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Sample records for thermohydraulic hot pool

  1. Development of a Two-dimensional Thermohydraulic Hot Pool Model and ITS Effects on Reactivity Feedback during a UTOP in Liquid Metal Reactors

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Jeong, Hae Yong; Cho, Chung Ho; Kwon, Young Min; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong; Hahn, Do Hee

    2009-01-01

    The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect

  2. One component, volume heated, boiling pool thermohydraulics

    International Nuclear Information System (INIS)

    Bede, M.; Perret, C.; Pretrel, H.; Seiler, J.M.

    1993-01-01

    Prior work on boiling pools provided heat exchange correlations valid for bubbly flow with laminar or turbulent boundary layers. New experiments performed with water (SEBULON) and UO 2 (SCARABEE BF2) in a churn-turbulent flow configuration show unexpected heat flux distributions for which the maximum heat flux may be situated well below the pool surface. The origin of this behaviour is attributed to condensation effects, very unstable boundary layer flow and surface oscillation. A calculation model is discussed which permits to approach the experimental heat flux distribution with reasonable accuracy. (authors). 7 figs., 2 appendix., 14 refs

  3. 3-dimensional thermohydraulic analysis of KALIMER reactor pool during unprotected accidents

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Hahn Do Hee

    2003-01-01

    During a normal reactor scram, the heat generation is reduced almost instantaneously while the coolant flow rate follows the pump coastdown. This mismatch between power and flow results in a situation where the core flow entering the hot pool is at a lower temperature than the temperature of the bulk pool sodium. This temperature difference leads to thermal stratification. Thermal stratification can occur in the hot pool region if the entering coolant is colder than the existing hot pool coolant and the flow momentum is not large enough to overcome the negative buoyancy force. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response. Hence, it is necessary to predict the pool coolant temperature distribution with sufficient accuracy to determine the inlet temperature conditions for the IHXs and its contribution to the net buoyancy head. Therefore, two-dimensional hot pool thermohydraulic model named HP2D has been developed. In this report code-to-code comparison analysis between HP2D and COMMIX-1AR/P has been performed in the case of steady-state and UTOP.

  4. Thermohydraulic behavior of liquid metal pool submitted to electronic bombardment

    International Nuclear Information System (INIS)

    Brun, Patrice

    1998-01-01

    This thesis deals with the thermohydraulics of liquid metal molten by an electron beam. We study the relationship between the liquid metal pool and the vapor rate. The aim is to find good conditions increasing the metal vapor rate. In first place, energy losses are identified. Mains are convection (buoyancy and thermo-capillary) strengthen by the deformation of the molten pool. The first action is to reduce the liquid interface deformation with a transient spot realized by scanning the electron beam. I find that in this case, the optimum vapor rate is obtained when the crossing time of the beam is smaller than characteristic time of formation of the cavity, but greater than the heating time of the surface. Secondly, I impose forces to change the morphology of the flow. Two actions are tried: magnetic field application and rotating motion of the crucible. External magnetic field application may reduce convective flow, by the creation of a magnetic brake. But in my experiment, magnetic field deteriorates electron beam before to be effective. Results obtained by the rotating motion of the crucible approve this choice to reduce energy losses and increase vapor rate. This growth of vapor rate is due to an expansion of the emitted vapor source and an increase of the central temperature of the molten pool. Nevertheless with the increase of the rotation velocity and after the optimum vapor rate, I note that the flow is not axisymmetric. My observation give to think about instabilities that are developed by baroclinic waves. The comparison of my works with the Eady's linear theory gives good results. (author) [fr

  5. (Microbiological studies of small hot-bath-pools and hot-whirl-pools (author's transl))

    Energy Technology Data Exchange (ETDEWEB)

    Exner, M; Havenith, N

    1981-01-01

    Hot small bathing pools and hot whirl-pools have the following characteristics: small watervolume, thick squeeze of swimmers, high water temperature (37-40 degrees C) and small dimension of filters. By this, the quality of bathing-water is influenced detrimentally. To elaborate the hygienic problems, bathing-water samples were taken before, during and after the visiting-hours and were tested for facultative-pathogenic microorganisms. During this investigation E. coli was isolated in 25 degrees, Coliforms and Proteus species in 37.3%, P. aeruginosa in 36%, S. aureus in 26.3%, Enterococci in 42.3 %, Candida albicans in 3.6% and yeast totally in 8.3%.

  6. Thermohydraulic behaviour of the hot channel in a PWR type reactor under loss-of-coolant accident conditions (LOCA)

    International Nuclear Information System (INIS)

    Costa, J.R.

    1978-12-01

    An analysis is done of the core behavior for a 1861 MW(th) pressurized water reactor with two coolant loops, during the blowdown phase of a double-ended cold leg rupture, between the main feedwater pump, and the pressure vessel. The analysis is done through a detailed thermohydraulic study of the hot pin channel with RELAP4/MOD 5 code, including the Evaluatin Model options. The problem is solved separately for two values of discharge coefficient (C sub(D)= 1,0 and 0,4). The results show that the maximum clad temperature is lower than the limit value for licensing purposes. Concerning clad material oxidation, the maximum value obtained is also under the limit of acceptance. (author) [pt

  7. Allegheny County Public Swimming Pool, Hot Tub, and Spa Inspections

    Data.gov (United States)

    Allegheny County / City of Pittsburgh / Western PA Regional Data Center — Public swimming pool, hot tub, and spa facilities are licensed and inspected once each year to assure proper water quality, sanitation, lifeguard coverage and...

  8. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    Belem, J.A.T.

    1993-09-01

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs

  9. Radiological performance of hot water layer system in open pool type reactor

    Directory of Open Access Journals (Sweden)

    Amr Abdelhady

    2013-06-01

    Full Text Available The paper presents the calculated dose rate carried out by using MicroShield code to show the importance of hot water layer system (HWL in 22 MW open pool type reactor from the radiation protection safety point of view. The paper presents the dose rate profiles over the pool surface in normal and abnormal operations of HWL system. The results show that, in case of losing the hot water layer effect, the radiation dose rate profiles over the pool surface will increase from values lower than the worker permissible dose limits to values very higher than the permissible dose limits.

  10. Radiological performance of hot water layer system in open pool type reactor

    OpenAIRE

    Amr Abdelhady

    2013-01-01

    The paper presents the calculated dose rate carried out by using MicroShield code to show the importance of hot water layer system (HWL) in 22 MW open pool type reactor from the radiation protection safety point of view. The paper presents the dose rate profiles over the pool surface in normal and abnormal operations of HWL system. The results show that, in case of losing the hot water layer effect, the radiation dose rate profiles over the pool surface will increase from values lower than th...

  11. Theoretical and experimental studies of thermal stratification in hot and cold pools of PFBR

    International Nuclear Information System (INIS)

    Velusamy, K.; Titus, G.; Rajakumar, A.; Ravichandran, G.; Padmakumar, G.; Vaidyanathan, G.; Kale, R.D.; Chetal, S.C.; Bhoje, S.B.

    1994-01-01

    Results of experimental studies carried out in two water models of size 1/24 and 1/15, to assess the free level fluctuation in the hot pool of PFBR are presented. The results when extrapolated to the prototype gives a ripple height of 50 mm. The results of thermal stratification studies carried out in 1/24 scale model, using hot and cold water indicates that the interface velocity can be correlated with the Richardson number. The paper also gives the details of computer codes developed for the estimation of flow and temperature fields in the pools. (author)

  12. Experimental Investigation of the Hot Water Layer Effect on Upward Flow Open Pool Reactor Operability

    International Nuclear Information System (INIS)

    Abou Elmaaty, T.

    2014-01-01

    The open pool reactor offers a high degree of reliability in the handling and manoeuvring, the replacement of reactor internal components and the suing of vertical irradiation channels. The protection of both the operators and the reactor hall environment against radiation hazards is considered a matter of interest. So, a hot water layer is implemented above many of the research reactors main pool, especially those whose flow direction is upward flow. An experimental work was carried out to ensure the operability of the upward flow open pool research reactor with / without the hot water layer. The performed experiment showed that, the hot water layer is produced an inverse buoyant force make the water to diffuse downward against the ordinary natural circulation from the reactor core. An upward flow - open pool research reactor (with a power greater than 20 M watt) could not wok without a hot water layer. The high temperature of the hot water layer surface could release a considerable amount of water vapour into the reactor hall, so a heat and mass transfer model is built based on the measured hot water layer surface temperature to calculate the amount of released water vapour during the reactor operating period. The effects of many parameters like the ambient air temperature, the reactor hall relative humidity and the speed of the pushed air layer above the top pool end on the evaporation rate is studied. The current study showed that, the hot water layer system is considered an efficient shielding system against Gamma radiation for open pool upward flow reactor and that system should be operated before the reactor start up by a suitable period of time. While, the heat and mass transfer model results showed that, the amount of the released water vapour is increased as a result of both the increase in hot water layer surface temperature and the increase in air layer speed. As the increase in hot water layer surface temperature could produce a good operability

  13. Experimental Investigation of the Hot Water Layer Effect on Upward Flow Open Pool Reactor Operability

    International Nuclear Information System (INIS)

    Abou Elmaaty, T.

    2015-01-01

    The open pool reactor offers a high degree of reliability in the handling and manoeuvring, the replacement of reactor internal components and the swing of vertical irradiation channels. The protection of both the operators and the reactor hall environment against radiation hazards is considered a matter of interest. So, a hot water layer implemented above many of the research reactors main pool, especially those whose flow direction is upward flow. An experimental work was carried out to ensure the operability of the upward flow open pool research reactor with / without the hot water layer. The performed experiment showed that, the hot water layer produced an inverse buoyant force making the water to diffuse downward against the ordinary natural circulation from the reactor core. An upward flow-open pool research reactor (with a power greater than 20 Mw) could not wok without a hot water layer. The high temperature of the hot water layer surface could release a considerable amount of water vapour into the reactor hall, so a heat and mass transfer model is built based on the measured hot water layer surface temperature to calculate the amount of released water vapour during the reactor operating period. The effects of many parameters like the ambient air temperature, the reactor hall relative humidity and the speed of the pushed air layer above the top pool end on the evaporation rate is studied. The current study showed that, the hot water layer system is considered an efficient shielding system against gamma radiation for open pool upward flow reactor and that system should be operated before the reactor start up by a suitable period of time. While, the heat and mass transfer model results showed that, the amount of the released water vapour is increased as a result of both the increase in hot water layer surface temperature and the increase in air layer speed. As the increase in hot water layer surface temperature could produce a good operability conditions from

  14. Change in color of the hot spring deposits at the Chinoike-Jigoku hot pool, Beppu geothermal field

    Energy Technology Data Exchange (ETDEWEB)

    Kazuthoshi, Oue; Ohsawa, Shinji; Yusa, Yuki [Kyoto University, Beppu (Japan). Beppu Geothermal Research Laboratory, Graduate School of Science

    2002-06-01

    The Chinoike-Jigoku hot pool in Beppu geothermal field, Central Kyushu, Japan, displays a blood-red color due to the hematite (Fe{sub 2}O{sub 3}) deposited at the bottom of the pool. The colors of the deposits collected on 1 October 1990, on 27 March 1995, and on 6 March 1996 were measured with a colorimeter. The results show that the red deposits became yellower in 1995 and 1996 than they were in 1990. X-ray diffraction (XRD) patterns and chemical compositions of the deposits indicate that the discoloration of the Chinoike-Jigoku pool water is caused by an increase in the content of jarosite [KFe{sub 3}(SO{sub 4}){sub 2}(OH){sub 6}]. The temperature of the subsurface thermal water beneath the Chinoike-Jigoku hot pool, as estimated by the anhydrite chemical geothermometer, has declined from 200 to 150{sup o}C over the past 25 years. The Na and Cl concentrations of the hot spring water discharging from Chinoike-Jigoku have decreased, while the SO{sub 4} concentration has increased. The temporal variations in subsurface temperature and dissolved ion concentrations suggest that the mixing ratio between the high-temperature, neutral Na-Cl type water and the relatively low-temperature, acid H-SO{sub 4} type water that form the thermal water of Chinoike-Jigoku has changed over the last 25 years. Hydrothermal studies of jarosite stability have confirmed that the increase in jarosite content in the deposits was caused by a temperature drop of the mixed thermal water beneath Chinoike-Jigoku pool, due to an increase in the contribution of the cooler H-SO{sub 4} water type to the thermal mixture. (author)

  15. Chemical analyses of waters from geysers, hot springs, and pools in Yellowstone National Park, Wyoming from 1974 to 1978

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, J.M.; Yadav, S.

    1979-01-01

    Waters from geysers, hot springs, and pools of Yellowstone National Park have been analyzed. We report 422 complete major ion analyses from 330 different locations of geysers, hot springs, and pools, collected from 1974 to 1978. Many of the analyses from Upper, Midway, Lower, and Norris Geyser Basin are recollections of features previously reported.

  16. Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Padmakumar, G.; Pandey, G.K.; Vaidyanathan, G.

    2009-01-01

    Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained. Studies were conducted simulating both low and full power operating conditions. This paper discusses the model simulation, similarity criteria, the various thermal hydraulic studies that were carried out, the results obtained and the comparison with the prototype measurements.

  17. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  18. Thermohydraulic relationships for advanced water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  19. Thermohydraulics of reactors

    International Nuclear Information System (INIS)

    Delhaye, J.M.

    2008-01-01

    This scientific and technical handbook about PWR reactors thermohydraulics is the result of many years of teaching in the framework of the CEA-INSTN's atomic engineering training courses, in engineering schools and during continuing training activities. Its main goal is to present in a rigorous and pedagogical way the basic knowledge necessary for the understanding and modeling of single phase and two-phase thermohydraulic phenomena encountered during the design and operation of nuclear reactors. In particular, heat transfers in two-phase flows are presented in a detailed way. Most chapters include some nuclear engineering examples of application of the studied concepts, and some exercises aiming at mastering these concepts. Each example or exercise is accompanied by its detailed solution. Content: - thermohydraulic characteristics of reactors; - design and thermal dimensioning of reactors; - thermal engineering of the fuel element; - two-phase flow configurations in ducts; - recalls about single-phase flow equations; - basic equations for two-phase flows; - modeling of two-phase flows inside ducts; - pressure drops in ducts; - boiling and vapor condensation heat transfers; - two-phase flow instabilities in ducts; - two-phase flow blocking; thermohydraulics of naval propulsion reactors

  20. Justify of implementation of a hot water layer system in swimming pool research reactor IEA-R1m

    International Nuclear Information System (INIS)

    Toyoda, Eduardo Yoshio; Gordon, Ana Maria Pinho Leite; Sordi, Gian-Maria A.A.

    2001-01-01

    The IPEN/CNEN-SP has a swimming pool research reactor (IEA-R1m) in operation since 1957 at 2 MW. In 1998, after some modifications, its nominal power increased to 5 MW. Among these modifications some adaptations had to be accomplished in the radiological protection and operational procedure. The present work aim to study the need of implementation of a hot water layer in order to reduce the dose in the workers in the vicinity of the reactor swimming pool. Applying the principles of radioprotection optimization, it was concluded that the decision of the construction of one hot water layer system in the reactor swimming pool, is not necessary. (author)

  1. Enhancing the actinide sciences in Europe through hot laboratories networking and pooling: from ACTINET to TALISMAN

    International Nuclear Information System (INIS)

    Bourg, S.; Poinssot, C.

    2013-01-01

    Since 2004, Europe supports the strengthening of the European actinides sciences scientific community through the funding of dedicated networks: (i) from 2004 to 2008, the ACTINET6 network of excellence (6. Framework Programme) gathered major laboratories involved in nuclear research and a wide range of academic research organisations and universities with the specific aims of funding and implementing joint research projects to be performed within the network of pooled facilities; (ii) from 2009 to 2013, the ACTINET-I3 integrated infrastructure initiative (I3) supports the cost of access of any academics in the pooled EU hot laboratories. In this continuation, TALISMAN (Trans-national Access to Large Infrastructures for a Safe Management of Actinides) gathers now the main European hot laboratories in actinides sciences in order to promote their opening to academics and universities and strengthen the EU-skills in actinides sciences. Furthermore, a specific focus is set on the development of advanced cutting-edge experimental and spectroscopic capabilities, the combination of state-of-the art experimental with theoretical first-principle methods on a quantum mechanical level and to benefit from the synergy between the different scientific and technical communities. ACTINET-I3 and TALISMAN attach a great importance and promote the Education and Training of the young generation of actinides scientists in the Trans-national access but also by organizing Schools (general Summer Schools or Theoretical User Lab Schools) or by granting students to attend International Conference on actinide sciences. (authors)

  2. Experimental and computational analysis of the hot water layer for the radiological protection in swimming pool reactor

    International Nuclear Information System (INIS)

    Ribeiro, Rogerio.

    1995-01-01

    Pool reactors are research reactors, which allow easy access to the core and rare simple to operate. Reactors of this kind operating at power levels higher than about one megawatt need a hot water layer at the surface of the pool, in order to keep surface activity below acceptable levels and enable free access to the upper part of the reactor. An experimental apparatus was constructed to study the hot water layer stability. Thermocouples were used to measure the temperature field. A numerical analysis was conducted simultaneously. Regarding experimental results, representative temperature contour lines of the hot water layer were plotted. The temperature field was determined in the numerical analysis and temperature contour lines corresponding to those of the experimental results were plotted. The hot water layer kept stable for experimental and numerical results. Good agreement between the results for the hot water layer position and thickness has been obtained. (author). 21 refs., 40 figs., 15 tabs

  3. Film boiling heat transfer from a hot sphere falling in two-phase pool

    International Nuclear Information System (INIS)

    Bang, K. H.; Kim, K. Y.

    1998-01-01

    The purpose of the present study is to experimentally investigate film boiling heat trasfer from a hot sphere falling in steam-water two-phase pool, which is the key heat transfer mode in molten fuel and coolant mixing. To measure film boiling heat transfer coefficients on a spere falling in two-phase pool, a heated sphere with a thermocouple embedded at the center is dropped in a vertical tube filled with steam-water mixture. The present experiment is unique in making the heated sphere fall through the two-phase pool while the previous experiments were performed with stationary spheres in flowing fluid. The falling speed of the sphere is measured using a set of magnet pickup coils distributed along the tube. The ranges of the experimental conditions are: spere fall speed 0-0.5 m/s, average void fraction 0-25,% steam superficial velocity 0-0.25 m/s. The results show that the forced convection film boiling heat transfer coefficient decrease slightly as the steam superficial velocity (void fraction) is increased

  4. The TOPFLOW multi-purpose thermohydraulic test facility

    International Nuclear Information System (INIS)

    Schaffrath, Andreas; Kruessenberg, A.-K.; Weiss, F.-P.; Prasser, H.-M.

    2002-01-01

    The TOPFLOW (Transient Two Phase Flow Test Facility) multi-purpose thermohydraulic test facility is being built for studies of steady-state and transient flow phenomena in two-phase flows, and for the development and validation of the models contained in CFD (Computational Fluid Dynamics) codes. The facility is under construction at the Institute for Safety Research of the Rossendorf Research Center (FZR). It will be operated together with the Dresden Technical University and the Zittau/Goerlitz School for Technology, Economics and Social Studies within the framework of the Nuclear Technology Competence Preservation Program. TOPFLOW, with its test sections and its flexible concept, is available as an attractive facility also to users from all European countries. Experiments are planned in these fields, among others: - Transient two-phase flows in vertical and horizontal pipes and pipes of any inclination as well as in geometries typical of nuclear reactors (annulus, hot leg). - Boiling in large vessels and water pools (measurements of steam generation, 3D steam content distribution, turbulence, temperature stratification). - Test of passive components and safety systems. - Condensation in horizontal pipes in the absence and presence of non-condensable gases. The construction phase of TOPFLOW has been completed more or less on schedule. Experiments can be started after a commissioning phase in the 3rd quarter of 2002. (orig.) [de

  5. CAREM reactor thermohydraulic essays laboratory

    International Nuclear Information System (INIS)

    Horro, R.; Mazzi, R.; Rossini, A.

