appropriate heat and mass transfer models in the present analysis is addressed. and additional assessment and validation of the proposed models is proposed

  12. 46 CFR 4.03-40 - Public vessels.

    Science.gov (United States)

    2010-10-01

    ... INVESTIGATIONS Definitions § 4.03-40 Public vessels. Public vessel means a vessel that— (a) Is owned, or demise... Department (except a vessel operated by the Coast Guard or Saint Lawrence Seaway Development Corporation...

  13. What is cerebral small vessel disease?

    International Nuclear Information System (INIS)

    Onodera, Osamu

    2011-01-01

    An accumulating amount of evidence suggests that the white matter hyperintensities on T 2 weighted brain magnetic resonance imaging predict an increased risk of dementia and gait disturbance. This state has been proposed as cerebral small vessel disease, including leukoaraiosis, Binswanger's disease, lacunar stroke and cerebral microbleeds. However, the concept of cerebral small vessel disease is still obscure. To understand the cerebral small vessel disease, the precise structure and function of cerebral small vessels must be clarified. Cerebral small vessels include several different arteries which have different anatomical structures and functions. Important functions of the cerebral small vessels are blood-brain barrier and perivasucular drainage of interstitial fluid from the brain parenchyma. Cerebral capillaries and glial endfeet, take an important role for these functions. However, the previous pathological investigations on cerebral small vessels have focused on larger arteries than capillaries. Therefore little is known about the pathology of capillaries in small vessel disease. The recent discoveries of genes which cause the cerebral small vessel disease indicate that the cerebral small vessel diseases are caused by a distinct molecular mechanism. One of the pathological findings in hereditary cerebral small vessel disease is the loss of smooth muscle cells, which is an also well-recognized finding in sporadic cerebral small vessel disease. Since pericytes have similar character with the smooth muscle cells, the pericytes should be investigated in these disorders. In addition, the loss of smooth muscle cells may result in dysfunction of drainage of interstitial fluid from capillaries. The precise correlation between the loss of smooth muscle cells and white matter disease is still unknown. However, the function that is specific to cerebral small vessel may be associated with the pathogenesis of cerebral small vessel disease. (author)

  14. Experimental investigation on the feasibility of heat pipe cooling for HEV/EV lithium-ion battery

    International Nuclear Information System (INIS)

    Tran, Thanh-Ha; Harmand, Souad; Desmet, Bernard; Filangi, Sebastien

    2014-01-01

    In this paper, the use of flat heat pipe as an effective and low-energy device to mitigate the temperature of a battery module designed for a HEV application was investigated. For this purpose, nominal heat flux generated by a battery module was reproduced and applied to a flat heat pipe cooling system. The thermal performance of the flat heat pipe cooling system was compared with that of a conventional heat sink under various cooling conditions and under several inclined positions. The results show that adding heat pipe reduced the thermal resistance of a common heat sink of 30% under natural convection and 20% under low air velocity cooling. Consequently, the cell temperature was kept below 50 °C, which cannot be achieved using heat sink. According to the space allocated for the battery pack in the vehicle, flat heat pipe can be used in vertical or horizontal position. Furthermore, flat heat pipe works efficiently under different grade road conditions. The transient behaviour of the flat heat pipe was also studied under high frequency and large amplitude variable input power. The flat heat pipe was found to handle more efficiently instant increases of the heat flux than the conventional heat sink. -- Highlights: • Constant heat flux was applied to a flat heat pipe cooling system. • Its thermal performance was compared with that of a heat sink under several cooling conditions. • The influence of the inclination was evaluated. • The heat pipe transient behaviour was also studied under variable input power. • Heat pipe was found to be an effective and low-energy solution for HEV/EV battery cooling

  15. A structure for the protection of nuclear-reactor pressurized-vessels against rupture

    International Nuclear Information System (INIS)

    Marcellin, J.-P.; Aubert, Gilles

    1974-01-01

    Description is given of a structure for the protection of nuclear-reactor pressurized-vessels against rupture. Said structure comprises a pre-stressed concrete tank adapted to surround the tank side-wall and bottom, said tank being higher than said vessel, said tank being provided with ports for passing cooling fluid ducts therethrough, and a crown adapted to rest along the periphery of the reactor-cover and made integral therewith. This can be applied to reactors of the PWR type [fr

  16. Convection in complex shaped vessel; Convection dans des enceintes de forme complexe

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    The 8 november 2000, the SFT (Societe Francaise de Thermique) organized a technical day on the convection in complex shaped vessels. Nine papers have been presented in the domains of the heat transfers, the natural convection, the fluid distribution, the thermosyphon effect, the steam flow in a sterilization cycle and the transformers cooling. Eight papers are analyzed in ETDE and one paper dealing with the natural convection in spent fuels depository is analyzed in INIS. (A.L.B.)

  17. Metallographic post-test investigations for the scaled core-meltdown-experiments FOREVER-1 and -2

    International Nuclear Information System (INIS)

    Mueller, G.; Boehmert, J.

    2000-08-01

    FOREVER (Failure Of Reactor Vessel Rentention) experiments have been carried out in order to simulate the behaviour the lower head of a reactor pressure vessel under the conditions of a depressurized core melt down scenario. In particular the creep behaviour and the vessel failure mode have been investigated. Metallographic post test investigations have complemented the experimental programme. Samples of different height positions of the vessel of the FOREVER-C1 and -C2 experiments were metallographically examined and characteristic microstructural appearances were identified. Additionally samples with ineffected microstructure were annealed at different temperatures and cooled by different rates and afterwards investigated. In this way the microstructural effects of the temperature regime, the thermomechanical loads and the environmental attack could be characterized. Remarkable effects were characteristic for the FOREVER-C2 experiment where the highest-loaded region below the welding joint reached temperatures of approx. 1100 C and a strong creep damage occurred. In the FOREVER-C1 experiment creep damage could not be observed and the maximum temperature did not exceed 900 C. Environmental attack generated decarburization and oxidation but the effect was restricted to a narrow surface layer. There was almost no chemical interaction between the oxidic melt and the vessel material. (orig.)

  18. CFD investigations of natural circulation between the RPV and the cooling pond of VVER-440 type reactors in incidental conditions during maintenance performed with the code CFX-4.3

    International Nuclear Information System (INIS)

    Legradi, G.; Aszodi, A.

    2002-01-01

    During the annual maintenance of the VVER-440 type reactors, the RPV, the cooling pond and the transfer pond form a connected flow domain. The reactor is cooled by the natural circulation, which develops in one or two main loops. The cooling pond has its own cooling loops. CFD calculations have been performed with the CFX-4.3 code to investigate whether it is possible to cool the reactor core in case the main loops are lost and other emergency systems are not available. The results point out that the cooling system of the cooling pond is not capable of cooling the reactor core with the present connection. Therefore, modifications of the cooling system are investigated which would make it suitable for removing the remanent heat from the core.(author)

  19. Design, Analysis and R&D of the EAST In-Vessel Components

    Science.gov (United States)

    Yao, Damao; Bao, Liman; Li, Jiangang; Song, Yuntao; Chen, Wenge; Du, Shijun; Hu, Qingsheng; Wei, Jing; Xie, Han; Liu, Xufeng; Cao, Lei; Zhou, Zibo; Chen, Junling; Mao, Xinqiao; Wang, Shengming; Zhu, Ning; Weng, Peide; Wan, Yuanxi

    2008-06-01

    In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R&D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m3/s pumping rate at a pressure of 10-1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event (VDE) will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 °C. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.

  20. Experimental and Numerical Investigations of Air Cooling for a Large-Scale Motor

    Directory of Open Access Journals (Sweden)

    Chih-Chung Chang

    2009-01-01

    Full Text Available This article experimentally and numerically investigates the thermal performance of a 2350-kW completely enclosed motor, which is cooled through an air-to-air heat exchanger. The air in the heat exchanger includes external and internal flow paths. The external air driven by the rotation of the centrifugal fan goes through the heat exchanger mounted on the top of the frame. The internal air absorbs heat released from the stator and the rotor and then transfers the heat to the heat exchanger through the motion of two axial fans and the rotor. Several test rigs have been set up to measure the performance of the fan and the motor. The Fluent software package is adopted to analyze the complicated thermal-fluid interactions among the centrifugal fan, two axial fans, heat exchanger, stator, and rotor. The measured data, including the fan performance curves and the temperature profiles of the heat exchanger and the stator, show good agreement with the simulated results. The numerical calculations also show that the nonuniform external flow distribution through the heat exchanger and the air leakage between the axial fan and the rotor reduces the cooling ability of the motor. A detailed discussion is also included to improve the motor cooling performance.

  1. Analysis of the accident with the coolant discharge into the plasma vessel of the W7-X fusion experimental facility

    Energy Technology Data Exchange (ETDEWEB)

    Ušpuras, E.; Kaliatka, A.; Kaliatka, T., E-mail: tadas@mail.lei.lt

    2013-06-15

    Highlights: • The accident with water ingress into the plasma vessel in Wendelstein nuclear fusion device W7-X was analyzed. • The analysis of the processes in the plasma vessel and ventilation system was performed using thermal-hydraulic RELAP5 Mod3.3 code. • The suitability of pressure increase prevention system was assessed. • All analyses results will be used for the optimization of W7-X design and to ensure safe operation of this nuclear fusion device. -- Abstract: Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Starting 2007, Lithuanian Energy Institute (LEI) is a member of European Fusion Development Agreement (EFDA) organization. LEI is cooperating with Max Planck Institute for Plasma Physics (IPP, Germany) in the frames of EFDA project by performing safety analysis of fusion device W7-X. Wendelstein 7-X (W7-X) is an experimental stellarator facility currently being built in Greifswald, Germany, which shall demonstrate that in the future energy could be produced in such type of fusion reactors. In this paper the safety analysis of 40 mm inner diameter coolant pipe rupture in cooling circuit and discharge of steam–water mixture through the leak into plasma vessel during the W7-X no-plasma “baking” operation mode is presented. For the analysis the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers) and plasma vessel was developed by employing system thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code. This paper demonstrated that the developed RELAP5 model enables to analyze the processes in divertor cooling system and plasma vessel. The results of analysis demonstrated that the proposed burst disc, connecting the plasma vessel with venting system, opens and pressure inside plasma vessel does not exceed the limiting 1.1 × 10{sup 5} Pa absolute pressure. Thus, the plasma vessel remains intact after loss

  2. Nuclear reactor core support incorporating also a cooling fluid flow system

    International Nuclear Information System (INIS)

    Pennell, W.E.

    1975-01-01

    A description is given of a core bearing plate with several modular intake units having cooling fluid intake openings on their lower extensions, and on their upper ends located above the bearing plate, at least one fuel assembly which is thus in communication with the area under the bearing plate through the modular intake unit. The means for introducing the cooling fluid into the reactor vessel area are located under the bearing plate. The lower ends of the modular intake have ribs arranged essentially on a plane and join together with openings provided between the seals, in such a manner that the ribs form a barrier. The cooling fluid intake openings are located above this barrier, so that the cooling fluid is compelled to cross it before penetrating into the modular intake units [fr

  3. Investigation of an Alternative Fuel Form for the Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

    Much of the recent studies investigating the use of liquid salts as reactor coolants have utilized a core configuration of graphite prismatic fuel block assemblies with TRISO particles embedded into cylindrical fuel compacts arranged in a triangular pitch lattice. Although many calculations have been performed for this fuel form in gas cooled reactors, it would be instructive to investigate whether an alternative fuel form may yield improved performance for the liquid salt-cooled Very High Temperature Reactor (LS-VHTR). This study investigates how variations in the fuel form will impact the performance of the LS-VHTR during normal and accident conditions and compares the results with a similar analysis that was recently completed for a LS-VHTR core made up of prismatic block fuel. (author)

  4. Summary of ORNL high-temperature gas-cooled reactor program

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-01-01

    Oak Ridge National Laboratory (ORNL) efforts on the High-Temperature Gas-Cooled Reactor (HTGR) Program have been on HTGR fuel development, fission product and coolant chemistry, prestressed concrete reactor vessel (PCRV) studies, materials studies, graphite development, reactor physics and shielding studies, application assessments and evaluations and selected component testing

  5. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    Energy Technology Data Exchange (ETDEWEB)

    Urbonavičius, E., E-mail: Egidijus.Urbonavicius@lei.lt; Povilaitis, M., E-mail: Mantas.Povilaitis@lei.lt; Kontautas, A., E-mail: Aurimas.Kontautas@lei.lt

    2015-11-15

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  6. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    International Nuclear Information System (INIS)

    Urbonavičius, E.; Povilaitis, M.; Kontautas, A.

    2015-01-01

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  7. Lay-out of the He-cooled solid breeder model B in the European power plant conceptual study

    International Nuclear Information System (INIS)

    Hermsmeyer, S.; Malang, S.; Fischer, U.; Gordeev, S.