    1990-01-01

    The main characteristics, essays projected and the present state of the Thermohydraulic Essays Laboratory -under construction at present- prepared to meet the experimental needs resulting from a power reactor design of the CAREM type, are herein described. (Author) [es

  6. Development of two-dimensional hot pool model and analysis of the ULOHS accident in KALIMER design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Jeong, K. S.; Hahn, H. D

    2000-10-01

    In the new version of HP2D program, the variation model of the hot pool sodium level is added so that the temperature and velocity profiles can be predicted more accurately than old version. To verify and validate the developed new version model, comparison of the MONJU experimental data with the predicted one is performed and analyzed. And also the ULOHS(Unprotected Loss of Heat Sink) accident in the KALIMER design is performed and analyzed.

  7. Development of two-dimensional hot pool model and analysis of the ULOHS accident in KALIMER design

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Jeong, K. S.; Hahn, H. D.

    2000-10-01

    In the new version of HP2D program, the variation model of the hot pool sodium level is added so that the temperature and velocity profiles can be predicted more accurately than old version. To verify and validate the developed new version model, comparison of the MONJU experimental data with the predicted one is performed and analyzed. And also the ULOHS(Unprotected Loss of Heat Sink) accident in the KALIMER design is performed and analyzed

  8. Microbial contributions to coupled arsenic and sulfur cycling in the acid-sulfide hot spring Champagne Pool, New Zealand.

    Science.gov (United States)

    Hug, Katrin; Maher, William A; Stott, Matthew B; Krikowa, Frank; Foster, Simon; Moreau, John W

    2014-01-01

    Acid-sulfide hot springs are analogs of early Earth geothermal systems where microbial metal(loid) resistance likely first evolved. Arsenic is a metalloid enriched in the acid-sulfide hot spring Champagne Pool (Waiotapu, New Zealand). Arsenic speciation in Champagne Pool follows reaction paths not yet fully understood with respect to biotic contributions and coupling to biogeochemical sulfur cycling. Here we present quantitative arsenic speciation from Champagne Pool, finding arsenite dominant in the pool, rim and outflow channel (55-75% total arsenic), and dithio- and trithioarsenates ubiquitously present as 18-25% total arsenic. In the outflow channel, dimethylmonothioarsenate comprised ≤9% total arsenic, while on the outflow terrace thioarsenates were present at 55% total arsenic. We also quantified sulfide, thiosulfate, sulfate and elemental sulfur, finding sulfide and sulfate as major species in the pool and outflow terrace, respectively. Elemental sulfur concentration reached a maximum at the terrace. Phylogenetic analysis of 16S rRNA genes from metagenomic sequencing revealed the dominance of Sulfurihydrogenibium at all sites and an increased archaeal population at the rim and outflow channel. Several phylotypes were found closely related to known sulfur- and sulfide-oxidizers, as well as sulfur- and sulfate-reducers. Bioinformatic analysis revealed genes underpinning sulfur redox transformations, consistent with sulfur speciation data, and illustrating a microbial role in sulfur-dependent transformation of arsenite to thioarsenate. Metagenomic analysis also revealed genes encoding for arsenate reductase at all sites, reflecting the ubiquity of thioarsenate and a need for microbial arsenate resistance despite anoxic conditions. Absence of the arsenite oxidase gene, aio, at all sites suggests prioritization of arsenite detoxification over coupling to energy conservation. Finally, detection of methyl arsenic in the outflow channel, in conjunction with

  9. Microbial contributions to coupled arsenic and sulfur cycling in the acid-sulfide hot spring Champagne Pool, New Zealand

    Directory of Open Access Journals (Sweden)

    Katrin eHug

    2014-11-01

    Full Text Available Acid-sulfide hot springs are analogs of early Earth geothermal systems where microbial metal(loid resistance likely first evolved. Arsenic is a metalloid enriched in the acid-sulfide hot spring Champagne Pool (Waiotapu, New Zealand. Arsenic speciation in Champagne Pool follows reaction paths not yet fully understood with respect to biotic contributions and coupling to biogeochemical sulfur cycling. Here we present quantitative arsenic speciation from Champagne Pool, finding arsenite dominant in the pool, rim and outflow channel (55-75% total arsenic, and dithio- and trithioarsenates ubiquitously present as 18-25% total arsenic. In the outflow channel, dimethylmonothioarsenate comprised ≤9% total arsenic, while on the outflow terrace thioarsenates were present at 55% total arsenic. We also quantified sulfide, thiosulfate, sulfate and elemental sulfur, finding sulfide and sulfate as major species in the pool and outflow terrace, respectively. Elemental sulfur reached a maximum at the terrace. Phylogenetic analysis of 16S rRNA genes from metagenomic sequencing revealed the dominance of Sulfurihydrogenibium at all sites and an increased archaeal population at the rim and outflow channel. Several phylotypes were found closely related to known sulfur- and sulfide-oxidizers, as well as sulfur- and sulfate-reducers. Bioinformatic analysis revealed genes underpinning sulfur redox transformations, consistent with sulfur speciation data, and illustrating a microbial role in sulfur-dependent transformation of arsenite to thioarsenate. Metagenomic analysis also revealed genes encoding for arsenate reductase at all sites, reflecting the ubiquity of thioarsenate and a need for microbial arsenate resistance despite anoxic conditions. Absence of the arsenite oxidase gene, aio, at all sites suggests prioritization of arsenite detoxification over coupling to energy conservation. Finally, detection of methyl arsenic in the outflow channel, in conjunction with

  10. Greenland Tidal Pools as Hot Spots for Ecosystem Metabolism and Calcification

    KAUST Repository

    Duarte, Carlos M.

    2018-01-18

    The hypothesis that Arctic tidal pools provide environmental conditions suitable for calcifiers during summer, thereby potentially providing refugia for calcifiers in an acidifying Arctic Ocean, was tested on the basis of measurements conducted during two midsummers (2014 and 2016) in tidal pools colonised by a community composed of macroalgae and calcifiers in Disko Bay, Greenland (69° N). The tidal pools exhibited steep diurnal variations in temperature from a minimum of about 6 °C during the night to a maximum of almost 18 °C in the afternoon, while the temperature of the surrounding shore water was much lower, typically in the range 3 to 8 °C. O2 concentrations in the tidal pools were elevated relative to those in the adjacent open waters, by up to 11 mg O2 L−1, and exhibited heavy super-saturation (up to > 240%) during daytime emersion, reflecting intense and sustained photosynthetic rates of the tidal macroalgae. The intense photosynthetic activity of the seaweeds resulted in the drawdown of pCO2 concentrations in the pools during the day to levels down to average (±SE) values of 66 ± 18 ppm, and a minimum recorded value of 14.7 ppm, corresponding to pH levels as high as 8.69 ± 0.08, as compared to CO2 levels of 256 ± 4 and pH levels of 8.14 ± 0.01 in the water flooding the pools during high tide. The corresponding Ωarag reached 5.04 ± 0.49 in the pools as compared to 1.55 ± 0.02 in the coastal waters flooding the pools. Net calcification averaged 9.6 ± 5.6 μmol C kg−1 h−1 and was strongly and positively correlated with calculated net ecosystem production rates, which averaged 27.5 ± 8.6 μmol C kg−1 h−1. Arctic tidal pools promote intense metabolism, creating conditions suitable for calcification during the Arctic summer, and can, therefore, provide refugia from ocean acidification to vulnerable calcifiers as extended periods of continuous light during summer are conducive to suitable conditions

  11. Greenland Tidal Pools as Hot Spots for Ecosystem Metabolism and Calcification

    KAUST Repository

    Duarte, Carlos M.; Krause-Jensen, Dorte

    2018-01-01

    The hypothesis that Arctic tidal pools provide environmental conditions suitable for calcifiers during summer, thereby potentially providing refugia for calcifiers in an acidifying Arctic Ocean, was tested on the basis of measurements conducted during two midsummers (2014 and 2016) in tidal pools colonised by a community composed of macroalgae and calcifiers in Disko Bay, Greenland (69° N). The tidal pools exhibited steep diurnal variations in temperature from a minimum of about 6 °C during the night to a maximum of almost 18 °C in the afternoon, while the temperature of the surrounding shore water was much lower, typically in the range 3 to 8 °C. O2 concentrations in the tidal pools were elevated relative to those in the adjacent open waters, by up to 11 mg O2 L−1, and exhibited heavy super-saturation (up to > 240%) during daytime emersion, reflecting intense and sustained photosynthetic rates of the tidal macroalgae. The intense photosynthetic activity of the seaweeds resulted in the drawdown of pCO2 concentrations in the pools during the day to levels down to average (±SE) values of 66 ± 18 ppm, and a minimum recorded value of 14.7 ppm, corresponding to pH levels as high as 8.69 ± 0.08, as compared to CO2 levels of 256 ± 4 and pH levels of 8.14 ± 0.01 in the water flooding the pools during high tide. The corresponding Ωarag reached 5.04 ± 0.49 in the pools as compared to 1.55 ± 0.02 in the coastal waters flooding the pools. Net calcification averaged 9.6 ± 5.6 μmol C kg−1 h−1 and was strongly and positively correlated with calculated net ecosystem production rates, which averaged 27.5 ± 8.6 μmol C kg−1 h−1. Arctic tidal pools promote intense metabolism, creating conditions suitable for calcification during the Arctic summer, and can, therefore, provide refugia from ocean acidification to vulnerable calcifiers as extended periods of continuous light during summer are conducive to suitable conditions

  12. Aerosol release from a hot sodium pool and behaviour in inert gas atmosphere

    International Nuclear Information System (INIS)

    Sauter, H.; Schuetz, W.

    1986-01-01

    In the KfK-NALA program, experiments were carried out on the subject of aerosol release from a contaminated sodium pool into inert gas atmosphere under various conditions. Besides the determination of retention factors for fuel and fission products, the sodium aerosol system was investigated and characterized, concerning aerosol generation (evaporation rate), particle size, mass concentration, and deposition behaviour. Pool temperatures were varied between 700 and 1000 K at different geometrical and convective conditions. Technical scale experiments with a 531-cm 2 pool surface area were performed at natural convection in a 2.2-m 3 heated vessel, as well as additional small scale experiments at forced convection and 38.5-cm 2 pool surface area. A best-fit formula is given for the specific evaporation rate into a 400 K argon atmosphere. Approximately, the very convenient relation (dm/dt) (kg/m 2 /h) = 0.1 p (mm Hg) was found. The sodium aerosol diameter lay between 0.6 μm, less than 1 sec after production, and 2.5 μm at maximum concentration. The deposition behaviour was characterized by very small quantities ( 80%) on the bottom cover of the vessel. In the model theoretic studies with the PARDISEKO code, calculations were performed of the mass concentration, particle diameter and deposition behaviour. Agreement with the experimental values could not be achieved until a modulus was introduced to allow for turbulent deposition. (author)

  13. Probabilistic risk analysis for Test Area North Hot Shop Storage Pool Facility

    International Nuclear Information System (INIS)

    Meale, B.M.; Satterwhite, D.G.

    1990-01-01

    A storage pool facility used for storing spent fuel and radioactive debris from the Three Mile Island (TMI) accident was evaluated to determine the risk associated with its normal operations. Several hazards were identified and examined to determine if any any credible accident scenarios existed. Expected annual occurrence frequencies were calculated for hazards for which accident scenarios were identified through use of fault trees modeling techniques. Fault tree models were developed for two hazards: (1) increased radiation field and (2) spread of contamination. The models incorporated facets of the operations within the facility as well as the facility itself. 6 refs

  14. EBR-II in-vessel natural circulation experiments on hot and cold pool stratification

    International Nuclear Information System (INIS)

    Ragland, W.A.; Feldman, E.E.

    1990-01-01

    The Experimental Breeder Reactor II is located in a cylindrical pool of liquid sodium which is part of the cold-leg of the primary flow circuit. A vertical string of 32 thermocouples spans the 8 m tank height, at each of two diametrically opposed locations in the primary tank. Local temperatures were measured with these 64 thermocouples during dynamic tests. The instantaneous spacial temperature distribution obtained from a string of thermocouples can be viewed on a personal computer. The animation which results from displaying successive spacial distributions provide a very effective way to quickly obtain physical insights. The design of the two strings of thermocouples, the software used to create the animation, measured data from three different types of tests--two unprotected reactor transients, and one with the reactor at decay power levels and the reactor cover lifted, are discussed. 5 refs., 3 figs

  15. Pius, self-protective thermohydraulics transient without safety system intervention

    International Nuclear Information System (INIS)

    Fredell, J.; Bredolt, V.

    1989-01-01

    In this paper, the self-protective thermohydraulic feedback of the PIUS reactor system is illustrated by an in-depth discussion of one typical transient. The selected transient is an undetected total loss of feedwater in the complete absence of conventional safety system intervention. The reactor shuts itself down to residual power in two steps. First, the power decreases due to the strongly negative moderator temperature reactivity coefficient, and then a complete shutdown occurs by ingress of cold, highly borated water from the reactor pool. The transient is terminated without any harm to the fuel or paint systems

  16. Investigation on the hot melting temperature field simulation of HDPE water supply pipeline in gymnasium pool

    Science.gov (United States)

    Cai, Zhiqiang; Dai, Hongbin; Fu, Xibin

    2018-06-01

    In view of the special needs of the water supply and drainage system of swimming pool in gymnasium, the correlation of high density polyethylene (HDPE) pipe and the temperature field distribution during welding was investigated. It showed that the temperature field distribution has significant influence on the quality of welding. Moreover, the mechanical properties of the welded joint were analyzed by the bending test of the weld joint, and the micro-structure of the welded joint was evaluated by scanning electron microscope (SEM). The one-dimensional unsteady heat transfer model of polyethylene pipe welding joints was established by MARC. The temperature field distribution during welding process was simulated, and the temperature field changes during welding were also detected and compared by the thermo-couple temperature automatic acquisition system. Results indicated that the temperature of the end surface of the pipe does not reach the maximum value, when it is at the end of welding heating. Instead, it reaches the maximum value at 300 sand latent heat occurs during the welding process. It concludes that the weld quality is the highest when the welding pressure is 0.2 MPa, and the heating temperature of HDPE heat fusion welding is in the range of 210 °C-230 °C.

  17. Experiment on thermohydraulics of simulated control rod

    International Nuclear Information System (INIS)

    Ogawa, Masuro; Ouchi, Mitsuo; Akino, Norio; Fujimura, Kaoru; Shiina, Yasuaki; Kawamura, Hiroshi

    1984-10-01

    A thermohydraulic study of a control rod channel is required for the core design of the Very High Temperature Gas Cooled Reactor (VHTR). A non-heating experiment with air flow was performed prior to heating experiment with helium flow. Experimental results on stability of flow, flow rate distribution and pressure drop of the control rod channel are reported. In a test section of the experimental apparatus, five simulated control subrods were suspended vertically in a circular duct. Their dimension was in coincide with those of the Detailed Disign (I) of the VHTR. Air of atomospheric pressure was used as a coolant gas, which flowed in inner and outer paths of the subrods. Total flow rate ranged from 0.0011 to 0.0062 kg/s. Flow rate distribution and pressure drop were obtained for various flow rates. Velocity fluctuation in the channel was also observed using a hot wire anemometer. From these experiments, it was found that the flow rate distribution was nearly the same as a disigned value and that turbulent and laminar flows were simultaneously realized in outer and inner paths respectively. These observations supported a feasibility of the present design. (author)

  18. Hot flush frequency and severity at baseline as predictors of time to transient and stable treatment success: pooled analysis of two CE/BZA studies.

    Science.gov (United States)

    Pinkerton, JoAnn V; Bushmakin, Andrew G; Bobula, Joel; Lavenberg, Joanne; Komm, Barry S; Abraham, Lucy

    2017-12-01

    To evaluate the impact of baseline hot flush frequency and severity on time to symptom improvement during treatment with conjugated estrogens/bazedoxifene (CE/BZA). Data were pooled through week 12 from two randomized placebo-controlled trials (SMART-1 and SMART-2) of nonhysterectomized postmenopausal women with hot flushes treated with CE 0.45 mg/BZA 20 mg or CE 0.625 mg/BZA 20 mg. Time to transient and stable improvement (≥ 50% reduction in hot flush frequency/severity) was estimated using nonparametric models. Transient improvement in hot flush frequency occurred earlier in women treated with CE 0.45 mg/BZA 20 mg with less frequent versus more frequent baseline hot flushes per day: median time to transient improvement was 2, 7, and 11 days for women with hot flushes per day at baseline, respectively (P = 0.0009). Transient improvement in severity occurred earlier for women with less severe versus more severe baseline hot flushes: median time to transient improvement was 2, 6, and 16 days for women with mild, moderate, and severe hot flushes at baseline, respectively (P hot flushes take longer to achieve transient improvements with CE/BZA and should be encouraged to continue treatment, as it may take longer than a few weeks to achieve significant improvement.

  19. LMFR core thermohydraulics: Status and prospects

    International Nuclear Information System (INIS)

    2000-06-01

    One of the fundamental steps for a successful reactor core thermohydraulic design is the capability to predict, reliably and accurately, the temperature distribution in the core assemblies. A detailed knowledge of the assembly and fuel pin thermohydraulic behaviour in the steady state and transient conditions is an indispensable prerequisite to safe and stable operation of the reactor. Considerable experimental and theoretical studies on various aspects of LMFR core thermohydraulics are necessary to acquire such knowledge. During the last decade, there have been substantial advances in fast reactor core thermohydraulic design and operation in several countries with fast reactor programmes (notably in France, the Russian Federation, Japan, the United Kingdom, Germany and the United States of America). Chief among these has been the demonstration of reliable operation of reactor cores at a high burnup. During the last years, some additional countries such as China, India and the Republic of Korea have launched new fast reactor programmes. International exchange of information and experience on LMFR development including core thermohydraulic design is becoming of increasing importance to these countries. It is with this focus that the IAEA convened the Technical Committee on 'Methods and Codes for Calculations of Thermohydraulic Parameters for Fuel, Absorber Pins and Assemblies of LMFR's with Traditional and Burner Cores'. This meeting, attended by participants from seven countries, brought together a group of international experts to review and discuss the thermohydraulic advances and design approaches providing a reliable, safe and robust reactor core, as well as to exchange the experience accumulated in different countries of using the codes for thermohydraulic calculations and to discuss the issues requiring further research and development. A total of thirty technical papers presented covered theoretical and computational issues as well as experiments under

  20. Vertical, horizontal, and temporal changes in temperature in the Atlantis II and Discovery hot brine pools, Red Sea

    KAUST Repository

    Swift, Stephen A.

    2012-06-01

    In October 2008, we measured temperature and salinity in hot, hypersaline brine filling the Atlantis II and Discovery Deeps on the Red Sea spreading center west of Jeddah, Saudi Arabia. In agreement with previous observations in the Atlantis II Deep, we found a stack of four convective layers with vertically uniform temperature profiles separated by thin interfaces with high vertical temperature gradients. Temperature in the thick lower convective layer in the Atlantis II Deep continued to slowly increase at 0.1 °C/year since the last observations in 1997. Previously published data show that the temperature of all four convective layers increased since the 1960s at the same rate, from which we infer that diffusive vertical heat flux between convective layers is rapid on time scales of 3-5 years and, thus, heat is lost from the brine pools to overlying Red Sea Deep Water. Heat budgets suggest that the heat flux from hydrothermal venting has decreased from 0.54. GW to 0.18. GW since 1966. A tow-yo survey found that temperature in the upper convective layers changes about 0.2 °C over 5-6. km but the temperature in the lower brine layer remains constant. Temperature in the lower convective layer in the Discovery Deep remains unchanged at 48 °C. To explain these results, we hypothesize that heat flux from a hydrothermal vent in the floor of the Discovery Deep has been stable for 40 years, whereas temperature of the brine in the Atlantis II Deep is adjusting to the change in hydrothermal heat flux from the vent in the Southwest Basin. We found no changes in the upper transition layer at 1900-1990. m depth that appeared between 1976 and 1992 and suggest that this layer originated from the seafloor elsewhere in the rift. © 2012 Elsevier Ltd.