    2003-01-01

    The European helium cooled pebble bed (HCPB) blanket concept is the basis for one of two limited-extrapolation plant models that are being elaborated within the European power plant conceptual study (PPCS). In addition to addressing the case for fusion safety and environmental compatibility, following earlier studies like SEAFP or SEAL, this reactor study puts emphasis on plant availability and economic viability, which are closely related to specific plant models and require a detailed lay-out of the fusion power core and a consideration of the overall plant (balance of plant). Within the development of in-vessel components for the plant model, the major tasks to be carried out were: (i) adaptation of the HCPB concept--featuring separate pebble beds of ceramic breeder and Beryllium neutron multiplier and reduced-activation ferritic-martensitic steel EUROFER as structural material--to the large module segmentation chosen for reasons of plant availability in part II of the PPCS; (ii) proposal of a concept for a Helium cooled divertor compatible with a maximum of 10 MW/m 2 heat flux to satisfy the requirements of reasonably extrapolated plasma physics; (iii) lay-out of the major plant model components and integration into the in-vessel dimensions found from system code calculations for a power plant of 1500 MW electrical output and iterated data on the plant model performance. The paper defines all major in-vessel components of plant model B, as it is called in the PPCS, namely (i) the unit of FW, blanket and high temperature shield that is to be replaced regularly; (ii) the low temperature shield that is laid out as a lifetime component of the reactor; (iii) the divertor; and (iv) the in-vessel manifolding. Results are presented for the thermal-hydraulic performance of the components and for the thermal-mechanical behaviour of the blanket and the divertor target plate. These results suggest, together with results from the wider exploration of the plant model within

  8. Performance experiments on the in-vessel core catcher during severe accidents

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Park, Rae Joon; Cho, Young Rho; Kim, Sang Baik

    2004-01-01

    A US-Korean International Nuclear Energy Research Initiative (INERI) project has been initiated by the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korean Atomic Energy Research Institute (KAERI) to determine if IVR is feasible for high power reactors up to 1500 MWe by investigating the performance of enhanced ERVC and in-vessel core catcher. This program is initially focusing on the Korean Advanced Power Reactor 1400 MWe (APR1400) design. As for the enhancement of the coolability through the ERVC, boiling tests are conducted by using appropriate coating material on the vessel outer surface to promote downward facing boiling and selecting an improved vessel/insulation design to facilitate water flow and steam venting through the insulation in this program. Another approach for successful IVR are investigated by applying the in-vessel core catcher to provide an 'engineered gap' between the relocated core materials and the water-filled reactor vessel and a preliminary design for an in-vessel core catcher was developed during the first year of this program. Feasibility experiments using the LAVA facility, named LAVA-GAP experiments, are in progress to investigate the core catcher performance based on the conceptual design of the in-vessel core catcher proposed in this INERI project. The experiments were performed using 60kg of Al 2 O 3 thermite melt as a core material simulant with a 1/8 linear scale mock-up of the reactor vessel lower plenum. The hemispherical in-vessel core catcher was installed inside the lower head vessel maintaining a uniform gap of 10mm from the inner surface of the lower head vessel. Two types of the core catchers were used in these experiments. The first one was a single layered in-vessel core catcher without internal coating and the second one was a two layered in-vessel core catcher with an internal coating of 0.5mm-thick ZrO 2 via the plasma

  9. Chooz A: a model for the dismantling of water-cooled reactors

    International Nuclear Information System (INIS)

    Anon.

    2017-01-01

    The specificity of Chooz-A, the first French pressurized water reactor (PWR), is that the reactor and its major components (pumps, exchangers and cooling circuits) are installed in 2 caves dug out in a hill slope. Chooz-A was operating from 1967 to 1991, in 1993 the fuel was removed and in 2007 EDF received the authorization to dismantle the reactor. In 2012, EDF completed the dismantling of the cave containing the elements of the cooling circuit, a cornerstone was the removing of the four 14 m high steam generators. The dismantling of the pressure vessel began in march 2017, it is the same tools and the same processes that were used for the dismantling of the pressure vessel of the Zorita plant (Spain) in 2016. The end of the Chooz-A dismantling is expected in 2022. The feedback experience will help to standardize practices for the French fleet of PWRs. (A.C.)

  10. Estimation on the Pressure Loss of the Conceptual Primary Cooling System and Design of the Primary Cooling Pump for a Research Reactor

    International Nuclear Information System (INIS)

    Seo, Kyoung Woo; Oh, Jae Min; Park, Jong Hark; Chae, Hee Taek; Seo, Jae Kwang; Park, Cheon Tae; Yoon, Ju Hyeon; Lee, Doo Jeong

    2009-01-01

    A new conceptual primary cooling system (PCS) for a research reactor has been designed for an adequate cooling to the reactor core which has various powers ranging from 30MW through 80MW. The developed primary cooling system consisted of decay tanks, pumps, heat exchangers, vacuum breakers, some isolation and check valves, connection piping, and instruments. Because the system flow rate should be determined by the thermal hydraulic design analysis for the core, the heads to design the primary cooling pumps (PCPs) in a PCS will be estimated by the variable system flow rates. The heads of the part of a research reactor vessel was evaluated by the previous study. The various pressure losses of the PCS can be calculated by the dimensional analysis of the pipe flow and the head loss coefficient of the components. The purpose of this research is to estimate the various pressure losses and to design the PCPs

  11. Reactor-core isolation cooling system with dedicated generator

    International Nuclear Information System (INIS)

    Nazareno, E.V.; Dillmann, C.W.

    1992-01-01

    This patent describes a nuclear reactor complex. It comprises a dual-phase nuclear reactor; a main turbine for converting phase-conversion energy stored by vapor into mechanical energy for driving a generator; a main generator for converting the mechanical energy into electricity; a fluid reservoir external to the reactor; a reactor core isolation cooling system with several components at least some of which require electrical power. It also comprises an auxiliary pump for pumping fluid from the reservoir into the reactor pressure vessel; an auxiliary turbine for driving the pump; control means for regulating the rotation rate of the auxiliary turbine; cooling means for cooling the control means; and an auxiliary generator coupled to the auxiliary turbine for providing at least a portion of the electrical power required by the components during a blackout condition

  12. A modified Seeded Region Growing algorithm for vessel segmentation in breast MRI images for investigating the nature of potential lesions

    Science.gov (United States)

    Glotsos, D.; Vassiou, K.; Kostopoulos, S.; Lavdas, El; Kalatzis, I.; Asvestas, P.; Arvanitis, D. L.; Fezoulidis, I. V.; Cavouras, D.

    2014-03-01

    The role of Magnetic Resonance Imaging (MRI) as an alternative protocol for screening of breast cancer has been intensively investigated during the past decade. Preliminary research results have indicated that gadolinium-agent administrative MRI scans may reveal the nature of breast lesions by analyzing the contrast-agent's uptake time. In this study, we attempt to deduce the same conclusion, however, from a different perspective by investigating, using image processing, the vascular network of the breast at two different time intervals following the administration of gadolinium. Twenty cases obtained from a 3.0-T MRI system (SIGNA HDx; GE Healthcare) were included in the study. A new modification of the Seeded Region Growing (SRG) algorithm was used to segment vessels from surrounding background. Delineated vessels were investigated by means of their topology, morphology and texture. Results have shown that it is possible to estimate the nature of the lesions with approximately 94.4% accuracy, thus, it may be claimed that the breast vascular network does encodes useful, patterned, information, which can be used for characterizing breast lesions.

  13. Design considerations for economically competitive sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Zhang, Hongbin; Zhao, Haihua; Mousseau, Vincent; Szilard, Ronaldo

    2009-01-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phenix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design. (author)

  14. Report of the NII investigation into allegations concerning quality control during the construction of Heysham 2 power station

    International Nuclear Information System (INIS)

    1987-01-01

    The Nuclear Installations Inspectorate have investigated allegations of poor quality control in the manufacture and installation of pipework carrying cooling water for the reactor vessel and various auxiliary systems and gas, mainly carbon dioxide, for treatment. After considerable investigation of each allegation it was concluded that none provided a cause for concern over the safety standards at Heysham-2 reactor. (U.K.)

  15. INVESTIGATION OF THE PERFORMANCE OF AN ATMOSPHERIC COOLING TOWER USING FRESH AND SALTED WATER

    Directory of Open Access Journals (Sweden)

    A Haddad

    2012-01-01

    Full Text Available Cooling towers are extensively used to evacuate large quantities of heat at modest temperatures through a change of phase of the flowing cooling fluid. Based on this classical principle, the present study investigates the influence of salty water on the heat exchange produced. For that purpose, experiments are carried out using fresh and salty water. Furthermore, a comparison with the results produced through an approach involving the solution of energy equation involving the flow of air on an evaporating film of fluid. The detailed results show a preponderance of fresh water over the salty.

  16. Numerical and experimental investigation of thermoelectric cooling in down-hole measuring tools; a case study

    Directory of Open Access Journals (Sweden)

    Rohitha Weerasinghe

    2017-09-01

    Full Text Available Use of Peltier cooling in down-hole seismic tooling has been restricted by the performance of such devices at elevated temperatures. Present paper analyses the performance of Peltier cooling in temperatures suited for down-hole measuring equipment using measurements, predicted manufacturer data and computational fluid dynamic analysis. Peltier performance prediction techniques is presented with measurements. Validity of the extrapolation of thermoelectric cooling performance at elevated temperatures has been tested using computational models for thermoelectric cooling device. This method has been used to model cooling characteristics of a prototype downhole tool and the computational technique used has been proven valid.

  17. Thermal Hydraulic Analysis of RPV Support Cooling System for HTGR

    International Nuclear Information System (INIS)

    Min Qi; Wu Xinxin; Li Xiaowei; Zhang Li; He Shuyan

    2014-01-01

    Passive safety is now of great interest for future generation reactors because of its reduction of human interaction and avoidance of failures of active components. reactor pressure vessel (RPV) support cooling system (SCS) for high temperature gas-cooled reactor (HTGR) is a passive safety system and is used to cool the concrete seats for the four RPV supports at its bottom. The SCS should have enough cooling capacity to ensure the temperature of the concrete seats for the supports not exceeding the limit temperature. The SCS system is composed of a natural circulation water loop and an air cooling tower. In the water loop, there is a heat exchanger embedded in the concrete seat, heat is transferred by thermal conduction and convection to the cooling water. Then the water is cooled by the air cooler mounted in the air cooling tower. The driving forces for water and air are offered by the density differences caused by the temperature differences. In this paper, the thermal hydraulic analysis for this system was presented. Methods for decoupling the natural circulation and heat transfer between the water loop and air flow were introduced. The operating parameters for different working conditions and environment temperatures were calculated. (author)

  18. State-of-the-Art Report on Molten Corium Concrete Interaction and Ex-Vessel Molten Core Coolability

    International Nuclear Information System (INIS)

    Bonnet, Jean-Michel; Cranga, Michel; Vola, Didier; Marchetto, Cathy; Kissane, Martin; ); Robledo, Fernando; Farmer, Mitchel T.; Spengler, Claus; Basu, Sudhamay; Atkhen, Kresna; Fargette, Andre; Fisher, Manfred; Foit, Jerzi; Hotta, Akitoshi; Morita, Akinobu; Journeau, Christophe; Moiseenko, Evgeny; Polidoro, Franco; Zhou, Quan

    2017-01-01

    Activities carried out over the last three decades in relation to core-concrete interactions and melt coolability, as well as related containment failure modes, have significantly increased the level of understanding in this area. In a severe accident with little or no cooling of the reactor core, the residual decay heat in the fuel can cause the core materials to melt. One of the challenges in such cases is to determine the consequences of molten core materials causing a failure of the reactor pressure vessel. Molten corium will interact, for example, with structural concrete below the vessel. The reaction between corium and concrete, commonly referred to as MCCI (molten core concrete interaction), can be extensive and can release combustible gases. The cooling behaviour of ex-vessel melts through sprays or flooding is also complex. This report summarises the current state of the art on MCCI and melt coolability, and thus should be useful to specialists seeking to predict the consequences of severe accidents, to model developers for severe-accident computer codes and to designers of mitigation measures

  19. Argon in hornblende, biotite and muscovite in geologic cooling - Ar-40/Ar-39-investigations

    International Nuclear Information System (INIS)

    Rittman, K.L.

    1984-01-01

    The results of the Ar-40/Ar-39 studies are discussed under the aspect of whether the age data of the minerals indicate a cooling process. The author hopes that isotope dating of minerals with different closing temperatures will describe the temperature/time history of an area in the temperature range of 600 to 200 0 C. The findings are analyzed under three aspects: How much do they contribute to the initial methodological question, what do they contribute to the regional geology of the areas investigated, and in what respects do they extent the present knowledge of the geochronological analysis, i.e. its techniques and interpretation. (orig.) [de

  20. Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

    International Nuclear Information System (INIS)

    Park, Jong Woon

    2009-01-01

    This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal-hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.