  1. Geothermal heat for Erding. 2. Energy and wellness, geothermal heating station and hot-water indoor swimming pool; Geowaerme fuer Erding 2. Energie und Wellness, Geothermieheizwerk und Thermalbad

    Energy Technology Data Exchange (ETDEWEB)

    Tenzer, H. (comp.); Bussmann, W.

    1999-07-01

    This 17:20 minute VHS-PAL video film describes the project 'Geothermal heat for Erding 2', i.e. the construction of the geothermal heating station and a modern hot-water indoor swimming pool. [German] Der vorliegende VHS-PAL-Videofilm beschreibt innerhalb von 17:20 Min. Lauflaenge das Projekt 'Geowaerme fuer Erding 2'. Gezeigt werden die Entstehungsphasen dieses Projektes bestehend aus einem Geothermieheizwerk und einem modernen Thermalbad. (AKF)

  2. Chemistry of Hot Spring Pool Waters in Calamba and Los Banos and Potential Effect on the Water Quality of Laguna De Bay

    Science.gov (United States)

    Balangue, M. I. R. D.; Pena, M. A. Z.; Siringan, F. P.; Jago-on, K. A. B.; Lloren, R. B.; Taniguchi, M.

    2014-12-01

    Since the Spanish Period (1600s), natural hot spring waters have been harnessed for balneological purposes in the municipalities of Calamba and Los Banos, Laguna, south of Metro Manila. There are at more than a hundred hot spring resorts in Brgy. Pansol, Calamba and Tadlac, Los Banos. These two areas are found at the northern flanks of Mt. Makiling facing Laguna de Bay. This study aims to provide some insights on the physical and chemical characteristics of hot spring resorts and the possible impact on the lake water quality resulting from the disposal of used water. Initial ocular survey of the resorts showed that temperature of the pool water ranges from ambient (>300C) to as high as 500C with an average pool size of 80m3. Water samples were collected from a natural hot spring and pumped well in Los Banos and another pumped well in Pansol to determine the chemistry. The field pH ranges from 6.65 to 6.87 (Pansol springs). Cation analysis revealed that the thermal waters belonged to the Na-K-Cl-HCO3 type with some trace amount of heavy metals. Methods for waste water disposal are either by direct discharge down the drain of the pool or by discharge in the public road canal. Both methods will dump the waste water directly into Laguna de Bay. Taking in consideration the large volume of waste water used especially during the peak season, the effect on the lake water quality would be significant. It is therefore imperative for the environmental authorities in Laguna to regulate and monitor the chemistry of discharges from the pool to protect both the lake water as well as groundwater quality.

  3. Specialists' meeting on correlation between material properties and thermohydraulics conditions in LMFRs

    International Nuclear Information System (INIS)

    1994-01-01

    In a liquid metal fast reactor (LMFR), temperature fluctuations in the fluid close to a structure occur in many areas: core outlet zone, lower part of hot pool, free surface of pool, IHX outlet, secondary circuit, water steam interface in steam generators. In some conditions, these temperature fluctuations can lead to mechanical damage to structures. Consequently, knowledge of temperature fluctuations and induced thermomechanical damage to structures is essential to support design and maintenance during the plant life-time. In response to a recommendation from the IWGFR, the IAEA convened a Specialist Meeting on 'Correlation between material properties and thermohydraulics conditions in LMFRs' in November 1994. The purpose of the meeting was to exchange information on the state of the art on thermalhydraulic aspects of temperature fluctuations (mixing jet phenomena, temperature gradient fluctuations, transfer of fluctuations from the fluid to the wall), and associated thermomechanical studies (thermal striping, thermal ratchetting, high strain fatigue) as well as design criteria to avoid damage. The main areas discussed by the delegates were: thermalhydraulics and thermomechanics. The objective of thermalhydraulic activities is the characterization of the temperature fluctuations on the wall. Three main items can be identified, for which both the experimental and calculational approaches were considered: identification of the areas where the fluctuations may occur; characterization of the fluctuations in the fluid; and transfer of the fluid fluctuations to the walls. For thermomechanical studies, which cover the effect of the fluctuations in the structures, the following subjects are of great importance: determination of the damage modes induced by thermal loadings in structures (thermal striping, ratchetting, high strain fatigue), and study of all damage modes so as to take them into account in the design criteria and to provide rules for avoiding failure of the

  4. Specialists' meeting on correlation between material properties and thermohydraulics conditions in LMFRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-01

    In a liquid metal fast reactor (LMFR), temperature fluctuations in the fluid close to a structure occur in many areas: core outlet zone, lower part of hot pool, free surface of pool, IHX outlet, secondary circuit, water steam interface in steam generators. In some conditions, these temperature fluctuations can lead to mechanical damage to structures. Consequently, knowledge of temperature fluctuations and induced thermomechanical damage to structures is essential to support design and maintenance during the plant life-time. In response to a recommendation from the IWGFR, the IAEA convened a Specialist Meeting on 'Correlation between material properties and thermohydraulics conditions in LMFRs' in November 1994. The purpose of the meeting was to exchange information on the state of the art on thermalhydraulic aspects of temperature fluctuations (mixing jet phenomena, temperature gradient fluctuations, transfer of fluctuations from the fluid to the wall), and associated thermomechanical studies (thermal striping, thermal ratchetting, high strain fatigue) as well as design criteria to avoid damage. The main areas discussed by the delegates were: thermalhydraulics and thermomechanics. The objective of thermalhydraulic activities is the characterization of the temperature fluctuations on the wall. Three main items can be identified, for which both the experimental and calculational approaches were considered: identification of the areas where the fluctuations may occur; characterization of the fluctuations in the fluid; and transfer of the fluid fluctuations to the walls. For thermomechanical studies, which cover the effect of the fluctuations in the structures, the following subjects are of great importance: determination of the damage modes induced by thermal loadings in structures (thermal striping, ratchetting, high strain fatigue), and study of all damage modes so as to take them into account in the design criteria and to provide rules for avoiding failure of the

  5. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    Frid, W.

    1991-08-01

    This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

  6. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor; Modelo simplificado para simulacao do comportamento termohidraulico do canal quente de reator nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Belem, J A.T.

    1993-09-01

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs.

  7. Neutronics and thermohydraulics of the reactor C.E.N.E. Part II

    International Nuclear Information System (INIS)

    Caro, R.

    1976-01-01

    In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

  8. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

    Energy Technology Data Exchange (ETDEWEB)

    Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

    1985-07-01

    This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

  9. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Veloso, M.A.

    1980-01-01

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

  10. Vertical, horizontal, and temporal changes in temperature in the Atlantis II and Discovery hot brine pools, Red Sea

    KAUST Repository

    Swift, Stephen A.; Bower, Amy S.; Schmitt, Raymond W.

    2012-01-01

    convective layers is rapid on time scales of 3-5 years and, thus, heat is lost from the brine pools to overlying Red Sea Deep Water. Heat budgets suggest that the heat flux from hydrothermal venting has decreased from 0.54. GW to 0.18. GW since 1966. A tow

  11. An Origin of Life in Cycling Hot Spring Pools: Emerging Evidence from Chemistry, Geology and Computational Studies

    Science.gov (United States)

    Deamer, D. W.; Damer, B. F.; Van Kranendonk, M. J.; Djokic, T.

    2017-07-01

    New evidence for an origin of life in a hot spring setting on land is supported by three studies: chemical (polymerization in wet-dry cycles), geological (stromatolites in a 3.48 Ga geothermal field) and computational (verifying the kinetic trap).

  12. Does addition of `mud-pack and hot pool treatment' to patient education make a difference in fibromyalgia patients? A randomized controlled single blind study

    Science.gov (United States)

    Bağdatlı, Ali Osman; Donmez, Arif; Eröksüz, Rıza; Bahadır, Güler; Turan, Mustafa; Erdoğan, Nergis

    2015-12-01

    The aim of this randomized controlled single-blind study is to explore whether addition of mud-pack and hot pool treatments to patient education make a significant difference in short and mild term outcomes of the patients with fibromyalgia. Seventy women with fibromyalgia syndrome were randomly assigned to either balneotherapy with mud-pack and hot pool treatments (35) or control (35) groups. After randomization, five patients from balneotherapy group and five patients from control group were dropped out from the study with different excuses. All patients had 6-h patient education programme about fibromyalgia syndrome and were given a home exercise programme. The patients in balneotherapy group had heated pool treatment at 38 °C for 20 min a day, and mud-pack treatment afterwards on back region at 45 °C. Balneotherapy was applied on weekdays for 2 weeks. All patients continued to take their medical treatment. An investigator who was blinded to the intervention assessed all the patients before and after the treatment, at the first and the third months of follow-up. Outcome measures were FIQ, BDI and both patient's and physician's global assessments. Balneotherapy group was significantly better than control group at after the treatment and at the end of the first month follow-up assessments in terms of patient's and physician's global assessment, total FIQ score, and pain intensity, fatigue, non-refreshed awaking, stiffness, anxiety and depression subscales of FIQ. No significant difference was found between the groups in terms of BDI scores. It is concluded that patient education combined with 2 weeks balneotherapy application has more beneficial effects in patients with fibromyalgia syndrome as compared to patient education alone.

  13. Does addition of 'mud-pack and hot pool treatment' to patient education make a difference in fibromyalgia patients? A randomized controlled single blind study.

    Science.gov (United States)

    Bağdatlı, Ali Osman; Donmez, Arif; Eröksüz, Rıza; Bahadır, Güler; Turan, Mustafa; Erdoğan, Nergis

    2015-12-01

    The aim of this randomized controlled single-blind study is to explore whether addition of mud-pack and hot pool treatments to patient education make a significant difference in short and mild term outcomes of the patients with fibromyalgia. Seventy women with fibromyalgia syndrome were randomly assigned to either balneotherapy with mud-pack and hot pool treatments (35) or control (35) groups. After randomization, five patients from balneotherapy group and five patients from control group were dropped out from the study with different excuses. All patients had 6-h patient education programme about fibromyalgia syndrome and were given a home exercise programme. The patients in balneotherapy group had heated pool treatment at 38 °C for 20 min a day, and mud-pack treatment afterwards on back region at 45 °C. Balneotherapy was applied on weekdays for 2 weeks. All patients continued to take their medical treatment. An investigator who was blinded to the intervention assessed all the patients before and after the treatment, at the first and the third months of follow-up. Outcome measures were FIQ, BDI and both patient's and physician's global assessments. Balneotherapy group was significantly better than control group at after the treatment and at the end of the first month follow-up assessments in terms of patient's and physician's global assessment, total FIQ score, and pain intensity, fatigue, non-refreshed awaking, stiffness, anxiety and depression subscales of FIQ. No significant difference was found between the groups in terms of BDI scores. It is concluded that patient education combined with 2 weeks balneotherapy application has more beneficial effects in patients with fibromyalgia syndrome as compared to patient education alone.

  14. Library thermohydraulic components for training simulators

    International Nuclear Information System (INIS)

    Castelao Caruana, M. J.; Di Benedetto, A.; Pierini, J.P.

    2013-01-01

    The thermohydraulic components Library was modeled in MatLab/Simulink®. This library owns Pipe type components (pump, control valve and / or heaters), storage tanks (Open, Closed and Equilibrium Water Vapor-Air) and Heat Exchangers (Co-Current, Counter-Current and U-tubes). Each component can be attached to other components through the component library Header, in order to create a more complex thermal-hydraulic system which in turn can interact with other thermal-hydraulic systems. (author)

  15. Research activity on thermohydraulic problems of PWR reactors

    International Nuclear Information System (INIS)

    Szabados, L.

    1976-06-01

    The general review of the experimental and theoretical research works on thermohydraulic investigation of pressurized water type reactors being done in the Central Research Institute for Physics is given. The main results of the theoretical and theoretical-numerical research are summarized. The most important result of the past years is the construction of the High Pressure Water Cooled Loop (NVH) thermohydraulic loop. Another significant achievement was the development of the reactor thermohydraulic program system. (Sz.N.Z.)

  16. Basic researches on thermo-hydraulic non-equilibrium phenomena related to nuclear reactor safety

    International Nuclear Information System (INIS)

    Sakurai, Akira; Kataoka, Isao; Aritomi, Masanori.

    1989-01-01

    A review was made of recent developments of fundamental researches on thermo-hydraulic non-equilibrium phenomena related to light water reactor safety, in relation to problems to be solved for the improvement of safety analysis codes. As for the problems related to flow con ditions, fundamental researches on basic conservation equations and constitutive equations for transient two-phase flow were reviewed. Regarding to the problems related to thermal non-equilibrium phenomena, fundamental researches on film boiling in pool and forced convection, transient boiling heat transfer and flow behavior caused by pressure transients were reviewed. (author)

  17. Thermohydraulic tests in nuclear fuel model

    International Nuclear Information System (INIS)

    Ladeira, L.C.D.; Navarro, M.A.

    1984-01-01

    The main experimental works performed in the Thermohydraulics Laboratory of the NUCLEBRAS Nuclear Technology Development Center, in the field of thermofluodynamics are briefly described. These works include the performing of steady-state flow tests in single tube test sections, and the design and construction of a rod bundle test section, which will be also used for those kind of testes. Mention is made of the works to be performed in the near future, related to steady-state and transient flow tests. (Author) [pt

  18. Thermohydraulic model of WWER-1000 core

    International Nuclear Information System (INIS)

    Maroti, L.; Szabados, L.

    1987-11-01

    Safe and economic operation of the WWER-1000 type reactor requires more accurate calculation of the thermohydraulic processes than the one satisfactory for the 440 type cores. The high degree of accuracy is needed both for reactor physics calculations and for the determination of the operational safety limits of the core. The paper illustrates the most important differences between the 1000 and 440 type reactors and presents the main fields of the development work necessary to reach the required accuracy. A prediction for the capability of the computer programs after the proposed development is also given and some suggestions for the further improvement is outlined. (author) 7 refs

  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. Evaluation of filters in RSPCS (Reactor Service Pool Cooling System) and HWL (Hot Water Layer) in OPAL research reactor at ANSTO (Australian Nuclear Science and Technology Organization) using Gamma Spectrometry System and Liquid Scintillation Counter

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jim In; Foy, Robin; Jung, Seong Moon; Park, Hyeon Suk; Ye, Sung Joon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Australian Nuclear Science and Technology Organization(ANSTO) has a research reactor, OPAL (Open Pool Australian Lightwater reactor) which is a state-of-art 20 MW reactor for various purposes. In OPAL reactor, there are many kinds of radionuclides produced from various reactions in pool water and those should be identified and quantified for the safe use of OPAL. To do that, it is essential to check the efficiency of filters which are able to remove the radioactive substance from the reactor pool water. There are two main water circuits in OPAL which are RSPCS (Reactor Service Pool Cooling System) and HWL (Hot Water Layer) water circuits. The reactor service pool is connected to the reactor pool via a transfer canal and provides a working area and storage space for the spent and other materials. Also, HWL is the upper part of the reactor pool water and it minimize radiation dose rates at the pool surface. We collected water samples from these circuits and measured the radioactivity by using Gamma Spectrometry System (GSS) and Liquid Scintillation Counter (LSC) to evaluate the filters. We could evaluate the efficiency of filters in RSPCS and HWL in OPAL research reactor. Through the measurements of radioactivity using GSS and LSC, we could conclude that there is likely to be no alpha emitter in water samples, and for beta and gamma activity, there are very big differences between inlet and outlet results, so every filter is working efficiently to remove the radioactive substance.

  1. A Vision Controlled Robot to Detect and Collect Fallen Hot Cobalt60 Capsules inside Wet Storage Pool of Cobalt60 Irradiators

    International Nuclear Information System (INIS)

    Solyman, A.E.M.

    2015-01-01

    In a typical irradiator that use radioactive cobalt-60 capsules source is one of the peaceful uses of atomic energy, it originated strategy in terms of its importance in the sterilization of medical products and food processing from bacteria and fungi before being exported. However, there are several well-known problems related to the fall of the cobalt-60 capsules inside the wet storage pool as a result of manufacturing defects, defects welds or a problem occurs in the vertical movement of the radioactive source rack. Therefore it is necessary to study this problem and solve it in a scientific way so as to keep the human as much as possible from radiation exposure, according to the principles of radiation protection and safety issued by the International Atomic Energy Agency. The present work considers the possibility to use a vision based control arm robot to collect fallen hot cobalt-60 capsules inside wet storage pool. A 5-DOF arm robot is designed and vision algorithms are established to pick the fallen capsule on the bottom surface of the storage pool, read the information printed on its edge (cap) and move it to a safe storage place. Two object detection approaches are studied; RGB-based filter and background subtraction technique. Vision algorithms and camera calibration are done using MATLAB/SIMULINK program. Robot arm forward and inverse kinematics are developed and programmed using an embedded micro controller system. Experiments show the validity of the proposed system and prove its success. The collecting process will be done without interference of operators, so radiation safety will be increased. The results showed camera calibration equations accuracy. And also the presence of vibrations in the hands of the movement of the robot and thus were seized motor rotation speed to 10 degrees per second to avoid these vibrations.This scientific application keeps the operators as much as possible from radiation exposure so it leads to increase radiation

  2. 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

  3. Thermohydraulic calculations in rectangular channels for RA-6 type reactors with transition regime

    International Nuclear Information System (INIS)

    Sillin, N; Vertullo, A.; Masson, V.; Hilal, R

    2009-01-01

    In August 2000 and within the framework of the RA-6 core conversion from high to low enrichment (20%), a preliminary analysis was performed to evaluate the maximum power that the reactor could operate with the new kernel without makeing substantial changes. This meant keeping intact, for example, the concrete shield of the pool and the nucleus inlet and outlet pipes embedded in the walls. Preliminary results indicated that for these boundary conditions a maximum power of about 3 MWt could be achieved. In August 2005 the project was resumed and new calculations performed taking as a starting point the ECBE plate fuel element(U3O8-Al). A core was developed with cooling channle widths of 2.6 mm for the control fuel elements and 2.7 mm for standard fuel elements. The thermo-hydraulic calculation puts in evidence that coolant flow into the core was in the transitional regime for the vast majority of configurations. While TERMIC code, used for thermo-hydraulic design, has been extensively tested and validated for use in research reactors under turbulent and laminar flows, this is not so for transition conditions. The transition regime is strongly dependent on conditions such as flow inlet characteristics, channel geometry, etc.. and therefore there are no reliable correlations for general use. For this reason we found it convenient to carry out experiments simulating the working conditions in order to adjust the code results with experimental data. In the present work we show the experimental results, the simulation of the experiences using the TERMIC code, and the adjustments made to the correlations used by the code so that it can be applied to the thermo-hydraulic design of the new core. [es

  4. Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

    International Nuclear Information System (INIS)

    Kobayashi, Michiyuki; Murata, Hiroyuki; Sawada, Kenichi; Inasaka, Fujio; Aya, Izuo; Shiozaki, Koki

    1999-01-01

    By inputting the experimental data, information and others on thermo-hydraulic characteristics of integrated ship propulsion reactor accumulated hitherto by the Ship Research Institute and some recent cooperation results into the nuclear ship engineering simulation system, it was conducted not only to contribute an improvement study on next ship reactor by executing general analysis and evaluation on motion characteristics under ship body motion conditions, safety at accidents, and others of the integrated ship reactor but also to investigate and prepare some measures to apply fundamental experiment results based on obtained here information to safety countermeasure of the nuclear ships. In 1997 fiscal year, on safety of the integrated ship propulsion reactor loading nuclear ship, by adding experimental data on unstable flow analysis and information on all around of the analysis to general data base fundamental program, development to intellectual data base program was intended; on effect of pulsation flow on thermo-hydraulic characteristics of ship propulsion reactor; after pulsation flow visualization experiment, experimental equipment was reconstructed into heat transfer type to conduct numerical analysis of pulsation flow by confirming validity of numerical analysis code under comparison with the visualization experiment results; and on thermo-hydraulic behavior in storage container at accident of active safety type ship propulsion reactor; a flashing vibration test using new apparatus finished on its higher pressurization at last fiscal year to examine effects of each parameter such as radius and length of exhausting nozzle and pool water temperature. (G.K.)