  1. In-vessel core debris retention through external flooding of the reactor pressure vessel. SCDAP/RELAP5 assessment for the SBWR lower head

    International Nuclear Information System (INIS)

    Heel, A.M.J.M. van.

    1995-09-01

    In this report the results are discussed from various analyses on the feasibility and phenomenology of the External Flooding (EF) concept for an SBWR lower head, filled with a large heat generating corium mass. In applying External Flooding as an accident management strategy after or during core melt down, the lower drywell is filled with water up to a level where a large portion of the Reactor Pressure Vessel (RPV) is flooded. The purpose of this method is to establish cooling of the vessel wall, that is challenged by the heat load resulting from the corium, in such a way that its structural integrity if not endangered. The analysis discussed in this report focus on the thermal response of the vessel wall and the ex-vessel boiling processes under the conditions described above. For these analyses the SCDAP/REALP5 MOD 3.1 code was used. The major outcome of the calculations is, that a major part of the vessel wall remains well below themelting temperature of carbon steel, as long as flooding of the external surface of the lower head is established. The SCDAP/RELAP5 analyses indicated that low-quality Critical Heat Flux (CHF) was not exceeded, under all the conditions that had been tested. However, a comaprison of the heat fluxes, as calculated in RELAP5, with the CHF values obtained from the Zuber correlation and the Vishnev correction factor (for boiling at inclined surfaces) proved that CHF values, based on these criteria, were exceeded in several surface points of the lower head mesh. The correlations, as resident in the current version of RELAP5 MOD 3.1, might lead to over-estimation of CHF for the EF analyses discussed in this report. The use of the more conservative Zuber correlation with the Vishnev correction factor is recommended for EF analyses. (orig.)

  2. Ground experimental investigations into an ejected spray cooling system for space closed-loop application

    Directory of Open Access Journals (Sweden)

    Zhang Hongsheng

    2016-06-01

    Full Text Available Spray cooling has proved its superior heat transfer performance in removing high heat flux for ground applications. However, the dissipation of vapor–liquid mixture from the heat surface and the closed-loop circulation of the coolant are two challenges in reduced or zero gravity space environments. In this paper, an ejected spray cooling system for space closed-loop application was proposed and the negative pressure in the ejected condenser chamber was applied to sucking the two-phase mixture from the spray chamber. Its ground experimental setup was built and experimental investigations on the smooth circle heat surface with a diameter of 5 mm were conducted with distilled water as the coolant spraying from a nozzle of 0.51 mm orifice diameter at the inlet temperatures of 69.2 °C and 78.2 °C under the conditions of heat flux ranging from 69.76 W/cm2 to 311.45 W/cm2, volume flow through the spray nozzle varying from 11.22 L/h to 15.76 L/h. Work performance of the spray nozzle and heat transfer performance of the spray cooling system were analyzed; results show that this ejected spray cooling system has a good heat transfer performance and provides valid foundation for space closed-loop application in the near future.

  3. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    International Nuclear Information System (INIS)

    Corradin, Michael; Dominguez, A.; Tokuhiro, Akira; Hamman, K.

    2014-01-01

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  4. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  5. Summary and implications of out-of-pile investigations of local cooling disturbances in LMFBR subassembly geometry under single-phase and boiling conditions

    International Nuclear Information System (INIS)

    Huber, F.; Peppler, W.

    1985-05-01

    The consequences of local cooling disturbances in subassemblies of LMFBRs have been investigated out-of-pile at KfK. Flow and temperature distributions in the disturbed region as well as cooling under boiling conditions up to loss of cooling were investigated. Fission gas release was simulated by gas injection. A total of 16 different blockages in 20 test set-ups were used, four of them under sodium and the rest under water conditions. Mainly planar plates of different sizes and arrangements were used as blockages. In some of the experiments performed in water also porous blockages were investigated. The test sections consisted of electrically heated pin bundles with a thermal-hydraulic characteristic corresponding to that of an SNR 300 subassembly. With different parameter settings the single-phase tests in water furnished a multitude of test results on flow and temperature fields and on the behaviour of gas in the recirculation zone. In the experiments involving boiling two boiling patterns were observed: steady-state boiling and oscillating boiling. With increasing boiling intensity the boiling region grew to some extent, but it remained always confined to the blocked zone because of the relatively cold sodium flow around this zone. In the experiments simulating fission gas release it was found that under certain conditions gas accumulates in the reverse flow region behind a blockage and leads to loss of cooling. (orig./GL) [de

  6. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    Improvements in the design of the thermally insulating material used to shield the concrete containment walls in liquid metal cooled nuclear reactors are described in detail. The insulating material is composed of two layers and is placed between the primary vessel (usually steel) and the steel lined concrete containment vault. The outer layer, which clads the inner wall surface of the vault, is generally impervious to liquid metal coolant whilst the inner layer is pervious to the coolant. In normal operation, both layers protect the concrete from heat radiated from the reactor. In the event of a breach of the containment vessel, the resulting leakage of liquid metal coolant permeates the inner layer of insulating material, provides a means of heat transfer by conduction and hence reduces the overall insulating properties of the two layers. The outer layer continues to protect the wall surface of the vault from substantial direct contact with the liquid metal. Thus the two apparently conflicting requirements of good thermal insulation during normal operation and of heat transfer during loss of coolant accidents are satisfied by this novel design. Suggestions are given for possible materials for use as the insulating layers. (U.K.)

  7. Cooling rates and intensity limitations for laser-cooled ions at relativistic energies

    Science.gov (United States)

    Eidam, Lewin; Boine-Frankenheim, Oliver; Winters, Danyal

    2018-04-01

    The ability of laser cooling for relativistic ion beams is investigated. For this purpose, the excitation of relativistic ions with a continuous wave and a pulsed laser is analyzed, utilizing the optical Bloch equations. The laser cooling force is derived in detail and its scaling with the relativistic factor γ is discussed. The cooling processes with a continuous wave and a pulsed laser system are investigated. Optimized cooling scenarios and times are obtained in order to determine the required properties of the laser and the ion beam for the planed experiments. The impact of beam intensity effects, like intrabeam scattering and space charge are analyzed. Predictions from simplified models are compared to particle-in-cell simulations and are found to be in good agreement. Finally two realistic example cases of Carbon ions in the ESR and relativistic Titanium ions in SIS100 are compared in order to discuss prospects for future laser cooling experiments.

  8. Contribution for the improvement of pressurized thermal shock assessment methodologies in PWR pressure vessels

    International Nuclear Information System (INIS)

    Gomes, Paulo de Tarso Vida

    2005-01-01

    The structural integrity assessment of nuclear reactor pressure vessel, concerned to Pressurized Thermal Shock (PTS) accidents, became a necessity and has been investigated since the eighty's. The recognition of the importance of PTS assessment has led the international nuclear technology community to devote a considerable research effort directed to the complete integrity assessment process of the Reactor Pressure Vessels (VPR). Researchers in Europe, Japan and U.S.A. have concentrated efforts in the VPR structural and fracture analysis, conducting experiments to best understand how specific factors act on the behavior of discontinuities, under PTS loading conditions. The main goal of this work is to study de structural behavior of an 'in scale' PWR nuclear reactor pressure vessel model, containing actual discontinuities, under loading conditions generated by a pressurized thermal shock. To construct the pressure vessel model utilized in this research, the approach developed by Barroso (1995) and based on likelihood studies, related to thermal-hydraulic behavior during the PTS was employed. To achieve the objective of this research, a new methodology to generate cracks, with known geometry and localization in the vessel model wall was developed. Additionally, an hydraulic circuit, able to flood the vessel model, heated to 300 deg C, with 10 m 3 of water at 8 deg C, in 170 seconds, was built. Thermo-hydraulic calculations using RELAP5/M0D 3.2.2γ computational code were done, to estimate the temperature profiles during the cooling time. The resulting data subsidized the thermo-structural calculations that were accomplished using ANSYS 7.01 computational code, for both 2D and 3D models. So, the stress profiles obtained with these calculations were associated with fracture mechanics concepts, to assess the crack growth behavior in the VPR model wall. After the PTS test, the VPR model was submitted to destructive and non-destructive inspections. The results

  9. Modeling of hydronic radiant cooling of a thermally homeostatic building using a parametric cooling tower

    International Nuclear Information System (INIS)

    Ma, Peizheng; Wang, Lin-Shu; Guo, Nianhua

    2014-01-01

    Highlights: • Investigated cooling of thermally homeostatic buildings in 7 U.S. cities by modeling. • Natural energy is harnessed by cooling tower to extract heat for building cooling. • Systematically studied possibility and conditions of using cooling tower in buildings. • Diurnal ambient temperature amplitude is taken into account in cooling tower cooling. • Homeostatic building cooling is possible in locations with large ambient T amplitude. - Abstract: A case is made that while it is important to mitigate dissipative losses associated with heat dissipation and mechanical/electrical resistance for engineering efficiency gain, the “architect” of energy efficiency is the conception of best heat extraction frameworks—which determine the realm of possible efficiency. This precept is applied to building energy efficiency here. Following a proposed process assumption-based design method, which was used for determining the required thermal qualities of building thermal autonomy, this paper continues this line of investigation and applies heat extraction approach investigating the extent of building partial homeostasis and the possibility of full homeostasis by using cooling tower in one summer in seven selected U.S. cities. Cooling tower heat extraction is applied parametrically to hydronically activated radiant-surfaces model-buildings. Instead of sizing equipment as a function of design peak hourly temperature as it is done in heat balance design-approach of selecting HVAC equipment, it is shown that the conditions of using cooling tower depend on both “design-peak” daily-mean temperature and the distribution of diurnal range in hourly temperature (i.e., diurnal temperature amplitude). Our study indicates that homeostatic building with natural cooling (by cooling tower alone) is possible only in locations of special meso-scale climatic condition such as Sacramento, CA. In other locations the use of cooling tower alone can only achieve homeostasis

  10. Solar thermally driven cooling systems: Some investigation results and perspectives

    International Nuclear Information System (INIS)

    Ajib, Salman; Günther, Wolfgang

    2013-01-01

    Highlights: ► Two types of solar thermally driven absorption refrigeration machines (ARMs) have been investigated. ► We investigated the influence of the operating conditions on the effectiveness of the ARMs. ► The influence of the flow rate of the work solution on the effectiveness of the ARMs has been tested. ► Two laboratory test plants have been built and tested under different operating conditions. - Abstract: A big increase in the number of solar thermal cooling installations and research efforts could be seen over the last years worldwide. Especially the producers of solar thermal collectors and systems have been looking for thermal chillers in the small capacity range to provide air conditioning for one or two family houses. Furthermore, many developments aim to increase the efficiency of the system and to decrease the specific costs of the produced refrigeration capacity. The growth in the use of solar thermal cooling systems amounted about 860% from 52 units in 2004 to 450 units in 2009 [1]. This tendency is expected to be continuously in the next years. The practical examinations on solar thermally driven absorption machines with refrigeration capacity of 15, 10 and 5 kW have shown that this technology has a good chance to be standardized and to replace partly the conventional one. These systems can save more primary energy at high fraction of solar thermally driving by suitable control and regulation of the system. The investing costs still higher as the conventional one, however, the operating costs are less than the conventional one. The Coefficient of Performance (COP) depends on the kind of the system, work temperatures and conditions as well as the refrigeration capacity of the systems. It lies between 0.4 and 1.2. In the framework of the research on this field, we built, tested and measured two prototypes. After measuring the first prototype, the chillers were redesigned to reduce internal heat losses and make the heat and mass transfer

  11. Investigation of magnetic nanoparticle targeting in a simplified model of small vessel aneurysm

    Energy Technology Data Exchange (ETDEWEB)

    Mirzababaei, S.N. [Department of Chemical Engineering, Noshirvani Babol University of Technology, Babol (Iran, Islamic Republic of); Gorji, Tahereh B., E-mail: gorji.tahereh@stu.nit.ac.ir [Department of Mechanical Engineering, Noshirvani Babol University of Technology, Babol (Iran, Islamic Republic of); Baou, M.; Gorji-Bandpy, M. [Department of Mechanical Engineering, Noshirvani Babol University of Technology, Babol (Iran, Islamic Republic of); Fatouraee, Nasser [Department of Biomedical Engineering, Amirkabir University of Technology, Tehran (Iran, Islamic Republic of)

    2017-03-15

    An in simulacra study was conducted to investigate the capture efficiency (CE) of magnetic nanoparticles (MNPs) in aneurysm model, under the effect of a bipolar permanent magnetic system positioned at the vicinity of the model vessel. The bipolar magnetic system with an active space of 9 cm was designed by FEMM software. The MNPs were magnetite nanoparticles synthesized by the hydrothermal method which were characterized by X-ray diffraction, Fourier transform infrared spectroscopy, scanning electron microscope and magnetometer measurements. Ferrofluid velocity, magnetic field strength, and aneurysm volume all proved to be important parameters which affect the capturing of MNPs. Overall, the results of this in simulacra study confirmed the effectiveness of magnetic targeting for possible aneurysm embolization. - Highlights: • An in simulacra investigation of the magnetic targeting in mechanical aneurysm embolization was conducted. • A bipolar permanent magnetic system with an active space of 9 cm was designed by FEMM software. • Magnetic nanofluid was synthetized and applied in an experimental setup to study the effect of different flow, magnetic field and geometry parameters on the capture efficiency of the magnetic particles acting as a dug carrier agent.