  5. Neutronics and thermohydraulics of the reactor C.E.N.E. Part II; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte II

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R

    1976-07-01

    In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

  6. Proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic

    International Nuclear Information System (INIS)

    1986-01-01

    The proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic - 6. ENFIR - allow to evaluate the present status of development in reactor physics and thermohydraulic fields. The mathematical models and methods for calculating neutronic of nuclear reactors, safety reactor analysis, measuring methods of neutronic parameters, computerized simulation of accidents, transients and thermohydraulic analysis are presented. (M.C.K.) [pt

  7. Assessment of TRAC-PD2 reflood core thermo-hydraulic model by CCTF Test C1-16

    International Nuclear Information System (INIS)

    Sugimoto, Jun

    1982-11-01

    The TRAC-PD2 reflood core thermo-hydraulic model was assessed by CCTF Test C1-16. The measured data were utilized as core boundary conditions in the TRAC calculations. The results indicate that the core inlet liquid temperature and the core heater rod temperatures are in reasonable agreement with data, but the pressure distribution in the core and water pool formation in the upper plenum are not in good agreement. The parametric effects of the droplet critical Weber number, the material properties of the heater rod, the noding of the upper plenum, and the minimum stable film boiling temperature are also discussed. (author)

  8. Thermohydraulic tests of 3x3-rod bundle maquette

    International Nuclear Information System (INIS)

    Ladeira, L.C.D.

    1986-10-01

    The results of a 3x3-rod bundle thermohydraulic research program, performed in the Thermohydraulics Laboratory of NUCLEBRAS' Nuclear Technology Development Center, are briefly described. This program included measurements of pressure drops in one and two-phase flows, heat transfer coefficients, mixing between interconnected subchannels in one-phase flow conditions and critical heat fluxes. The measurements covered the following parameter ranges: heat fluxes from zero to the critical values, pressure ranging from 1 to 15 ata, inlet temperature from 25 to 150 sup(0)C and flow rate from 20 to 300l/min. (author)

  9. Observation and control system of the thermohydraulic assays laboratory

    International Nuclear Information System (INIS)

    Santome, D.; Hualde, R.

    1990-01-01

    The Thermohydraulic Assays Laboratory (L.E.T.) is an installation whose purpose will be the components testing and the CAREM-25 reactor thermohydraulic processes operation dynamics. This plant is located at Pilcaniyeu, province of Rio Negro. Part of the tests which will be carried out consist in the use of different control strategies. The control of the systems by digital processors (control by software) has been decided to proceed with a maximum flexibility and capacity to make changes in the algorithms. This work describes the design and implementation of a digital control system to command the three circuits of the installation. (Author) [es

  10. Several new thermo-hydraulic test facilities in NPIC

    International Nuclear Information System (INIS)

    Ye Shurong; Sun Yufa; Ji Fuyun; Zong Guifang; Guo Zhongchuan

    1997-01-01

    Several new thermo-hydraulic test facilities are under construction in Nuclear Power Institute of Chinese (NPIC) at Chengdu. These facilities include: 1. Nuclear Power Component Comprehensive Test Facility. 2. Reactor Hydraulic Modeling Test Facility. 3. Control Rod Drive Line Hydraulic Test Facility. 4. Large Scale Thermo-Hydraulic Test Facility. The construction of these facilities will make huge progress in the research and development capability of nuclear power technology in CHINA. The author will present a brief description of the design parameters flowchart and test program of these facilities

  11. Utilization of Relap 5 computer code for analyzing thermohydraulic projects

    International Nuclear Information System (INIS)

    Silva Filho, E.

    1987-01-01

    This work deals with the design of a scaled test facility of a typical pressurized water reactor plant of the 1300 MW (electric) class. A station blackout has been choosen to investigate the thermohydraulic behaviour of the the test facility in comparison to the reactor plant. The computer code RELAPS/MOD1 has been utilized to simulate the blackout and to compare the test facility behaviour with the reactor plant one. The results demonstrate similar thermohydraulic behaviours of the two systems. (author) [pt

  12. Thermo-hydraulic Quench Propagation at the LHC Superconducting Magnet String

    CERN Document Server

    Rodríguez-Mateos, F; Serio, L

    1998-01-01

    The superconducting magnets of the LHC are protected by heaters and cold by-pass diodes. If a magnet quenches, the heaters on this magnet are fired and the magnet chain is de-excited in about two minu tes by opening dump switches in parallel to a resistor. During the time required for the discharge, adjacent magnets might quench due to thermo-hydraulic propagation in the helium bath and/or heat con duction via the bus bar. The number of quenching magnets depends on the mechanisms for the propagation. In this paper we report on quench propagation experiments from a dipole magnet to an adjacent ma gnet. The mechanism for the propagation is hot helium gas expelled from the first quenching magnet. The propagation changes with the pressure opening settings of the quench relief valves.

  13. Statistical Hot Channel Factors and Safety Limit CHFR/OFIR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byeonghee; Park, Suki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The fuel integrity of research reactors are usually judged by comparing the critical heat flux ratio (CHFR) and the maximum fuel temperature (MFT) with the safety limits. Onset of flow instability ratio (OFIR) can also be used for the examination with CHFR. Hot channel factors (HCFs) are incorporated when calculating the CHFR/OFIR and MFT, to consider the uncertainties of fuel properties and thermo-hydraulic variables affecting them. The HCFs and safety limit CHFR is sometimes estimated to include too much conservatism, deteriorating the design flexibilities and operating margins. In this paper, a statistical estimation of HCFs and the safety limit CHFR/OFIR is presented by a random sampling of uncertainty parameters. A 15MW pool type research reactor is selected as the sample reactor for the estimation. The HCFs and the safety limit CHFR/OFIR of a 15MW pool type research reactor are evaluated statistically. The parameters affecting the HCF and the safety limit CHFR/OFIR are listed and their uncertainties are estimated. The relevant parameter uncertainties are sampled randomly and the HCFs and the safety limits are evaluated from them. The HCFs and the safety limit CHFR/OFIR with 95% probability are smaller than those estimated deterministically because the statistical evaluation convolute the correlation uncertainties and the other uncertainties in probabilistic way, whereas the deterministic evaluation simply multiply them.

  14. The problems of thermohydraulics of prospective fast reactor concepts

    International Nuclear Information System (INIS)

    Sedov, A.A.

    2000-01-01

    In this report the main requirements to fast reactors in system of future multicomponent Nuclear Power with closed U-Pu fuel cycle are regarded. The peculiarities of different liquid-metal (sodium and lead-alloyed) coolants as well as the thermohydraulics problems of prospective fast reactors (FR) concepts are discussed. (author)

  15. Systems for neutronic, thermohydraulic and shielding calculation in personal computers

    International Nuclear Information System (INIS)

    Villarino, E.A.; Abbate, P.; Lovotti, O.; Santini, M.

    1990-01-01

    The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author) [es

  16. Thermohydraulic analysis for power increase of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Umbehaun, Pedro E.; Bastos, Jose L.F.

    1996-01-01

    In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30

  17. Cash pooling

    OpenAIRE

    Lozovaya, Karina

    2009-01-01

    This work makes a mention of cash management. At next chapter describes two most known theoretical models of cash management -- Baumol Model and Miller-Orr Model. Principal part of work is about cash pooling, types of cash pooling, cash pooling at Czech Republic and influence of cash pooling over accounting and taxes.

  18. Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

    International Nuclear Information System (INIS)

    2014-02-01

    The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

  19. Dynamic thermo-hydraulic model of district cooling networks

    International Nuclear Information System (INIS)

    Oppelt, Thomas; Urbaneck, Thorsten; Gross, Ulrich; Platzer, Bernd

    2016-01-01

    Highlights: • A dynamic thermo-hydraulic model for district cooling networks is presented. • The thermal modelling is based on water segment tracking (Lagrangian approach). • Thus, numerical errors and balance inaccuracies are avoided. • Verification and validation studies proved the reliability of the model. - Abstract: In the present paper, the dynamic thermo-hydraulic model ISENA is presented which can be applied for answering different questions occurring in design and operation of district cooling networks—e.g. related to economic and energy efficiency. The network model consists of a quasistatic hydraulic model and a transient thermal model based on tracking water segments through the whole network (Lagrangian method). Applying this approach, numerical errors and balance inaccuracies can be avoided which leads to a higher quality of results compared to other network models. Verification and validation calculations are presented in order to show that ISENA provides reliable results and is suitable for practical application.

  20. Exciting Pools

    Science.gov (United States)

    Wright, Bradford L.

    1975-01-01

    Advocates the creation of swimming pool oscillations as part of a general investigation of mechanical oscillations. Presents the equations, procedure for deriving the slosh modes, and methods of period estimation for exciting swimming pool oscillations. (GS)

  1. Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model

    International Nuclear Information System (INIS)

    Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko

    2002-01-01

    Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and

  2. Thermo-Hydraulic Modelling of Buffer and Backfill

    International Nuclear Information System (INIS)

    Pintado, X.; Rautioaho, E.

    2013-09-01

    The temporal evolution of saturation, liquid pressure and temperature in the components of the engineered barrier system was studied using numerical methods. A set of laboratory tests was conducted to calibrate the parameters employed in the models. The modelling consisted of thermal, hydraulic and thermo-hydraulic analysis in which the significant thermo-hydraulic processes, parameters and features were identified. CODE B RIGHT was used for the finite element modelling and supplementary calculations were conducted with analytical methods. The main objective in this report is to improve understanding of the thermo-hydraulic processes and material properties that affect buffer behaviour in the Olkiluoto repository and to determine the parametric requirements of models for the accurate prediction of this behaviour. The analyses consisted of evaluating the influence of initial canister temperature and gaps in the buffer, and the role played by fractures and the rock mass located between fractures in supplying water for buffer and backfill saturation. In the thermo-hydraulic analysis, the primary processes examined were the effects of buffer drying near the canister on temperature evolution and the manner in which heat flow affects the buffer saturation process. Uncertainties in parameters and variations in the boundary conditions, modelling geometry and thermo-hydraulic phenomena were assessed with a sensitivity analysis. The material parameters, constitutive models, and assumptions made were carefully selected for all the modelling cases. The reference parameters selected for the simulations were compared and evaluated against laboratory measurements. The modelling results highlight the importance of understanding groundwater flow through the rock mass and from fractures in the rock in order to achieve reliable predictions regarding buffer saturation, since saturation times could range from a few years to tens of thousands of years depending on the hydrogeological

  3. Fine numerical modelling of thermohydraulic phenomena in EDF PWR reactors

    International Nuclear Information System (INIS)

    Boulot, F.

    1993-01-01

    Over the last 20 years, EDF has developed a family of 2D and 3D industrial thermohydraulics software to solve problems encountered in existing PWR power plants and to design new reactors for the future. The equations used in the models are the averaged Navier-Stokes and energy equations. A brief description is given of the four main codes developed for single-phase and two-phase water-steam flows, some of which use finite differences or finite volumes methods, while others make use of finite elements methods. An example of application is given for each code. (author). 4 figs., 4 refs

  4. Simulation of thermohydraulic phenomena and model test for FBR

    International Nuclear Information System (INIS)

    Satoh, Kazuziro

    1994-01-01

    This paper summarizes the major thermohydraulic phenomena of FBRs and the conventional ways of their model tests, and introduces the recent findings regarding measurement technology and computational science. In the future commercial stage of FBRs, the design optimization will becomes important to improve economy and safety more and more. It is indispensable to use computational science to the plant design and safety evaluation. The most of the model tests will be replaced by the simulation analyses based on computational science. The measurement technology using ultrasonic and the numerical simulation with super parallel computing are considered to be the key technology to realize the design by analysis method. (author)

  5. Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

    International Nuclear Information System (INIS)

    Kamide, H.; Ieda, Y.; Toda, S.; Isozaki, T.; Sugawara, S.

    1993-01-01

    Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor core during natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

  6. Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, H; Ieda, Y; Toda, S; Isozaki, T; Sugawara, S [Reactor Engineering Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, Narita, O-arai, Ibaraki-ken (Japan)

    1993-02-01

    Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor coreduring natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

  7. Validation and verification of the MTRPC thermohydraulic package

    International Nuclear Information System (INIS)

    Doval, Alicia

    1998-01-01

    The MTR P C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)

  8. Thermohydraulic behaviour and heat transfer in the molten core

    International Nuclear Information System (INIS)

    Reineke, H.H.

    1977-01-01

    Increasing the application of nuclear reactors to produce electrical power extremely unprobable accidents should be investigated too. In the Federal Republic of Germany, a research program is performed for some years engaged in accidents at light water reactors in which the melting of the reactor core is presumed. A part of this program is to investigate the thermohydraulic and the heat transfer behavior in an accumulation of molten core material. The knowledge of these events is necessary to analyse the accident exactly. Further on the results of this work are of great importance to build a catcher for the molten core material. As a result of the decay heat the molten material is heated up and the density differences induce a free convection motion. In this work the thermohydraulic behavior and the distribution of the escaping heat fluxes for several accumulations of molten core material were determined. The numerical methods for solving the system of partial differential equation were used to develop computer codes, able to compute the average and local heat fluxes at the walls enclosing the molten core material and the inside increase of the temperature. The numerical computations were confirmed and verified by experimental investigations. In these investigations the molten core material was always assumed as a homogeneous fluid. In this case, the results could be reproduced by simple power laws

  9. Thermohydraulic stability coupled to the neutronic in a BWR

    International Nuclear Information System (INIS)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.

    2006-01-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the Laguna Verde

  10. Effects of thermohydraulics on clad ballooning, flow blockage and coolability in a LOCA

    International Nuclear Information System (INIS)

    Erbacher, F.J.; Neitzel, H.J.; Wiehr, K.

    1983-01-01

    Thermohydraulic boundary conditions have a dominating effect on clad ballooning, flow blockage and coolability: Increasing heat transfer to the fluid decreases the total circumferential strain; Countercurrent flow in a combined injection leads to a relatively small flow blockage; Burst claddings exhibit premature quenching. Differences in the test results obtained in several countries are mainly due to different thermohydraulic test conditions; all test data are consistent with the understanding elaborated within the REBEKA program. Core coolability in a LOCA can be maintained. (author)

  11. Pool scrubbing

    International Nuclear Information System (INIS)

    Lopez-Jimenez, J.; Herranz, J.; Escudero, M.J.; Espigares, M.M.; Peyres, V.; Polo, J.; Kortz, Ch.; Koch, M.K.; Brockmeier, U.; Unger, H.; Dutton, L.M.C.; Smedley, Ch.; Trow, W.; Jones, A.V.; Bonanni, E.; Calvo, M.; Alonso, A.

    1996-12-01

    The Source Term Project in the Third Frame Work Programme of the European Union Was conducted under and important joined effort on pool scrubbing research. CIEMAT was the Task Manager of the project and several other organizations participated in it: JRC-Ispra, NNC Limited, RUB-NES and UPM. The project was divided into several tasks. A peer review of the models in the pool scrubbing codes SPARC90 and BUSCA-AUG92 was made, considering the different aspects in the hydrodynamic phenomenology, particle retention and fission product vapor abortions. Several dominant risk accident sequences were analyzed with MAAP, SPARC90 and BUSCA-AUG92 codes, and the predictions were compared. A churn-turbulent model was developed for the hydrodynamic behaviour of the pool. Finally, an experimental programme in the PECA facility of CIEMAT was conducted in order to study the decontamination factor under jet injection regime, and the experimental observations were compared with the SPARC and BUSCA codes. (Author)

  12. A thermohydraulic analysis for LOCA accident of a CANDU 600 reactor core charged with SEU 43 fuel by means of FIREBIRD code

    International Nuclear Information System (INIS)

    Serbanel, M.; Catana, A.

    2001-01-01

    This report presents a comparative analysis of the behaviour of primary circuit during a LOCA 20% RIH accident for two types of reactor core, namely, normally charged, i.e., with clusters of 37 rods and charged with clusters of 43 rods, respectively. This type of accident was chosen since Canadian analyses showed that the associated transient regime stress the fuel elements. The void reactivity as a function of coolant average density was calibrated for a reference regime (LOCA 20% RIH) so that the results of the model be able to reproduce the average distribution in the reference transient regime. The computation makes use of CERBERUS and FIREBIRD codes externally coupled by files. The void reactivity of the hot pencil was obtained this way. An extremely conservative hypothesis was used, namely that the momentary power of the cluster hosting the pencil is the maximal power over the cluster for the corresponding half reactor core. To carry out this work the following steps were covered: 1. The scenario for the LOCA 20% RIH accident was worked out and the input data corresponding to the thermohydraulic and neutronic modules, for the complex model and the 37 rod clusters, were checked; 2. The input data corresponding to the thermohydraulic module for the complex model and the 43 rod cluster were checked; 3. The kinetic parameters corresponding to the 37 rod cluster were computed; 4. The kinetic parameters corresponding to the 43 rod cluster were computed and the file for the input data in the neutronic module was built; 5. A sub-routine for writing files with the thermohydraulic and neutronic quantities, in a format adequate to the other programs, was implemented; 6. The two transient regimes considered were implemented and the archives containing the quantities were built ;7. The results obtained were analyzed. The conclusion of this work is that in case of LOCA 20% RIH accident the 43 bar clusters have a better behaviour than the 37 bar clusters

  13. ALARM, Thermohydraulics of BWR with Jet Pumps During LOCA

    International Nuclear Information System (INIS)

    Araya, F.; Akimoto, M.

    1985-01-01

    1 - Nature of physical problem solved: ALARM-B2 which is an improved version of ALARM-B1 is a computer program to analyze thermo-hydraulic phenomena of BWR during a blowdown period under a large-break loss-of-coolant accident condition with special emphasis on the heat transfer phenomena in the core region. 2 - Method of solution: A so called volume-junction method is used to present fluid conservations. The primary system is divided into a number of special elements called 'control-volumes'. The system of partial differential equations describing fluid conservations for a stream-tube are integrated over a number of control volumes. The resulting set of simultaneous differential equations that is based on the assumptions of one-dimensional, homogeneous and thermal- equilibrium flow is linearized and solved for a small time increment by a simple explicit numerical technique. The one-dimensional heat conduction equations describing temperature profiles within solid material are written in finite difference forms which are linearized and solved by the Crank-Nicholson implicit method. In order to simulate the blowdown heat transfer phenomena, the code has correlation packages for heat transfer coefficient and critical heat flux. The heat generation in the core is given by a point reactor kinetics model with six groups of delayed neutrons and decay of eleven groups of fission products and actinides. The solution technique of the reactor kinetics is based on the Runge-Kutta method. ALARM-B2 has the models to simulate various components incorporated in BWRs such as jet pumps, recirculation pumps, steam separators, valves, and so on. The discharge and injection systems are modeled by leak and fill systems, respectively. 3 - Restrictions on the complexity of the problem: As this has been developed to simulate a blowdown thermo-hydraulic transient during a large break LOCA, users must pay attention when applying the code to any medium or small break LOCAs or to later phases

  14. Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio

    2002-10-01

    An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

  15. Convergence analysis of neutronic/thermohydraulic coupling behavior of SCWR

    International Nuclear Information System (INIS)

    Liu, Shichang; Cai, Jiejin

    2013-01-01

    The neutronic/thermohydraulic coupling (N–T coupling) calculations play an important role in core design and stability analysis. The traditional iterative method is not applicable for some new reactors (such as supercritical water-cooled reactor) which have intense N–T coupling behavior. In this paper, the mathematical model of N–T coupling based on fixed point theory is established firstly, with the convergent criterion, which can show the real-time convergence situation of iteration. Secondly, the self-adaptive relaxation factor and corresponding algorithm are proposed. Thirdly, the convergence analysis of the method of self-adaptive relaxation factor and common relaxation iteration has been performed, based on three calculation examples of SCWR fuel assembly. The results show that the proposed algorithm can efficiently reduce the calculation time and be adapted to different coupling cases and different initial distribution. It is easy to program, providing convenience for reactor design and analysis. This research also provides the theoretical basis for further study of N–T coupling behavior of new reactors such as SCWR

  16. Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2002-10-01

    An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

  17. Latest developments for a computer aided thermohydraulic network

    International Nuclear Information System (INIS)

    Alemberti, A.; Graziosi, G.; Mini, G.; Susco, M.