  12. Increasing the maximum daily operation time of MNSR reactor by modifying its cooling system

    International Nuclear Information System (INIS)

    Khamis, I.; Hainoun, A.; Al Halbi, W.; Al Isa, S.

    2006-08-01

    thermal-hydraulic natural convection correlations have been formulated based on a thorough analysis and modeling of the MNSR reactor. The model considers detailed description of the thermal and hydraulic aspects of cooling in the core and vessel. In addition, determination of pressure drop was made through an elaborate balancing of the overall pressure drop in the core against the sum of all individual channel pressure drops employing an iterative scheme. Using this model, an accurate estimation of various timely core-averaged hydraulic parameters such as generated power, hydraulic diameters, flow cross area, ... etc. for each one of the ten-fuel circles in the core can be made. Furthermore, distribution of coolant and fuel temperatures, including maximum fuel temperature and its location in the core, can now be determined. Correlation among core-coolant average temperature, reactor power, and core-coolant inlet temperature, during both steady and transient cases, have been established and verified against experimental data. Simulating various operating condition of MNSR, good agreement is obtained for at different power levels. Various schemes of cooling have been investigated for the purpose of assessing potential benefits on the operational characteristics of the syrian MNSR reactor. A detailed thermal hydraulic model for the analysis of MNSR has been developed. The analysis shows that an auxiliary cooling system, for the reactor vessel or installed in the pool which surrounds the lower section of the reactor vessel, will significantly offset the consumption of excess reactivity due to the negative reactivity temperature coefficient. Hence, the maximum operating time of the reactor is extended. The model considers detailed description of the thermal and hydraulic aspects of cooling the core and its surrounding vessel. Natural convection correlations have been formulated based on a thorough analysis and modeling of the MNSR reactor. The suggested 'micro model

  13. Experimental investigation of a solar adsorption chiller used for grain depot cooling

    International Nuclear Information System (INIS)

    Luo, H.L.; Dai, Y.J.; Wang, R.Z.; Wu, J.Y.; Xu, Y.X.; Shen, J.M.

    2006-01-01

    The solar cooling technology is attractive since cooling load of building is roughly in phase with solar energy availability. In this study, a solar adsorption chiller was built and tested with aim of developing an alternative refrigeration system used for grain cooling storage. This solar adsorption chiller consists of four subsystems, namely, a solar water heating unit with 49.4 m 2 solar collecting area, a silica gel-water adsorption chiller, a cooling tower and a fan coil unit. In order to achieve continuous refrigeration, two adsorption units are operated out-of-phase with mass recovery cycle in the adsorption chiller. Field test results show that, under the climatic conditions of daily solar radiation being about 16-21 MJ/m 2 , this solar adsorption chiller can furnish 14-22 deg. C chilled air with an average cooling output ranging from about 3.2-4.4 kW, its daily solar cooling COP (coefficient of performance) is about 0.1-0.13

  14. Cooling water conditioning and quality control for tokamaks

    International Nuclear Information System (INIS)

    Gootgeld, A.M.

    1995-01-01

    Designers and operators of Tokamaks and all associated water cooled, peripheral equipment, are faced with the task of providing and maintaining closed-loop, low conductivity, low impurity, cooling water systems. The primary reason for supplying low conductivity water to the DIII-D vacuum vessel coils, power supplies and auxiliary heating components is to assure, along with the use of a non-conducting break in the supply piping, sufficient electrical resistance and thus an acceptable current-leakage path to ground at operating voltage potentials. As important, good quality cooling water significantly reduces the likelihood of scaling and fouling of flow passages and heat transfer surfaces. Dissolved oxygen gas removal is also required in one major DIII-D cooling water system to minimize corrosion in the ion sources of the neutral beam injectors. Currently, the combined pumping capacity of the high quality cooling water systems at DIII-D is ∼5,000 gpm. Another area that receives close attention at DIII-D is the chemical treatment of the water used in the cooling towers. This paper discusses the DIII-D water quality requirements, the means used to obtain the necessary quality and the instrumentation used for control and monitoring. Costs to mechanically and chemically condition and maintain water quality are discussed as well as the various aspects of complying with government standards and regulations

  15. Ultrasonographic Examination of Some Vessels in Dogs and the Characteristics of Blood Flow in These Vessels

    Directory of Open Access Journals (Sweden)

    Figurová M.

    2017-12-01

    Full Text Available The examination by Doppler ultrasonography provides haemodynamic information about blood flow velocity in a respective vessel. It specifies high- and lowresistance flow patterns. The aim of our study was to record the flow in a. carotis communis, a. femoralis and aa. renales in 16 adult clinically healthy dogs of small and medium size; characterize the types of vessels and also determine the pulsatility index (PI and the resistive index (RI of these vessels. The a. femoralis is a high-resistance vessel with a pronounced three-peak waveform. The aa. renales gives a typical picture of a low-resistance flow pattern. The characteristics of a. carotis communis involves different images of its branches a. carotis interna and a. carotis externa. In the investigated groups we observed a medium degree of pulsatility (atypical highresistance flow pattern with an absence of reverse flow. The mean measured values of indices for a. carotis communis were: left side PI 1.824 and RI 0.742; right side PI 1.891 and RI 0.746, and for aa. renales: PI 1.366 ± 0.04 and RI 0.684 ± 0.05.