    1999-01-01

    Thermohydraulic networks are I-D systems characterized by a small number of basic components (pumps, valves, heat exchangers, etc) connected by pipes and limited spatially by a defined number of boundary conditions (tanks, atmosphere, etc). The network system is simulated by the well known computer program RELAPS/mod3. Information concerning the network geometry component behaviour, initial and boundary conditions are usually supplied to the RELAPS code using an ASCII input file by means of 'input cards'. CATNET (Computer Aided Thermalhydraulic NETwork) is a graphically user interface that, under specific user guidelines which completely define its range of applicability, permits a very high level of standardization and simplification of the RELAPS/mod3 input deck development process as well as of the output processing. The characteristics of the components (pipes, valves, pumps etc), defining the network system can be entered through CATNET. The CATNET interface is provided by special functions to compute form losses in the most typical bending and branching configurations. When the input of all system components is ready, CATNET is able to generate the RELAPS/mod3 input file. Finally, by means of CATNET, the RELAPS/mod3 code can be run and its output results can be transformed to an intuitive display form. The paper presents an example of application of the CATNET interface as well as the latest developments which greatly simplified the work of the users and allowed to reduce the possibility of input errors. (authors)

  18. Thermo-hydraulic design of earth-air heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Paepe, M. de [Ghent University (Belgium). Department of Flow, Heat and Combustion Mechanics; Janssens, A. [Ghent University (Belgium). Department of Architecture and Urbanism

    2003-05-01

    Earth-air heat exchangers, also called ground tube heat exchangers, are an interesting technique to reduce energy consumption in a building. They can cool or heat the ventilation air, using cold or heat accumulated in the soil. Several papers have been published in which a design method is described. Most of them are based on a discretisation of the one-dimensional heat transfer problem in the tube. Three-dimensional complex models, solving conduction and moisture transport in the soil are also found. These methods are of high complexity and often not ready for use by designers. In this paper, a one-dimensional analytical method is used to analyse the influence of the design parameters of the heat exchanger on the thermo-hydraulic performance. A relation is derived for the specific pressure drop, linking thermal effectiveness with pressure drop of the air inside the tube. The relation is used to formulate a design method which can be used to determine the characteristic dimensions of the earth-air heat exchanger in such a way that optimal thermal effectiveness is reached with acceptable pressure loss. The choice of the characteristic dimensions, becomes thus independent of the soil and climatological conditions. This allows designers to choose the earth-air heat exchanger configuration with the best performance. (author)

  19. Thermo-hydraulic design of earth-air heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    De Paepe, M. [Department of Flow, Heat and Combustion Mechanics, Ghent University, Ghent (Belgium); Janssens, A. [Department of Architecture and Urbanism, Ghent University, Ghent (Belgium)

    2003-07-01

    Earth-air heat exchangers, also called ground tube heat exchangers, are an interesting technique to reduce energy consumption in a building. They can cool or heat the ventilation air, using cold or heat accumulated in the soil. Several papers have been published in which a design method is described. Most of them are based on a discretisation of the one-dimensional heat transfer problem in the tube. Three-dimensional complex models, solving conduction and moisture transport in the soil are also found. These methods are of high complexity and often not ready for use by designers. In this paper, a one-dimensional analytical method is used to analyse the influence of the design parameters of the heat exchanger on the thermo-hydraulic performance. A relation is derived for the specific pressure drop, linking thermal effectiveness with pressure drop of the air inside the tube. The relation is used to formulate a design method which can be used to determine the characteristic dimensions of the earth-air heat exchanger in such a way that optimal thermal effectiveness is reached with acceptable pressure loss. The choice of the characteristic dimensions, becomes thus independent of the soil and climatological conditions. This allows designers to choose the earth-air heat exchanger configuration with the best performance. (author)

  20. Reduction of the pool-top radiation level in HANARO

    International Nuclear Information System (INIS)

    Lee, Choong-Sung; Park, Sang-Jun; Kim, Heonil; Park, Yong-Chul; Choi, Young-San

    1999-01-01

    HANARO is an open-tank-in-pool type reactor. Pool water is the only shielding to minimize the pool top radiation level. During the power ascension test of HANARO, the measured pool top radiation level was higher than the design value because some of the activation products in the coolant reached the pool surface. In order to suppress this rising coolant, the hot water layer system (HWL) was designed and installed to maintain l.2 meter-deep hot water layer whose temperature is 5degC higher than that of the underneath pool surface. After the installation of the HWL system, however, the radiation level of the pool-top did not satisfy the design value. The operation modes of the hot water layer system and the other systems in the reactor pool, which had an effect on the formation of the hot water layer, were changed to reduce pool-top radiation level. After the above efforts, the temperature and the radioactivity distribution in the pool was measured to confirm whether this system blocked the rising coolant. The radiation level at the pool-top was significantly reduced below one tenth of that before installing the HWL and satisfied the design value. It was also confirmed by calculation that this hot water layer system would significantly reduce the release of fission gases to the reactor hall and the environment during the hypothetical accident as well. (author)

  1. Comparison of the results of the fifth dynamic AER benchmark-a benchmark for coupled thermohydraulic system/three-dimensional hexagonal kinetic core models

    International Nuclear Information System (INIS)

    Kliem, S.

    1998-01-01

    The fifth dynamic benchmark was defined at seventh AER-Symposium, held in Hoernitz, Germany in 1997. It is the first benchmark for coupled thermohydraulic system/three-dimensional hexagonal neutron kinetic core models. In this benchmark the interaction between the components of a WWER-440 NPP with the reactor core has been investigated. The initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one control rod group stucking. This break causes an overcooling of the primary circuit. During this overcooling the scram reactivity is compensated and the scrammed reactor becomes re critical. The calculation was continued until the highly-borated water from the high pressure injection system terminated the power excursion. Each participant used own best-estimate nuclear cross section data. Only the initial subcriticality at the beginning of the transient was given. Solutions were received from Kurchatov Institute Russia with the code BIPR8/ATHLET, VTT Energy Finland with HEXTRAN/SMABRE, NRI Rez Czech Republic with DYN3/ATHLET, KFKI Budapest Hungary with KIKO3D/ATHLET and from FZR Germany with the code DYN3D/ATHLET.In this paper the results are compared. Beside the comparison of global results, the behaviour of several thermohydraulic and neutron kinetic parameters is presented to discuss the revealed differences between the solutions.(Authors)

  2. Development of real time visual evaluation system for sodium transient thermohydraulic experiments

    International Nuclear Information System (INIS)

    Tanigawa, Shingo

    1990-01-01

    A real time visual evaluation system, the Liquid Metal Visual Evaluation System (LIVES), has been developed for the Plant Dynamics Test Loop facility at O-arai Engineering Center. This facility is designed to provide sodium transient thermohydraulic experimental data not only in a fuel subassembly but also in a plant wide system simulating abnormal or accident conditions in liquid metal fast breeder reactors. Since liquid metal sodium is invisible, measurements to obtain experimental data are mainly conducted by numerous thermo couples installed at various locations in the test sections and the facility. The transient thermohydraulic phenomena are a result of complicated interactions among global and local scale three-dimensional phenomena, and short- and long-time scale phenomena. It is, therefore, difficult to grasp intuitively thermohydraulic behaviors and to observe accurately both temperature distribution and flow condition solely by digital data or various types of analog data in evaluating the experimental results. For effectively conducting sodium transient experiments and for making it possible to observe exactly thermohydraulic phenomena, the real time visualization technique for transient thermohydraulics has been developed using the latest Engineering Work Station. The system makes it possible to observe and compare instantly the experiment and analytical results while experiment or analysis is in progress. The results are shown by not only the time trend curves but also the graphic animations. This paper shows an outline of the system and sample applications of the system. (author)

  3. Operation and maintenance techniques of pool and pool water purification system in IMEF

    Energy Technology Data Exchange (ETDEWEB)

    Soong, Woong Sup

    1999-03-01

    IMEF pool is used pass way between pool and hot cell in order to inlet and outlet of fuel pin in cask. All operation is performed conforming with naked eyes. Therefore floating matter is filtered so as to easy under water handling. Also radioactivity in pool water is controlled according to the nuclear law, radioactivity ration maintained less than 15mR/hr on pool side. Perfect operation and maintenance can be achieved well trained operator. Result obtained from the perfection can give more influence over restrain, spreading contamination of radioactivity materials. This report describes operation and maintenance technique of pool water purification system in IMEF. (Author). 7 refs., 13 figs.

  4. Operation and maintenance techniques of pool and pool water purification system in IMEF

    International Nuclear Information System (INIS)

    Soong, Woong Sup

    1999-03-01

    IMEF pool is used pass way between pool and hot cell in order to inlet and outlet of fuel pin in cask. All operation is performed conforming with naked eyes. Therefore floating matter is filtered so as to easy under water handling. Also radioactivity in pool water is controlled according to the nuclear law, radioactivity ration maintained less than 15mR/hr on pool side. Perfect operation and maintenance can be achieved well trained operator. Result obtained from the perfection can give more influence over restrain, spreading contamination of radioactivity materials. This report describes operation and maintenance technique of pool water purification system in IMEF. (Author). 7 refs., 13 figs

  5. Effect of thermohydraulic parameter on the flux distribution and the effective multiplication factor

    International Nuclear Information System (INIS)

    Mello, J.C.; Valladares, G.L.

    1990-01-01

    The influence of two thermohydraulics parameters; the coolant flow velocity along the reactor channels and the increase of the average water temperature through the core, on the thermal flux distribution and on the effective multiplication factor, was studied in a radioisotopes production reactor. The results show that, for a fixed values of the thermohydraulics parameters reffered above, there are limits for the reactor core volume reduction for each value of the V sub(mod)/V sub(comb) ratio. These thermohydraulics conditions determine the higher termal flux value in the flux-trap and the lower value of the reactor effective multiplication factor. It is also show that there is a V sub(mod)/V sub(comb) ratio value that correspond to the higher value of the lower effective multiplication factor. These results was interpreted and comment using fundamentals concepts and relations of reactor physics. (author)

  6. Hot Flashes

    Science.gov (United States)

    Hot flashes Overview Hot flashes are sudden feelings of warmth, which are usually most intense over the face, neck and chest. Your skin might redden, as if you're blushing. Hot flashes can also cause sweating, and if you ...

  7. The Klinger hot gas double axial valve

    International Nuclear Information System (INIS)

    Kruschik, J.; Hiltgen, H.

    1984-01-01

    The Klinger hot gas valve is a medium controlled double axial valve with advanced design features and safety function. It was first proposed by Klinger early in 1976 for the PNP-Project as a containment shut-off for hot helium (918 deg. C and 42 bar), because a market research has shown that such a valve is not state of present techniques. In the first stage of development a feasibility study had to be made by detailed design, calculation and by basic experiments for key components in close collaboration with Interatom/GHT. This was the basis for further design, calculation, construction and experimental work for such a valve prototype within the new development contract. The stage of knowledge to that time revealed the following key priority development areas: Finite element stress analysis for the highly stressed high temperature main components; development of an insulation layout; Detailed experimental tests of functionally important structural components or units of the valve, partly at Klingers (gasstatic bearings, flexible metallic sealing element, aerodynamic and thermohydraulic tests), partly at Interatom (actuator unit and also gasstatic bearings), partly at HRB in Juelich (flexible metallic sealing system, aerodynamic and thermohydraulic tests); Design of a test valve for experimental work in the KVK (test circuit at Interatom) for evaluation of temperature distribution and reliability of operation; Design of a prototype and extensive testing in the KVK

  8. Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA

    International Nuclear Information System (INIS)

    Badulescu, A.; Groeneveld, D.C.

    2000-01-01

    Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)

  9. Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility

    International Nuclear Information System (INIS)

    Coragem, Helio Boemer de Oliveira

    1980-01-01

    A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

  10. HOT 2015

    DEFF Research Database (Denmark)

    Hannibal, Sara Stefansen

    2016-01-01

    HOT samler og formidler 21 literacykyndiges bud på, hvad der er hot, og hvad der bør være hot inden for literacy – og deres begrundelser for disse bud.......HOT samler og formidler 21 literacykyndiges bud på, hvad der er hot, og hvad der bør være hot inden for literacy – og deres begrundelser for disse bud....

  11. Investigation of pool thermal hydraulics and temperature distribution in inner vessel under mechanical seal leakage

    International Nuclear Information System (INIS)

    Abraham, Juby; Velusamy, K.; Selvaraj, P.

    2015-01-01

    The primary heat sink of prototype fast breeder reactor is a sodium pool which is partitioned into cold pool and hot pool. The inner vessel which separates the cold and hot pools is having penetrations for intermediate heat exchangers. The hot sodium from hot pool leaks into the cold pool through these penetrations and to reduce the leakage, mechanical seals are provided. Leakage of hot sodium into cold pool can lead to thermal stratification in the cold pool and also will affect the temperature distribution in inner vessel. 3-D CFD studies were performed focusing these features as a function of sodium leakage. The analyses indicate that the maximum temperature difference across the IV thickness is 65°C without any leakage of sodium. The temperature difference is found to decrease with increase in leakage through the seals. It is seen that a leakage of 2.5% is acceptable. (author)

  12. Studies of thermal stratification in water pool

    International Nuclear Information System (INIS)

    Verma, P.K.; Chandraker, D.K.; Nayak, A.K.; Vijayan, P.K.

    2015-01-01

    Large water pools are used as a heat sink for various cooling systems used in industry. In context of advance nuclear reactors like AHWR, it is used as ultimate heat sink for passive systems for decay heat removal and containment cooling. This system incorporates heat exchangers submerged in the large water pool. However, heat transfer by natural convection in pool poses a problem of thermal stratification. Due to thermal stratification hot layers of water accumulate over the relatively cold one. The heat transfer performance of heat exchanger gets deteriorated as a hot fluid envelops it. In the nuclear reactors, the walls of the pool are made of concrete and it may subject to high temperature due to thermal stratification which is not desirable. In this paper, a concept of employing shrouds around the heat source is studied. These shrouds provide a bulk flow in the water pool, thereby facilitating mixing of hot and cold fluid, which eliminate stratification. The concept has been applied to the a scaled model of Gravity Driven Water Pool (GDWP) of AHWR in which Isolation Condensers (IC) tubes are submerged for decay heat removal of AHWR using ICS and thermal stratification phenomenon was predicted with and without shrouds. To demonstrate the adequacy of the effectiveness of shroud arrangement and to validate the simulation methodology of RELAP5/Mod3.2, experiments has been conducted on a scaled model of the pool with and without shroud. (author)

  13. IAEA coordinated research program on `harmonization and validation of fast reactor thermomechanical and thermohydraulic codes using experimental data`. 1. Thermohydraulic benchmark analysis on high-cycle thermal fatigue events occurred at French fast breeder reactor Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    A benchmark exercise on `Tee junction of Liquid Metal Fast Reactor (LMFR) secondary circuit` was proposed by France in the scope of the said Coordinated Research Program (CRP) via International Atomic Energy Agency (IAEA). The physical phenomenon chosen here deals with the mixture of two flows of different temperature. In a LMFR, several areas of the reactor are submitted to this problem. They are often difficult to design, because of the complexity of the phenomena involved. This is one of the major problems of the LMFRs. This problem has been encountered in the Phenix reactor on the secondary loop, where defects in a tee junction zone were detected during a campaign of inspections after an operation of 90,000 hours of the reactor. The present benchmark is based on an industrial problem and deal with thermal striping phenomena. Problems on pipes induced by thermal striping phenomena have been observed in some reactors and experimental facilities coolant circuits. This report presents numerical results on thermohydraulic characteristics of the benchmark problem, carried out using a direct numerical simulation code DINUS-3 and a boundary element code BEMSET. From the analysis with both the codes, it was confirmed that the hot sodium from the small pipe rise into the cold sodium of the main pipe with thermally instabilities. Furthermore, it was indicated that the coolant mixing region including the instabilities agrees approximately with the result by eye inspections. (author)

  14. Comparison of thermohydraulic characteristics in the use of various coolants

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Suda, Kazunori; Yamaguchi, Akira

    2000-11-01

    Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiated based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1) There is no remarkable difference between liquid sodium and liquid Pb-Bi in characteristics of internal flows and free surface characteristics based on Fr number. (2) The AQUA-VOF code has a potential enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [Thermal Stratification Phenomena] (1) On-set position of thermal entrainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. On the other hand, the position in the case of CO 2 gas was shifted to upstream side with decreasing of Ri number. (2) Destruction speed of the thermal stratification interface was dependent on thermal diffusivity as fluid properties. Therefore it was concluded that an elimination method is necessary for the interface generated in CO 2 gas. [Thermal Striping Phenomena] (1) Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO 2 gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2) To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it is necessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phynomlena] (1) Fundamental behavior of the natural convection in various coolant follows buoyant jet

  15. Design of the Demineralized Water Make-up Line to Maintain the Normal Pool Water Level of the Reactor Pool in the Research Reactor

    International Nuclear Information System (INIS)

    Yoon, Hyun Gi; Choi, Jung Woon; Yoon, Ju Hyeon; Chi, Dae Young

    2012-01-01

    In many research reactors, hot water layer system (HWLS) is used to minimize the pool top radiation level. Reactor pool divided into the hot water layer at the upper part of pool and the cold part below the hot water layer with lower temperature during normal operation. Water mixing between these layers is minimized because the hot water layer is formed above cold water. Therefore the hot water layer suppresses floatation of cold water and reduces the pool top radiation level. Pool water is evaporated form the surface to the building hall because of high temperature of the hot water layer; consequently the pool level is continuously fallen. Therefore, make-up water is necessary to maintain the normal pool level. There are two way to supply demineralized water to the pool, continuous and intermittent methods. In this system design, the continuous water make-up method is adopted to minimize the disturbance of the reactor pool flow. Also, demineralized water make-up is connected to the suction line of the hot water layer system to raise the temperature of make-up water. In conclusion, make-up demineralized water with high temperature is continuously supplied to the hot water layer in the pool

  16. The state of art of the methods for thermohydraulics design of LMFBR fuel elements

    International Nuclear Information System (INIS)

    Fernandez y Fernandez, E.; Carajilescov, P.

    1981-09-01

    The present (experimental and analytical) state of art of the methods for thermohydraulics design of LMFBR fuel elements is analyzed. A development program is suggested, in order to obtain a computer code for modelling the distribution of coolant enthalpy in reactor core. This computer code is in development. (Author) [pt

  17. Thermohydraulic tests in the area of reactor safety done in CDTN

    International Nuclear Information System (INIS)

    Ladeira, L.C.D.

    1990-01-01

    The main experimental works performed in the last five years at the Thermohydraulics Laboratory of the Nuclear Technology Development Center, in the field of reactor safety are briefly described. This paper cover the performing and analysis of pressure drop, heat transfer and mixing tests in 3X3 rod bundle and rewetting tests in single tube section. (autor) [pt

  18. An overview of IPPE research on liquid metal fast reactor thermohydraulics

    International Nuclear Information System (INIS)

    Sorokin, A. P.; Efanov, A. D.; Zhukov, A. V.; Bogoslovskaia, G. P.

    2003-01-01

    The paper presents brief information on the most significant researches in the fields of liquid metal hydrodynamics and heat transfer performed in the State Scientific Center of Russian Federation 'Institute for Physics and Power Engineering' named after A.I.Leypunski applied to sodium-cooled fast reactors. Experimental methods for studying liquid metal thermohydraulics and applied measurement techniques are overviewed briefly in the paper. Some results of fundamental thermohydraulic investigations, such as quasi-universal character of velocity and temperature profile in liquid metals, if considered normally to the channel wall etc. are presented. Specific features of heat transfer in liquid metal cooled fuel subassembly are mentioned, among them there are: high level of coolant temperature; significant influence of an interchannel exchange on velocity and temperature distribution; an availability of contact thermal resistance; large azimuthal non-uniformity of velocity and temperature; 'conjugate' problem of heat transfer in combined geometry of fuel pin; an absence of stabilization of heat transfer in non-standard channels; an influence of non-uniform heat generation. Special attention is given to the temperature fields in fuel subassembly subjected to deformation because of radioactive swelling and creeping, as well as in case of blockage of a part of subassembly cross section. Some results of thermohydraulic investigation are demonstrated for intermediate heat exchangers, pressurized head collectors. Also the developed methods and codes of thermohydraulic calculations applied to fast reactor core are considered: subchannel approach, porous body model

  19. Hawaii ESI: POOLS (Anchialine Pool Points)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains sensitive biological resource data for anchialine pools in Hawaii. Anchialine pools are small, relatively shallow coastal ponds that occur...

  20. State space model extraction of thermohydraulic systems – Part I: A linear graph approach

    International Nuclear Information System (INIS)

    Uren, K.R.; Schoor, G. van

    2013-01-01

    Thermohydraulic simulation codes are increasingly making use of graphical design interfaces. The user can quickly and easily design a thermohydraulic system by placing symbols on the screen resembling system components. These components can then be connected to form a system representation. Such system models may then be used to obtain detailed simulations of the physical system. Usually this kind of simulation models are too complex and not ideal for control system design. Therefore, a need exists for automated techniques to extract lumped parameter models useful for control system design. The goal of this first paper, in a two part series, is to propose a method that utilises a graphical representation of a thermohydraulic system, and a lumped parameter modelling approach, to extract state space models. In this methodology each physical domain of the thermohydraulic system is represented by a linear graph. These linear graphs capture the interaction between all components within and across energy domains – hydraulic, thermal and mechanical. These linear graphs are analysed using a graph-theoretic approach to derive reduced order state space models. These models capture the dominant dynamics of the thermohydraulic system and are ideal for control system design purposes. The proposed state space model extraction method is demonstrated by considering a U-tube system. A non-linear state space model is extracted representing both the hydraulic and thermal domain dynamics of the system. The simulated state space model is compared with a Flownex ® model of the U-tube. Flownex ® is a validated systems thermal-fluid simulation software package. - Highlights: • A state space model extraction methodology based on graph-theoretic concepts. • An energy-based approach to consider multi-domain systems in a common framework. • Allow extraction of transparent (white-box) state space models automatically. • Reduced order models containing only independent state

  1. Spent fuel pool thermal-hydraulic analysis using RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Ramos, M. C.; Fernandes, G.H.N.; Costa, A.L.; Pereira, F.; Pereira, C., E-mail: marc5663@gmail.com, E-mail: ghnfernandes@pq.cnpq.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    In order to analyze the thermo-hydraulic behavior of spent fuel pools, and taking as reference a hypothetic PWR nuclear plant, a model of RELAP-3D for a spent fuel pool has been built. This model has been used to simulate a loss of coolant in SPF. This study focuses on the loss of coolant flow accident in spent fuel storage pool which is modelled by using RELAP5-3D code to observe the coolant level reduction and fuel uncovery because of decay heat generation of the spent fuel in the pool. The results have been compared with the available data. The developed model demonstrated that the RELAP5-3D is capable of reproduce the thermal behavior of SPF in a transient scenario. (author)

  2. Method of relative comparison of the thermohydraulic efficiency of heat exchange intensification in channels of heat-exchange surfaces

    International Nuclear Information System (INIS)

    Dubrovskij, E.V.; Vasil'ev, V.Ya.

    2002-01-01

    One introduces a technique to compare relatively thermohydraulic efficiency of heat transfer intensification in channels of heat exchange surfaces of any design types. It is shown that one should compare thermohydraulic efficiency of heat exchange intensification as to the thermal power of heat exchangers and pressure losses in channels with turbulators and in polished channels of heat exchange surfaces on the basis of dimensions of heat exchangers, their heat exchange surfaces and at similar (as to Re numbers) modes of coolant flow [ru

  3. Status of thermohydraulic studies of wire-wrapped bundles

    International Nuclear Information System (INIS)

    Khairallah, A.; Leteinturier, D.; Skok, J.

    1979-01-01

    A status review is presented of the work undertaken in CEA to acquire good understanding and description of the single-phase thermal-hydraulic problems in LMFBR wire-wrapped bundles. Design-type and reference-type calculational tools developed for the study of forced convection in nominal and distorted bundle geometries are briefly presented. Local hot spots and mixed convection situations are discussed in some more details. Out-of-pile and in-pile experimental programs designed in support to code development are described. (author)

  4. Single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3. Input data description

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-08-01

    This report explains the numerical methods and the set-up method of input data for a single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3 (Direct Numerical Simulation using a 3rd-order upwind scheme). The code was developed to simulate non-stationary temperature fluctuation phenomena related to thermal striping phenomena, developed at Power Reactor and Nuclear Fuel Development Corporation (PNC). The DINUS-3 code was characterized by the use of a third-order upwind scheme for convection terms in instantaneous Navier-Stokes and energy equations, and an adaptive control system based on the Fuzzy theory to control time step sizes. Author expect this report is very useful to utilize the DINUS-3 code for the evaluation of various non-stationary thermohydraulic phenomena in reactor applications. (author)

  5. Recommendations for analysis of stress corrosion in pipe systems exposed to thermohydraulic transients

    International Nuclear Information System (INIS)

    Bjoerndahl, Olof; Letzter, Adam; Marcinkiewicz, Jerzy; Segle, Peter

    2007-03-01

    Transient thermohydraulic events often control the design of piping systems in nuclear power plants. Water hammers due to valve closure, pressure transients caused by steam collapse and pipe break all result in structural loads that are characterised by a high frequency content. What also characterises these pressures/forces is the specific spatial and time dependence that is acting on the piping system and found in the wave propagation in the contained fluid. The aim with this project has been to develop recommendations for analysis of the stress response in piping systems subjected to thermohydraulic transients. Basis for this work is that the so called two-step-method is applied and that the structural response is calculated with modal superposition. Derived analysis criteria are based on the assumption that the associated volume strain energy in the wave propagation for the contained fluid may be well defined by a parameter, here called ε PN . The stress response in the piping system is assumed to be completely determined with certain accuracy for that part of the volume strain energy in the wave propagation associated with this parameter. A comprehensive work has been done to determine the accuracy in loadings calculated with RELAP5. Properties such as period elongation and associated spurious oscillations in the pressure wave transient have been investigated. Furthermore, has the characteristics of the artificial numerical damping in RELAP5 been identified. Based on desired accuracy of the thermohydraulic analysis together with knowledge about the duration of the thermohydraulic perturbation, the lowest upper frequency limit f Pipe , in the modal base that is required for the structure model is calculated. With perturbation is meant such as a valve closure. According to suggested criteria and with the upper frequency limit set, the essential parameters i) largest size of the elements in the structure model and ii) the largest applicable time step in the

  6. Development of a model for the primary system CAREM reactor's stationary thermohydraulic calculation

    International Nuclear Information System (INIS)

    Gaspar, C.; Abbate, P.

    1990-01-01

    The ESCAREM program oriented to CAREM reactors' stationary thermohydraulic calculation is presented. As CAREM gives variations in relation to models for BWR (Boiling Water Reactors)/PWR (Pressurized Water Reactors) reactors, it was decided to develop a suitable model which allows to calculate: a) if the Steam Generator design is adequate to transfer the power required; b) the circulation flow that occurs in the Primary System; c) the temperature at the entrance (cool branch) and d) the contribution of each component to the pressure drop in the circulation connection. Results were verified against manual calculations and alternative numerical models. An experimental validation at the Thermohydraulic Essays Laboratory is suggested. A parametric analysis series is presented on CAREM 25 reactor, demonstrating operating conditions, at different power levels, as well as the influence of different design aspects. (Author) [es

  7. A benchmark for coupled thermohydraulics system/three-dimensional neutron kinetics core models

    International Nuclear Information System (INIS)

    Kliem, S.

    1999-01-01

    During the last years 3D neutron kinetics core models have been coupled to advanced thermohydraulics system codes. These coupled codes can be used for the analysis of the whole reactor system. Although the stand-alone versions of the 3D neutron kinetics core models and of the thermohydraulics system codes generally have a good verification and validation basis, there is a need for additional validation work. This especially concerns the interaction between the reactor core and the other components of a nuclear power plant (NPP). In the framework of the international 'Atomic Energy Research' (AER) association on VVER Reactor Physics and Reactor Safety, a benchmark for these code systems was defined. (orig.)

  8. Prediction of the thermohydraulic performance of porous-media reservoirs for compressed-air energy storage

    Energy Technology Data Exchange (ETDEWEB)

    Wiles, L.E.; McCann, R.A.

    1981-09-01

    The numerical modeling capability that has been developed at the Pacific Northwest Laboratory (PNL) for the prediction of the thermohydraulic performance of porous media reservoirs for compressed air energy storage (CAES) is described. The capability of the numerical models was demonstrated by application to a variety of parametric analyses and the support analyses for the CAES porous media field demonstration program. The demonstration site analyses include calculations for the displacement of aquifer water to develop the air storage zone, the potential for water coning, thermal development in the reservoir, and the dehydration of the near-wellbore region. Unique features of the demonstration site reservoir that affect the thermohydraulic performance are identified and contrasted against the predicted performance for conditions that would be considered more typical of a commercial CAES site.

  9. Influence of the outlet air temperature on the thermohydraulic behaviour of air coolers

    Directory of Open Access Journals (Sweden)

    Đorđević Emila M.

    2003-01-01

    Full Text Available The determination of the optimal process conditions for the operation of air coolers demands a detailed analysis of their thermohydraulic behaviour on the one hand, and the estimation of the operating costs, on the other. One of the main parameters of the thermohydraulic behaviour of this type of equipment, is the outlet air temperature. The influence of the outlet air temperature on the performance of air coolers (heat transfer coefficient overall heat transfer coefficient, required surface area for heat transfer air-side pressure drop, fan power consumption and sound pressure level was investigated in this study. All the computations, using AirCooler software [1], were applied to cooling of the process fluid and the condensation of a multicomponent vapour mixture on two industrial devices of known geometries.

  10. DRESDYN: A new platform for liquid metal thermohydraulic studies and measurement technique developments

    International Nuclear Information System (INIS)

    Gerbeth, Gunter; Eckert, Sven; Stefani, Frank; Gundrum, Thomas

    2013-01-01

    DRESDYN: General features. DRESDYN: DREsden Sodium facility for DYNamo and thermohydraulic studies. A large-scale new infrastructure for liquid metal experiments. Features: • New building ~ 500 m 2 ; • Total sodium inventory: 12-15 tons; • Precession driven experiment with separate strong basement and containment for Argon flooding; • Big hall for SFR related experiments, including ISI, a sodium loop, X-ray lab; • Financing is given, construction will start soon in spring 2013; • First experiments 2015 (hopefully...)

  11. Application of CFD methods in research of SCWR thermo-hydraulics

    International Nuclear Information System (INIS)

    Zeng Xiaokang; Li Yongliang; Yan Xiao; Xiao Zejun; Huang Yanping

    2013-01-01

    The CFD method has been an important tool in the research of SCWR thermo- hydraulics. Currently, the CFD methods uses commonly the subcritical turbulence models, which can not accurately simulate the gravity and thermal expansion acceleration effect, and CFD numerical method is not applicable when the heat flux is large. The paper summarizes the application status of the CFD methods in the research of SCWR thermo-hydraulics in RETH. (authors)

  12. A model development for a thermohydraulic calculation material convection of MTR (Materials Testing Reactors)

    International Nuclear Information System (INIS)

    Abbate, P.

    1990-01-01

    The CONVEC program developed for the thermohydraulic calculation under a natural convection regime for MTR type reactors is presented. The program is based on a stationary, one dimensional model of finite differences that allow to calculate the temperatures of cooler, cladding and fuel as well as the flow for a power level specified by the user. This model has been satisfactorily validated by a water cooling (liquid phase) and air system. (Author) [es

  13. Evaluation of two-fluid and drift flux thermohydraulics in APROS code environment

    International Nuclear Information System (INIS)

    Miettinen, J.; Karppinen, I.; Haenninen, M.; Ylijoki, J.

    1999-01-01

    The characteristics of the thermohydraulic solutions in APROS are considered for the nuclear power plant modelling. The thermohydraulic model of the APROS plant analyzer includes three levels of solutions, homogeneous 3-equation model, 5-equation drift flux model and 6-equation two-fluid model. In practical modelling of versatile process systems different approaches are selected for different types of the power plant sections. The 3-equation model is used for turbines and auxiliary systems. The 5-equation model and 6-equation model are alternative models for main process sections of the primary and secondary sides. The 5-equation model has been typically selected for the real time applications and the 6-equation model for the safety analysis applications. The validation needs for both approaches are the same. Because the change of the solution mode is an easy task in APROS, the validation tasks are typically performed in parallel for 5-equation and 6-equation models. By calculating in parallel with both models systematic errors in solutions may be pointed out. The testing against both separate effects tests and integral tests is an essential part in the thermohydraulics. In different plant applications different physical features are important. The analysis requirements vary from one application to another. When nodalizations together with increased computer speed are growing up, the earlier validation cases may be insufficient. That is why the content of the code has to be known in detail. Such an expertise in the code development has to be gained that properties of the code against other thermohydraulics codes are known. (author)

  14. Preliminary study of the thermo-hydraulic behaviour of the binary breeder reactor

    International Nuclear Information System (INIS)

    Silveira Luz, M. da; Ferreira, W.J.

    1984-06-01

    Continuing the development of the Binary Breeder Reactor, its physical configuration and the advantages of differents types of spacers are analysed. In order to simulate the thermo-hydraulic behaviour and obtain data for a preliminary evaluation of the core geometry, the COBRA III C code was used to study the effects of the lenght and diameter of the fuel element, the coolant inlet temperature, the system pressure, helicoidal pitch and the pitch to diameter ratio. (Author) [pt

  15. Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA

    International Nuclear Information System (INIS)

    Sugimoto, Jun

    1989-01-01

    This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)

  16. Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

    International Nuclear Information System (INIS)

    Wiles, L.E.; McCann, R.A.

    1981-09-01

    This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

  17. Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall

    International Nuclear Information System (INIS)

    Kim, Byoung-Yoon; Ahn, Hee-Jae

    2011-01-01

    The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.

  18. HOT 2012

    DEFF Research Database (Denmark)

    Lund, Henriette Romme

    Undersøgelse af, hvad der er hot - og hvad der burde være hot på læseområdet med 21 læsekyndige. Undersøgelsen er gennemført siden 2010. HOT-undersøgelsen er foretaget af Nationalt Videncenter for Læsning - Professionshøjskolerne i samarb. med Dansklærerforeningen......Undersøgelse af, hvad der er hot - og hvad der burde være hot på læseområdet med 21 læsekyndige. Undersøgelsen er gennemført siden 2010. HOT-undersøgelsen er foretaget af Nationalt Videncenter for Læsning - Professionshøjskolerne i samarb. med Dansklærerforeningen...

  19. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    International Nuclear Information System (INIS)

    2014-08-01

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

  20. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-08-15

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

  1. HOT 2014

    DEFF Research Database (Denmark)

    Lund, Henriette

    Undersøgelse af, hvad der er hot - og hvad der burde være hot på læseområdet med 21 læsekyndige. Undersøgelsen er gennemført siden 2010. HOT-undersøgelsen er foretaget af Nationalt Videncenter for Læsning - Professionshøjskolerne i samarb. med Dansklærerforeningen...

  2. HOT 2011

    DEFF Research Database (Denmark)

    Lund, Henriette Romme

    En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager 21 læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet.......En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager 21 læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet....

  3. Swimming pool cleaner poisoning

    Science.gov (United States)

    Swimming pool cleaner poisoning occurs when someone swallows this type of cleaner, touches it, or breathes in ... The harmful substances in swimming pool cleaner are: Bromine ... copper Chlorine Soda ash Sodium bicarbonate Various mild acids

  4. Swimming pool granuloma

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/001357.htm Swimming pool granuloma To use the sharing features on this page, please enable JavaScript. A swimming pool granuloma is a long-term (chronic) skin ...

  5. On the phenomenon of the reversal of the cooling current in the hot pipes of a swimming-pool type pile cooled by forced convection; Sur un phenomene de renversement du courant de refrigeration dans les canaux chauds d'une pile piscine refroidie en convection forcee

    Energy Technology Data Exchange (ETDEWEB)

    Boure, J [Commissariat a l' Energie Atomique, Grenoble (France).Centre d' Etudes Nucleaires

    1961-07-01

    It is shown, for a swimming-pool type pile cooled by forced convection (general flow downwards), that a permanently stable regime with downward flow in all the channels is not possible when the flow is below a critical value for a given power. In the hot channels the natural convection then becomes preponderant, the direction of the flow is reversed and a permanently stable regime exists for which the flow is upwards in the hot channels. Calculations are made, with simplifying hypotheses in the case of Melusine. (author) [French] Pour une pile piscine refrigeree en convection forcee (ecoulement global descendant), on montre qu'un regime permanent stable avec ecoulement descendant dans tous les canaux est impossible lorsque le debit est inferieur a une valeur critique pour une puissance donnee. Dans les canaux chauds, la convection naturelle l'emporte alors, le sens du courant s'inverse et un regime permanent stable existe, pour lequel le courant est ascendant dans les canaux chauds. On fait les calculs, avec des hypotheses simplificatrices, dans le cas de Melusine. (auteur)

  6. Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

    International Nuclear Information System (INIS)

    Kobayashi, Michiyuki; Aya, Izuo; Inasaka, Fujio; Murata, Hiroyuki; Odano, Naoteru; Shiozaki, Koki

    1998-01-01

    A research project from 1995-1999 had a plan to make experimental studies on (1) safety of nuclear ship loaded with an integral ship propulsion reactor (2) effects of pulsating flow on the thermo-hydraulic characteristics of ship propulsion reactor and (3) thermo-hydraulic behaviors of the reactor container at the time of accident in a passively safe ship propulsion reactor. Development of a data base for ship propulsion reactor was attempted using previous experimental data on the thermo-hydraulic characteristics of the reactor in the institute in addition to the present results aiming to make general analytical evaluation for the safety of the engineering-simulation system for nuclear ship. A general data base was obtained by integrating the data list and the analytical program for static characteristics. A test equipment which allows to visualize the pulsating flow was produced and visualization experiments have started. (M.N.)

  7. Solar swimming pool

    Energy Technology Data Exchange (ETDEWEB)

    1985-01-01

    This report examines the feasibility of using solar collectors to heat the water in a previously unheated outdoor swimming pool. The solar system is used in conjunction with a pool blanket, to conserve heat when the pool is not in use. Energy losses through evaporation can be reduced by as much as 70% by a pool blanket. A total of 130 m{sup 2} of highly durable black synthetic collectors were installed on a support structure at a 30{degree} angle from the horizontal, oriented to the south. Circulation of pool water though the collectors, which is controlled by a differential thermostat, was done with the existing pool pump. Before installation the pool temperature averaged 16{degree}C; after installation it ranged from 20{degree} to 26{degree}C. It was hard to distinguish how much pool heating was due to the solar system and how much heat was retained by the pool blanket. However, the pool season was extended by five weeks and attendance tripled. 2 figs.