  16. Experimental investigation of the influence of the air jet trajectory on convective heat transfer in buildings equipped with air-based and radiant cooling systems

    DEFF Research Database (Denmark)

    Le Dreau, Jerome; Heiselberg, Per; Jensen, Rasmus Lund

    2015-01-01

    -state and dynamic conditions. With the air-based cooling system, a dependency of the convective heat transfer on the air jet trajectory has been observed. New correlations have been developed, introducing a modified Archimedes number to account for the air flow pattern. The accuracy of the new correlations has been...... evaluated to±15%. Besides the study with an air-based cooling system, the convective heat transfer with a radiant cooling system has also been investigated. The convective flow at the activated surface is mainly driven by natural convection. For other surfaces, the complexity of the flow and the large......The complexity and diversity of airflow in buildings make the accurate definition of convective heat transfer coefficients (CHTCs) difficult. In a full-scale test facility, the convective heat transfer of two cooling systems (active chilled beam and radiant wall) has been investigated under steady...

  17. Investigation of Stratified Thermal Storage Tank Performance for Heating and Cooling Applications

    Directory of Open Access Journals (Sweden)

    Azharul Karim

    2018-04-01

    Full Text Available A large amount of energy is consumed by heating and cooling systems to provide comfort conditions for commercial building occupants, which generally contribute to peak electricity demands. Thermal storage tanks in HVAC systems, which store heating/cooling energy in the off-peak period for use in the peak period, can be used to offset peak time energy demand. In this study, a theoretical investigation on stratified thermal storage systems is performed to determine the factors that significantly influence the thermal performance of these systems for both heating and cooling applications. Five fully-insulated storage tank geometries, using water as the storage medium, were simulated to determine the effects of water inlet velocity, tank aspect ratio and temperature difference between charging (inlet and the tank water on mixing and thermocline formation. Results indicate that thermal stratification enhances with increased temperature difference, lower inlet velocities and higher aspect ratios. It was also found that mixing increased by 303% when the temperature difference between the tank and inlet water was reduced from 80 °C to 10 °C, while decreasing the aspect ratio from 3.8 to 1.0 increased mixing by 143%. On the other hand, increasing the inlet water velocity significantly increased the storage mixing. A new theoretical relationship between the inlet water velocity and thermocline formation has been developed. It was also found that inlet flow rates can be increased, without increasing the mixing, after the formation of the thermocline.

  18. Modeling skin cooling using optical windows and cryogens during laser induced hyperthermia in a multilayer vascularized tissue

    International Nuclear Information System (INIS)

    Singh, Rupesh; Das, Koushik; Okajima, Junnosuke; Maruyama, Shigenao; Mishra, Subhash C.

    2015-01-01

    This article deals with the spatial and the temporal evolution of tissue temperature during skin surface cooled laser induced hyperthermia. Three different skin surface cooling methodologies viz., optical window contact cooling, cryogenic spray cooling and cryogen cooled optical window contact cooling are considered. Sapphire, yttrium aluminum garnet, lithium tantalate, and magnesium oxide doped lithium niobate are the considered optical windows. The cryogens considered are liquid CO_2 and R1234yf. Heat transfer in the multilayer skin tissue embedded with thermally significant blood vessels pairs is modeled using the Pennes and Weinbaum–Jiji bioheat equations. Weinbaum–Jiji bioheat equation is used for the vascularized tissue. Laser transport in the tissue is modeled using the radiative transfer equation. Axial and radial (skin surface) temperature distributions for different combinations of optical windows and cryogens are analyzed. Liquid CO_2 cooled yttrium aluminum garnet is found to be the best surface cooling mechanism. - Highlights: • Skin surface cooled laser induced hyperthermia is studied. • A multi-layer 2-D cylindrical tissue geometry is considered. • Both Pennes and Weinbaum–Jiji bioheat models are considered. • Laser transport in the tissue is modeled using discrete ordinate method. • Results for 4 optical windows and 2 cryogens for skin cooling are presented.

  19. Design improvements and R and D achievements for VV and in-vessel components towards ITER construction

    International Nuclear Information System (INIS)

    Ioki, K.; Barabaschi, P.; Barabash, V.

    2003-01-01

    During the preparation of the procurement specifications for long lead-time items, several detailed vacuum vessel (VV) design improvements are being pursued, such as elimination of the inboard triangular support, adding a separate interspace between inner and outer shells for independent leak detection of field joints, and revising the VV support system to gain a more comfortable structural performance margin. Improvements to the blanket design are also under investigation, an inter-modular key instead of two prismatic keys and a co-axial inlet outlet cooling connection instead of two parallel pipes. One of the most important achievements in the VV R and D has been demonstration of the necessary assembly tolerances. Further development of cutting, welding and nondestructive tests (NDT) for the VV has been continued, and thermal and hydraulic tests have been performed to simulate the VV cooling conditions. In FW/blanket and divertor, full-scale prototypical mock-ups of the FW panel, the blanket shield block, and the divertor components, have been successfully fabricated. These results make us confident in the validity of our design and give us possibilities of alternate fabrication methods. (author)

  20. Experimental investigation of impingement cooling with turbulators or surface enlarging elements

    Energy Technology Data Exchange (ETDEWEB)

    Persson, Johan

    2000-02-01

    For the materials in modern gas turbines to sustain, a considerable amount of cooling is required. In cases where large amounts of heat need to be removed, impingement cooling with its high heat transfer coefficients may be the only alternative. In this work the possibilities of enhancing impingement cooling by introducing surface enlarging - turbulence enhancing elements are examined experimentally. A configuration consisting of a staggered array of 45 impingement jets distributed over 10 rows is used for the purpose. A thermo camera is used to measure the temperature distribution on the target plate, giving an opportunity to separately evaluate the Nusselt number enhancement for different areas. Experiments are conducted for five different area enlarging geometries: triangle, wing, cylinder, dashed rib, and angel, all made from aluminium. Comparison between each area enlarged surface and a flat plate is made, with results presented as Nusselt number enhancement factors. The effect of pumping power required is also investigated in order to maximize the cooling efficiency. Parameters varied are Reynolds number and jet to plate distance. Overall Nusselt number enhancement factors show values of 1 to 1.3, the trend being decreasing with increased jet to plane distance and Reynolds number. When taking into account pumping power the enhancement factors drop to 0.4 to 1.2. The best results are achieved with the rib geometry and when not using a too large value of enlarger height over jet to plate distance (h/z). Row wise evaluation of Nusselt number enhancement shows an increased enhancement with row number and thereby crossflow ratio (Gc/Gj). Typical increases in enhancement of 1 to 1.5 with Gc/Gj from 0 to 0.8 are found. The thermo camera pictures reveal that the enhancement is found in three different areas, on the enlarger base area, the area just downstream the enlarger and in diagonal streaks with increased turbulence caused by the enlargers. Tests using an