  8. Experimental study on thermo-hydraulic instability on reduced-moderation natural circulation BWR concept

    International Nuclear Information System (INIS)

    Watanabe, Noriyuki; Subki, M.H.; Kikura, Hiroshige; Aritomi, Masanori

    2003-01-01

    Reduced-moderation natural circulation BWR has been promoted to solve the recent challenges in BWR nuclear power technology problems as one of advanced small and medium-sized reactors equipped with the passive safety features in conformity with the natural law. However, the elimination of recirculation pumps and a high-density core due to the increase of conversion ratio could cause various thermo-hydraulic instabilities especially during the start-up stage. The occurrences of the thermo-hydraulic instabilities are not desirable and it is one of the main challenges in establishing reduced-moderation natural circulation BWR as a commercial reactor. The purpose of this present study is to experimentally investigate the driving mechanism of the thermo-hydraulic instabilities and the effect of system pressure on the unstable flow patterns. Hence, as the fundamental research for this study, a natural circulation loop that carries boiling fluid with parallel boiling channel has been constructed. Channel gap that has been set at 2 mm in order to simulate reduced-moderation reactor core. Pressure ranges of 0.1 up to 0.7 MPa, input heat flux range of 0 ou to 577 kW/m 2 , and inlet subcooling temperatures of 5, 10, and 15 K respectively, are imposed in the experiments. This experiment clarifies that changes in unstable flow patterns with increase in heat flux can be classified into two in response to system pressure range. In case of atmospheric pressure, unstable flow patters has been classified in beyond order, (1) in-phase geysering, (2) transition oscillation combined with both features of in-phase geysering and natural circulation oscillation, (3) natural circulation oscillation induced by hydrostatic head fluctuation, (4) density wave oscillation, and finally (5) stable boiling two-phase flow. On the other hand, in the system pressure range from 0.2 to 0.7 MPa, unstable patters have been dramatically changed in the following order (1) out-of-phase geysering, (2

  9. The significance of thermohydraulic conditions for the corrosion safety of PWR steam generators

    International Nuclear Information System (INIS)

    Gulich, J.F.

    1975-04-01

    In several PWR nuclear power plants leakages have occurred in the steam generator which were caused by localised corrosion attack. While the attention of manufacturers and operators is focused on the influences of feedwater chemistry and tube material, the present work highlights the fact that the damage always occurred in those places where flow regimed are poorly defined. The investigation leads to the result that local dry out of the heating surface can be contributing cause of damage. A method is indicated for estimating the thermohydraulic conditions in the inflow region over the tube plate and measures to improve corrosion safety are discussed. (author)

  10. Validation of thermohydraulic codes by comparison of experimental results with computer simulations

    International Nuclear Information System (INIS)

    Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

    1989-01-01

    The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

  11. ASCOT-1, Thermohydraulics of Axisymmetric PWR Core with Homogeneous Flow During LOCA

    International Nuclear Information System (INIS)

    1978-01-01

    1 - Nature of the physical problem solved: ASCOT-1 is used to analyze the thermo-hydraulic behaviour in a PWR core during a loss-of-coolant accident. 2 - Method of solution: The core is assumed to be axisymmetric two-dimensional and the conservation laws are solved by the method of characteristics. For the temperature response of fuel in the annular regions into which the core is divided, the heat conduction equations are solved by an explicit method with averaged flow conditions. 3 - Restrictions on the complexity of the problem: Axisymmetric two-dimensional homogeneous flows

  12. Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, A.K.

    1978-02-01

    Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation.

  13. Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)

    International Nuclear Information System (INIS)

    Agrawal, A.K.

    1978-02-01

    Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation

  14. Reliability analysis of PWR thermohydraulic design by the Monte Carlo method

    International Nuclear Information System (INIS)

    Silva Junior, H.C. da; Berthoud, J.S.; Carajilescov, P.

    1977-01-01

    The operating power level of a PWR is limited by the occurence of DNB. Without affecting the safety and performance of the reactor, it is possible to admit failure of a certain number of core channels. The thermohydraulic design, however, is affect by a great number of uncertainties of deterministic or statistical nature. In the present work, the Monte Carlo method is applied to yield the probability that a number F of channels submitted to boiling crises will not exceed a number F* previously given. This probability is obtained as function of the reactor power level. (Author) [pt

  15. Fundamental study on thermo-hydraulic behaviors during power transient, 2

    International Nuclear Information System (INIS)

    Shinano, M.; Inoue, A.

    1988-01-01

    Thermo-hydraulic behaviors during power transient of nuclear reactors are studied. Boiling around test rod heated transiently forces to flow out liquid in the test section and generates high pressure pulse. In this study, it is investigated experimentally and analytically that magnitude of pressure pulse and energy conversion efficiency to the mechanical works in cases of fragmentation and non-fragmentation. In analysis, effects of increasing of heat transfer and of interaction area due to fragmentation is considered. Consequently, 1) magnitude of pressure pulse on fragmentation is about 10 times greater than that on non-fragmentation. 2) analytical model can show characteristics of fragmentation processes qualitatively. (author)

  16. HOT 2010

    DEFF Research Database (Denmark)

    Lund, Henriette Romme

    En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager en række læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet. Undersøgelsen er gentaget hvert år siden 2010.......En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager en række læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet. Undersøgelsen er gentaget hvert år siden 2010....

  17. HOT 2013

    DEFF Research Database (Denmark)

    Lund, Henriette Romme

    En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager en række læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet. Undersøgelsen er gentaget hvert år siden 2010.......En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager en række læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet. Undersøgelsen er gentaget hvert år siden 2010....

  18. Comparative Analysis of Thermohydraulic Margins in Embalse Power Station, CARA Vs. CANDU with Cobra IV-HW

    International Nuclear Information System (INIS)

    Daverio, H; Juanico, L

    2000-01-01

    Comparative analysis of thermohydraulic margins were studied of the CANDU 37 and CARA fuel bundles (FB) in Embalse power station with COBRA IV-HW code ., the geometry of the bundle laying on the channel was particularly modeled and discussing the results in comparison with former calculations with 1/6 simetry .The CARA design with enriched uranium (0.9 %) and extended burn up lets maintain the current thermohydraulic nominal margins , while compared with CANDU 37 rods FB enriched , the CARA design permits widely improve the current margins

  19. One-dimensional thermohydraulic code THESEUS and its application to chilldown process simulation in two-phase hydrogen flows

    Science.gov (United States)

    Papadimitriou, P.; Skorek, T.

    THESUS is a thermohydraulic code for the calculation of steady state and transient processes of two-phase cryogenic flows. The physical model is based on four conservation equations with separate liquid and gas phase mass conservation equations. The thermohydraulic non-equilibrium is calculated by means of evaporation and condensation models. The mechanical non-equilibrium is modeled by a full-range drift-flux model. Also heat conduction in solid structures and heat exchange for the full spectrum of heat transfer regimes can be simulated. Test analyses of two-channel chilldown experiments and comparisons with the measured data have been performed.

  20. THYDE-P, PWR LOCA Thermohydraulic Transient Analysis

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2001-01-01

    after all the normal and linkage nodes. (d) Among junctions, normal and guillotine break junctions should be numbered first. Then the mixing junctions should be numbered according to the direction of the steady-state flow. After these, the injection junctions and finally the dead-end junctions should be numbered. (e) In the present version of THYDE-P, it is required that either of the hot leg nodes adjacent to the upper plenum mixing junction must be numbered as one and that the upper plenum should be numbered first among the mixing junctions

  1. Swimming-pool piles

    International Nuclear Information System (INIS)

    Trioulaire, M.

    1959-01-01

    In France two swimming-pool piles, Melusine and Triton, have just been set in operation. The swimming-pool pile is the ideal research tool for neutron fluxes of the order of 10 13 . This type of pile can be of immediate interest to many research centres, but its cost must be reduced and a break with tradition should be observed in its design. It would be an advantage: - to bury the swimming-pool; - to reject the experimental channel; - to concentrate the cooling circuit in the swimming-pool; - to carry out all manipulations in the water; - to double the core. (author) [fr

  2. A novel thermo-hydraulic coupling model to investigate the crater formation in electrical discharge machining

    Science.gov (United States)

    Tang, Jiajing; Yang, Xiaodong

    2017-09-01

    A novel thermo-hydraulic coupling model was proposed in this study to investigate the crater formation in electrical discharge machining (EDM). The temperature distribution of workpiece materials was included, and the crater formation process was explained from the perspective of hydrodynamic characteristics of the molten region. To better track the morphology of the crater and the movement of debris, the level-set method was introduced in this study. Simulation results showed that the crater appears shortly after the ignition of the discharge, and the molten material is removed by vaporizing in the initial stage, then by splashing at the following time. The driving force for the detachment of debris in the splashing removal stage comes from the extremely large pressure difference in the upper part of the molten region, and the morphology of the crater is also influenced by the shearing flow of molten material. It was found that the removal ratio of molten material is only about 7.63% under the studied conditions, leaving most to form the re-solidification layer on the surface of the crater. The size of the crater reaches the maximum at the end of discharge duration then experiences a slight reduction because of the reflux of molten material after the discharge. The results of single pulse discharge experiments showed that the morphologies and sizes between the simulation crater and actual crater are good at agreement, verifying the feasibility of the proposed thermo-hydraulic coupling model in explaining the mechanisms of crater formation in EDM.

  3. On three-dimensional nuclear thermo-hydraulic computation techniques for ATR

    International Nuclear Information System (INIS)

    1997-08-01

    The three-dimensional computation code for nuclear thermo-hydraulic combination core LAYMON-2A is used for the calculation of the power distribution and the control rod reactivity value of the ATR. This code possesses various functions which are required for planning the core operation such as the search function for critical boric acid concentration, and can do various simulation calculations such as core burning calculation. Further, the three-dimensional analysis code for xenon dynamic characteristics in the core LAYMON-2C, in which the dynamic characteristic equation of xenon-samarium was incorporated into the LAYMON-2A code can take the change with time lapse of xenon-samarium concentration accompanying the change of power level and power distribution into account, and it is used for the analysis of the spatial vibration characteristics of power and the regional power control characteristics due to xenon in the core. As to the LAYMON-2A, the computation flow, power distribution and thermo-hydraulic computation models, and critical search function are explained. As to the LAYMON-2C, the computation flow is described. The comparison of the calculated values by using the LAYMON-2A code and the operation data of the Fugen is reported. (K.I.)

  4. Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

    International Nuclear Information System (INIS)

    Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

    1988-04-01

    As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

  5. EBaLM-THP - A neural network thermohydraulic prediction model of advanced nuclear system components

    International Nuclear Information System (INIS)

    Ridluan, Artit; Manic, Milos; Tokuhiro, Akira

    2009-01-01

    In lieu of the worldwide energy demand, economics and consensus concern regarding climate change, nuclear power - specifically near-term nuclear power plant designs are receiving increased engineering attention. However, as the nuclear industry is emerging from a lull in component modeling and analyses, optimization for example using ANN has received little research attention. This paper presents a neural network approach, EBaLM, based on a specific combination of two training algorithms, error-back propagation (EBP), and Levenberg-Marquardt (LM), applied to a problem of thermohydraulics predictions (THPs) of advanced nuclear heat exchangers (HXs). The suitability of the EBaLM-THP algorithm was tested on two different reference problems in thermohydraulic design analysis; that is, convective heat transfer of supercritical CO 2 through a single tube, and convective heat transfer through a printed circuit heat exchanger (PCHE) using CO 2 . Further, comparison of EBaLM-THP and a polynomial fitting approach was considered. Within the defined reference problems, the neural network approach generated good results in both cases, in spite of highly fluctuating trends in the dataset used. In fact, the neural network approach demonstrated cumulative measure of the error one to three orders of magnitude smaller than that produce via polynomial fitting of 10th order

  6. Analysis of the influences of thermal correlations on neutronic–thermohydraulic coupling calculation of SCWR

    International Nuclear Information System (INIS)

    Xu, Weifeng; Cai, Jiejin; Liu, Shichang; Tang, Qi

    2015-01-01

    Highlights: • Different thermal correlations for supercritical water are summarized. • Influences of thermal correlations on neutronic–thermohydraulic coupling calculation are analyzed. • Sensitivity analysis has been done for the thermal correlations. - Abstract: The neutronic–thermohydraulic coupling (N–T coupling) calculation is important on core design, security and stability analysis of supercritical water-coolant reactor (SCWR), and a suitable thermal correlation is also necessary for the N–T coupling calculation. In this paper, the scheme of the U.S. SCWR design and the process of the N–T coupling will be introduced as well as some of different thermal correlations firstly. Then, based on the N–T coupling system ARNT, the U.S. SCWR design is simulated to analyze the influences of thermal correlations on N–T coupling calculation of SCWR so as to find out which correlation is best. The result shows that all thermal correlations are suitable. However, using different correlations for calculation leads to a great difference in safety margin of SCWR. What's more, the Bishop and Jackson correlations are more suitable and conservative, but the Griem correlation is not very precise. And the effect of buoyancy lift makes little influence on the calculation of heat transfer of SCWR. This research is also of great significance for the further study of N–T coupling of SCWR

  7. Thermo-hydraulic analysis of the generic equatorial port plug design

    International Nuclear Information System (INIS)

    Rodríguez, E.; Guirao, J.; Ordieres, J.; Cortizo, J.L.; Iglesias, S.

    2012-01-01

    Highlights: ► Thermo-hydraulic transient performance evaluation and optimization of the GEPP structure cooling/heating system under neutronic heating and baking conditions. ► The optimization of the GEPP box structure's cooling system includes positioning and minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions. - Abstract: The port-based ITER diagnostic systems are housed primarily in two locations, the equatorial and upper port plugs. The port plug structure provides confinement function, maintains ultra-high vacuum quality and the first confinement barrier for radioactive materials at the ports. The port plug structure design, from the ITER International Organisation (IO), is cooled and heated by pressurized water which flows through a series of gun-drilled water channels and water pipes. The cooling function is required to remove nuclear heating due to radiation during operation of ITER, while the heating function is intended to heat up uniformly the machine during baking condition. The work presented provides coupled thermo-hydraulic analysis and optimization of a Generic Equatorial Port Plug (GEPP) structure cooling and heating system. The optimization performed includes positioning, minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions.

  8. Development of a computer code for thermohydraulic analysis of a heated channel in transients

    International Nuclear Information System (INIS)

    Jafari, J.; Kazeminejad, H.; Davilu, H.

    2004-01-01

    This paper discusses the thermohydraulic analysis of a heated channel of a nuclear reactor in transients by a computer code that has been developed by the writer. The considered geometry is a channel of a nuclear reactor with cylindrical or planar fuel rods. The coolant is water and flows from the outer surface of the fuel rod. To model the heat transfer in the fuel rod, two dimensional time dependent conduction equations has been solved by combination of numerical methods, O rthogonal Collocation Method in radial direction and finite difference method in axial direction . For coolant modelling the single phase time dependent energy equation has been used and solved by finite difference method . The combination of the first module that solves the conduction in the fuel rod and a second one that solved the energy balance in the coolant region constitute the computer code (Thyc-1) to analysis thermohydraulic of a heated channel in transients. The Orthogonal collocation method maintains the accuracy and computing time of conventional finite difference methods, while the computer storage is reduced by a factor of two. The same problem has been modelled by RELAP5/M3 system code to asses the validity of the Thyc-1 code. The good agreement of the results qualifies the developed code

  9. Thermohydraulic and safety analysis on China advanced research reactor under station blackout accident

    International Nuclear Information System (INIS)

    Tian Wenxi; Qiu Suizheng; Su Guanghui; Jia Dounan; Liu Xingmin; Zhang Jianwei

    2007-01-01

    A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules

  10. Validation and verification of the MTR{sub P}C thermohydraulic package

    Energy Technology Data Exchange (ETDEWEB)

    Doval, Alicia [INVAP S.E., Bariloche, Rio Negro (Argentina). Nuclear Engineering Dept.]. E-mail: doval@invap.com.ar

    1998-07-01

    The MTR{sub P}C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)

  11. Spent fuel storage pool

    International Nuclear Information System (INIS)

    Murakami, Naoshi.

    1996-01-01

    Fences are disposed to a fuel exchange floor surrounding the upper surface of a fuel pool for preventing overflow of pool water. The fences comprise a plurality of flat boards arranged in parallel with each other in the longitudinal direction while being vertically inclined, and slits are disposed between the boards for looking down the pool. Further, the fences comprise wide boards and are constituted so as to be laid horizontally on the fuel exchange floor in a normal state and uprisen by means of the signals from an earthquake sensing device. Even if pool water is overflow from the fuel pool by the vibrations occurred upon earthquake and flown out to the floor of the fuel exchange floor, the overflow from the fuel exchange floor is prevented by the fences. An operator who monitors the fuel pool can observe the inside of the fuel pool through the slits formed to the fences during normal operation. The fences act as resistance against overflowing water upon occurrence of an earthquake thereby capable of reducing the overflowing amount of water due to the vibrations of pool water. The effect of preventing overflowing water can be enhanced. (N.H.)

  12. Collaborative Car Pooling System

    OpenAIRE

    João Ferreira; Paulo Trigo; Porfírio Filipe

    2009-01-01

    This paper describes the architecture for a collaborative Car Pooling System based on a credits mechanism to motivate the cooperation among users. Users can spend the accumulated credits on parking facilities. For this, we propose a business model to support the collaboration between a car pooling system and parking facilities. The Portuguese Lisbon-s Metropolitan area is used as application scenario.

  13. Radiation heat transfer model in a spent fuel pool by TRACE code

    International Nuclear Information System (INIS)

    Sanchez-Saez, F.; Carlos, S.; Villanueva, J.F.; Martorell, S.

    2014-01-01

    Nuclear policies have experienced an important change since Fukushima Daiichi nuclear plant accident and the safety of spent fuels has been in the spot issue among all the safety concerns. The work presented consists of the thermohydraulic simulation of spent fuel pool behavior after a loss of coolant throughout transfer channel with loss of cooling transient is produced. The simulation is done with the TRACE code. One of the most important variables that define the behavior of the pool is cladding temperature, which evolution depends on the heat emission. In this work convection and radiation heat transfer is considered. When both heat transfer models are considered, a clear delay in achieving the maximum peak cladding temperature (1477 K) is observed compared with the simulation in which only convection heat transfer is considered. (authors)

  14. Fundamental study on thermo-hydraulics during start-up in natural circulation boiling water reactors, (1)

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Chiang Jing-Hsien; Takahashi, Tohru; Wataru, Masumi; Mori, Michitsugu.

    1992-01-01

    Recently, many concepts, in which passive and simplified functions are actively adapted, have been proposed for the next generation LWRs. The natural circulation BWR is one such considered from the requirements for next generation LWRs as compared with current BWRs. It is pointed out from this consideration that a thermo-hydraulic instability, which may appear during start-up, greatly influences concept feasibility because its occurence makes operation for raising power output difficult. Thermo-hydraulic instabilities are investigated experimentally under conditions simulating normal and abnormal start-up processes. It is clarified that three kinds of thermo-hydraulic instabilities may occur during start-up in the natural circulation BWR according to its procedure and reactor configuration, which are (1) geysering induced by condensation, (2) natural circulation instability induced by hydrostatic head fluctuation in steam separators and (3) density wave instability. Driving mechanisms of the geysering and the natural circulation instability, which have never understood enough, are inferred from the results. Finally, the difference of thermo-hydraulic behavior during start-up processes between thermal natural circulation boilers and the Dodewaard reactor is discussed. (author)

  15. Numerical simulation of fuel assembly thermohydraulics of fast reactors with the partial blockage of cross section under the coolant

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sorokin, A.P.

    2000-01-01

    The problems of numerical modeling of thermohydraulics in assembly of fuel elements of fast reactors with the partial blockage of cross-section under the coolant are considered. The information about existing codes constructed on use of subchannel technique and model of porous body are presented. The results of calculation obtained by these codes are presented. (author)

  16. 13 CFR 120.611 - Pools backing Pool Certificates.

    Science.gov (United States)

    2010-01-01

    ... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Pools backing Pool Certificates. 120.611 Section 120.611 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Secondary Market Certificates § 120.611 Pools backing Pool Certificates. (a) Pool characteristics. As set...

  17. Feasibility study on large pool-type LMFBR

    International Nuclear Information System (INIS)

    1984-01-01

    A feasibility study has been conducted from 1981 FY to 1983 FY, in order to evaluate the feasibility of a large pool-type LMFBR under the Japanese seismic design condition and safety design condition, etc. This study was aimed to establish an original reactor structure concept which meets those design conditions especially required in Japan. In the first year, preceding design concepts had been reviewed and several concepts were originated to be suitable to Japan. For typical two of them being selected by preliminary analysis, test programs were planned. In the second year, more than twenty tests with basic models had been conducted under severe conditions, concurrently analytical approaches were promoted. In the last year, larger model tests were conducted and analytical methods have been verified concerning hydrodynamic effects on structure vibration, thermo-hydraulic behaviours in reactor plena and so on. Finally the reactor structure concepts for a large pool-type LMFBR have been acknowledged to be feasible in Japan. (author)

  18. PDA: Pooled DNA analyzer

    Directory of Open Access Journals (Sweden)

    Lin Chin-Yu

    2006-04-01

    Full Text Available Abstract Background Association mapping using abundant single nucleotide polymorphisms is a powerful tool for identifying disease susceptibility genes for complex traits and exploring possible genetic diversity. Genotyping large numbers of SNPs individually is performed routinely but is cost prohibitive for large-scale genetic studies. DNA pooling is a reliable and cost-saving alternative genotyping method. However, no software has been developed for complete pooled-DNA analyses, including data standardization, allele frequency estimation, and single/multipoint DNA pooling association tests. This motivated the development of the software, 'PDA' (Pooled DNA Analyzer, to analyze pooled DNA data. Results We develop the software, PDA, for the analysis of pooled-DNA data. PDA is originally implemented with the MATLAB® language, but it can also be executed on a Windows system without installing the MATLAB®. PDA provides estimates of the coefficient of preferential amplification and allele frequency. PDA considers an extended single-point association test, which can compare allele frequencies between two DNA pools constructed under different experimental conditions. Moreover, PDA also provides novel chromosome-wide multipoint association tests based on p-value combinations and a sliding-window concept. This new multipoint testing procedure overcomes a computational bottleneck of conventional haplotype-oriented multipoint methods in DNA pooling analyses and can handle data sets having a large pool size and/or large numbers of polymorphic markers. All of the PDA functions are illustrated in the four bona fide examples. Conclusion PDA is simple to operate and does not require that users have a strong statistical background. The software is available at http://www.ibms.sinica.edu.tw/%7Ecsjfann/first%20flow/pda.htm.

  19. Swimming Pool Safety

    Science.gov (United States)

    ... Spread the Word Shop AAP Find a Pediatrician Safety & Prevention Immunizations All Around At Home At Play ... Español Text Size Email Print Share Swimming Pool Safety Page Content ​What is the best way to ...

  20. Vitamin D Pooling Project

    Science.gov (United States)

    The Vitamin D Pooling Project of Rarer Cancers brought together investigators from 10 cohorts to conduct a large prospective epidemiologic study of the association between vitamin D status and seven rarer cancers.

  1. Swimming pool special; Zwembadspecial

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-05-15

    This issue includes a few articles and messages on the use of heat pump systems in swimming pools. [Dutch] Dit nummer bevat onder meer een paar artikelen over het gebruik van warmtepompsystemen in zwembaden.

  2. Computer code HYDRO-ACE for analyzing thermo-hydraulic phenomena in the BWR core

    International Nuclear Information System (INIS)

    Abe, Kiyoharu; Naito, Yoshitaka

    1979-10-01

    A computer code HYDRO-ACE has been developed for analyzing thermo-hydraulic phenomena in the BWR core under forced or natural circulation of cooling water. The code is composed of two main calculation routines for single channels such as riser, separator, and downcommer and multiple channels such as the reactor core with a heated zone. Functionally the code is divided into many subroutines to be connected straightforwardly, and so that the user can choose a given course freely by simply arranging the subroutines. In the program, void fraction is calculated by Maurer's method, two-phase frictional pressure drop by Maltinelli-Nelson's, and critical heat flux ratio by Hench-Levy's. The coolant flow distributions in the JPDR-II core calculated by the code are in good agreement with those measured. (author)

  3. Analysis of natural circulation stability in a low pressure thermohydraulic test loop

    International Nuclear Information System (INIS)

    Jafari, J.; D'Auria, F.; Kazeminejad, H.; Davilu, H.

    2002-01-01

    This paper discusses an instability study of a natural circulation (NC) loop performed with the aid of Relap5 thermal-hydraulic system code. This loop has been designed and constructed for the analysis of relevant thermohydraulic parameters of a nuclear reactor. In this study, the main parameters for the stability of NC are identified and characterized through the execution of proper code runs. The obtained stability boundary (SB) in the dimensionless Zuber- Sub-cooling plane is compared with the SB reported in referenced literature. The agreement of predicted NC stability boundaries with the results of independent studies demonstrates both the capability of the mentioned code in assessing NC loop stability and the quality of the performed calculations.(author)

  4. The delay function in finite difference models for nuclear channels thermo-hydraulic transients

    International Nuclear Information System (INIS)

    Agazzi, A.

    1977-01-01

    The study of the thermo-hydraulic transients in a nuclear reactor core often requires a bi- or tri-dimensional mathematical simulation of a reactor channel. The equations involved are generally solved by means of finite-difference methods. The determination of the spatial mesh-width and the time interval is strongly conditioned by the necessity of a good accuracy in the description of the delay function which defines the transfer of thermal perturbations along the cooling channel. In this paper the effects of both space and time discretization on the delay function are considered and for the classical cases of inlet temperature step and ramp universal functions and diagrams are given in order to make possible the determination of optimal spatial mesh-width and time interval, once the requested accuracy of the model is fixed in advance

  5. Modifications in Compacted MX-80 Bentonite Due to Thermo-Hydraulic Treatment

    International Nuclear Information System (INIS)

    Gomez-Espina, R.; Villar, M. V.

    2013-01-01

    The thermo-hydraulic tests reproduce the thermal and hydraulic conditions to which bentonite is subjected in the engineered barrier of a deep geological repository of radioactive waste. The results of thermo-hydraulic test TBT1500, which was running for approximately 1500 days, are presented. This is a continuation to the Technical Report Ciemat 1199, which presented results of test TBT500, performed under similar conditions but with duration of 500 days. In both tests the MX-80 bentonite was used with initial density and water content similar to those of the large-scale test TBT. The bentonite column was heated at the bottom at 140 degree centigrade and hydrated on top with deionized water. At the end of the test a sharp water content gradient was observed along the column, as well as an inverse dry density gradient. Hydration modified also the bentonite microstructure. Besides, an overall decrease of the smectite content with respect to the initial value took place, especially in the most hydrated areas where the percentage of interest ratified illite increased and in the longer test. On the other hand, the content of cristobalite, feldspars and calcite increased. Smectite dissolution processes (probably colloidal) occurred, particularly in the more hydrated areas and in the longer test. Due to the dissolution of low-solubility species and to the loss of exchangeable positions in the smectite, the content of soluble salts in the pore water increased with respect to the original one, especially in the longer test. The solubilized ions were transported; sodium, calcium, magnesium and sulphate having a similar mobility, which was in turn lower than that of potassium and chloride. The cationic exchange complex was also modified. (Author)

  6. HOT 2017

    DEFF Research Database (Denmark)

    Hannibal, Sara Stefansen

    HOT er en kvalitativ undersøgelse, der hvert år diskuterer og undersøger en lille udvalgt skare af danskkyndige fagpersoners bud på, hvad de er optagede af på literacyområdet her og nu – altså hvilke emner, de vil vurdere som aktuelle at forholde sig til i deres nuværende praksis.......HOT er en kvalitativ undersøgelse, der hvert år diskuterer og undersøger en lille udvalgt skare af danskkyndige fagpersoners bud på, hvad de er optagede af på literacyområdet her og nu – altså hvilke emner, de vil vurdere som aktuelle at forholde sig til i deres nuværende praksis....

  7. Design, construction and evaluation of solar flat-plate collector simulator based on the thermohydraulic coefficient

    Directory of Open Access Journals (Sweden)

    H Rahmati Aidinlou

    2017-05-01

    Full Text Available Introduction Increasing the area of absorber plate between the flowed air through the duct can be accomplished by corrugating the absorber plate or by using the artificial roughness underside of the absorber plate as the commercial methods for enhancing the thermohydraulic performance of the flat plate solar air heaters. Evaluation of this requires the construction of separated solar air heater which is costly and time consuming. The constructed solar flat-plate collector simulator can be a sufficient solution for obtaining the heat transfer and thermodynamic parameters for evaluating the absorber plate. The inclined broken roughness was chosen as the optimum roughness which is surrounded by three aluminum smooth walls. Materials and Methods The duct for both smooth and roughened plate have been constructed based on the ASHRAE 93-2010 standard. In order to achieve a fully thermal and hydraulic developed flow, the plenum is constructed. The centrifugal fan is considered by applying the required air volume at the pressure drop obtained by the duct, plenum and the orifice meter. The TSI velocity-meter 8355 is used to measure the velocity of air crossing through the pipe connected to the centrifugal fan. The micro manometer Kimo CPE310-s with the resolution of 0.1 Pa is used to measure the pressure drop across the test section of the smooth and roughened duct. The LM35 sensors are used to measure the absorber plate and air temperature through the test section. Obtained parameters are used to calculate the Nusselt number and friction factor across the test section for smooth and roughened absorber plate. The Nusselt number and friction factor parameters which is obtained for smooth absorber plate based on experimental set-up, is compared with Dittus-Bolter and Blasius equations, respectively, for validating the simulator. By calculating the Nusselt number and friction factor, Stanton number is obtained based on the equation (6, and thermohydraulic

  8. Hot particles

    International Nuclear Information System (INIS)

    Merwin, S.E.; Moeller, M.P.

    1989-01-01

    Nuclear Regulatory Commission (NRC) licensees are required to assess the dose to skin from a hot particle contamination event at a depth of skin of7mg/cm 2 over an area of 1 cm 2 and compare the value to the current dose limit for the skin. Although the resulting number is interesting from a comparative standpoint and can be used to predict local skin reactions, comparison of the number to existing limits based on uniform exposures is inappropriate. Most incidents that can be classified as overexposures based on this interpretation of dose actually have no effect on the health of the worker. As a result, resources are expended to reduce the likelihood that an overexposure event will occur when they could be directed toward eliminating the cause of the problem or enhancing existing programs such as contamination control. Furthermore, from a risk standpoint, this practice is not ALARA because some workers receive whole body doses in order to minimize the occurrence of hot particle skin contaminations. In this paper the authors suggest an alternative approach to controlling hot particle exposures

  9. Analysis of steam generator plugging on core thermohydraulic performance of NPP Krsko; Analiza vpliva cepljenja cevi v upravljaniku na termohidravliko sredice JE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Kostadinov, V; Petelin, S; Sarler, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1988-07-01

    Nuclear safety analysis of NPP Krsko core operating at full power with 4% steam generator tubes plugged have been performed. Influence of individual parameters on core thermohydraulic performance have been evaluated. Using COBRA-III-C computer code we have analysed a core design (evaluation) model. The DNBR change was calculated as a consequence of 4% plugging. The influence of thermohydraulic parameters change on DNBR was analysed. (author)

  10. Pool water cleaning facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro [Hitachi Ltd., Tokyo (Japan); Asano, Takashi

    1998-05-29

    Only one system comprising a suppression poor water cleaning system (SPCU) and a filtration desalting tower (F/D) is connected for a plurality of nuclear power plants. Pipelines/valves for connecting the one system of the SPCU pump, the F/D and the plurality of nuclear power plants are disposed, and the system is used in common with the plurality of nuclear power plants. Pipelines/valves for connecting a pipeline for passing SP water to the commonly used SPCU pump and a skimmer surge tank are disposed, and fuel pool water is cooled and cleaned by the commonly used SPCU pump and the commonly used F/D. The number of SPCU pumps and the F/D facilities can be reduced, and a fuel pool water cooling operation mode and a fuel pool water cleaning operation mode which were conducted by an FPC pump so far are conducted by the SPCU pump. (N.H.)

  11. Simulation of pool scrubbing experiments using BUSCA

    International Nuclear Information System (INIS)

    Dehbi, A.; Guentay, S.

    1994-01-01

    BUSCA-PSI is a computer code which predicts the aerosol scrubbing taking place when gas bubbles containing fission products rise through stagnant pools of water after a postulated severe accident. A Lagrangian formulation is adopted to follow the path of a bubble as it rises toward the surface of the pool. The BUSCA model includes most aerosol removal mechanisms which are thought to be significant, namely: Jet Impaction at tile orifice, Convection/Diffusiophoresis during steam condensation, Thermophoresis, Sedimentation, Centrifugal Impaction during bubble rise, and Brownian Diffusion. The hydraulic modelling offers a variety of options for the initial globule volume, the stable bubble size, tile bubble rise velocity, and the bubble shape. The heat and mass transfer part of tile model uses correlations found in the relevant literature. BUSCA simulations were performed to determine the decontamination factor (DF) dependence on key aerosol and thermal hydraulic parameters. The decontamination factor increases with height, pool temperature subcooling, and steam content. The decontamination factor exhibits a parabolic dependence on the particle radius. At low particle sizes, the DF is high due to Brownian Diffusion which is the dominant removal mechanism. The DF hits a minimum and then increases with particle size as Centrifugal Impaction and Sedimentation become important. In separate calculations, BUSCA was used to the simulate the aerosol scrubbing experiments performed by EPRI. For cold pool tests, the predicted scrubbing efficiencies were in a good, conservative agreement with the data for both Tin and CsI, and the discrepancies were within the reported measurement errors. For hot pool tests, the code systematically underpredicted the scrubbing DF's; this is potentially due to condensation in the gas space above the pool, a situation not currently modelled by BUSCA. The code was also tested against data produced by the Tepco-Toshiba-Hitachi experiments. The

  12. Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC

    International Nuclear Information System (INIS)

    Gomez Garcia-Torano, I.; Jimenez, G.

    2013-01-01

    The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

  13. Liquid sodium pool fires

    Energy Technology Data Exchange (ETDEWEB)

    Casselman, C [DSN/SESTR, Centre de Cadarache, Saint-Paul-lez-Durance (France)

    1979-03-01

    Experimental sodium pool combustion results have led to a definition of the combustion kinetics, and have revealed the hazards of sodium-concrete contact reactions and the possible ignition of organic matter (paint) by hydration of sodium peroxide aerosols. Analysis of these test results shows that the controlling mechanism is sodium evaporation diffusion. (author)

  14. Income pooling within families

    DEFF Research Database (Denmark)

    Bonke, Jens; Uldall-Poulsen, Hans

    This paper analyses the phenomenon of income-pooling by applying the Danish household expenditure survey, merged with authoritative register information. Responses to additional questions on income sharing among 1696 couples also allows us to analyses whether the intra-household distribution...

  15. Liquid sodium pool fires

    International Nuclear Information System (INIS)

    Casselman, C.

    1979-01-01

    Experimental sodium pool combustion results have led to a definition of the combustion kinetics, and have revealed the hazards of sodium-concrete contact reactions and the possible ignition of organic matter (paint) by hydration of sodium peroxide aerosols. Analysis of these test results shows that the controlling mechanism is sodium evaporation diffusion. (author)

  16. Thermohydraulic Design Analysis Modeling for Korea Advanced NUclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    NTR engines have continued as a main stream based on the mature technology. The typical core design of the NERVA derived engines uses hexagonal shaped fuel elements with circular cooling channels and structural tie-tube elements for supporting the fuel elements, housing moderator and regeneratively cooling the moderator. The state-of-the-art NTR designs mostly use a fast or epithermal neutron spectrum core utilizing a HEU fuel to make a high power reactor with small and simple core geometry. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The KANUTER-HEU and -LEU are the innovative and futuristic NTR engines to reduce the reactor size and to implement a LEU fuel in the reactor by using thermal neutron spectrum. The KANUTERs have some features in the reactor design such as the integrated fuel element and the regeneratively cooling channels to increase room for moderator and heat transfer in the core, and ensuing rocket performance. To study feasible design points in terms of thermo-hydraulics and to estimate rocket performance of the KANUTERs, the NSES is under development. The model of the NSES currently focuses on thermo-hydraulic analysis of the peculiar and complex EHTGR design during the propulsion mode in steady-state. The results indicate comparable performance for future applications, even though it uses the heavier LEU fuel. In future, the NSES will be modified to obtain temperature distribution of the entire reactor components and then more extensive design analysis of neutronics, thermohydraulics and their coupling will be conducted to validate design feasibility and to optimize the reactor design enhancing the rocket performance.

  17. Response of Compacted Bentonites to Thermal and Thermo-Hydraulic Loadings at High Temperatures

    Directory of Open Access Journals (Sweden)

    Snehasis Tripathy

    2017-07-01

    Full Text Available The final disposal of high-level nuclear waste in many countries is preferred to be in deep geological repositories. Compacted bentonites are proposed for use as the buffer surrounding the waste canisters which may be subjected to both thermal and hydraulic loadings. A significant increase in the temperature is anticipated within the buffer, particularly during the early phase of the repository lifetime. In this study, several non-isothermal and non-isothermal hydraulic tests were carried on compacted MX80 bentonite. Compacted bentonite specimens (water content = 15.2%, dry density = 1.65 Mg/m3 were subjected to a temperature of either 85 or 150 °C at one end, whereas the temperature at the opposite end was maintained at 25 °C. During the non-isothermal hydraulic tests, water was supplied from the opposite end of the heat source. The temperature and relative humidity were monitored along predetermined depths of the specimens. The profiles of water content, dry density, and degree of saturation were established after termination of the tests. The test results showed that thermal gradients caused redistribution of the water content, whereas thermo-hydraulic gradients caused both redistribution and an increase in the water content within compacted bentonites, both leading to development of axial stress of various magnitudes. The applied water injection pressures (5 and 600 kPa and temperature gradients appeared to have very minimal impact on the magnitude of axial stress developed. The thickness of thermal insulation layer surrounding the testing devices was found to influence the temperature and relative humidity profiles thereby impacting the redistribution of water content within compacted bentonites. Under the influence of both the applied thermal and thermo-hydraulic gradients, the dry density of the bentonite specimens increased near the heat source, whereas it decreased at the opposite end. The test results emphasized the influence of

  18. Computation of 3D thermohydraulics in partially blocked bundles during the reflood phase of a LOCA

    International Nuclear Information System (INIS)

    Cicero, G.M.; Briere, E.; Fornaciari, G.

    1994-06-01

    In Pressurized Water Reactors (PWR), ballooning of the fuel rod claddings may occur during a LOCA, since the fuel rod claddings are heated up, and the system pressure is low. The severe blockages that may result induce cross-flow diversion and three-dimensional effects on thermohydraulics in the core bundle, during the reflood phase. To improve the knowledge of these phenomena and their physical modelling in the code CATHARE, 3D computer codes are needed. In 1990, EDF has started up a development and validation program of the 3D THYC computer code to analyze the thermohydraulics of the flow during the reflood phase, in partially blocked bundles. The main objective is to calculate the temperatures of the rods above the quench front, when they are cooled by superheated steam with saturated droplets. First, this paper introduces the THYC model developed for reflood studies. Secondly, we report the first qualification results on a Flooding Experiments with Blocked Array (FEBA) test. Thirdly, we analyze the model predictions on a large break LOCA transient, in a 900 MW PWR 11x11 core area with a 3x3 central blockage. THYC simulates the transient in the bundle around and above the blockage, until the quench front enters the computational domain. Previously, a 1D CATHARE simulation gives the boundary conditions and, in the reactor core case, the deformation of the blocked fuel rods. The results analysis focused on the time evolution of the clad temperatures in the blocked and in the bypass region. In the FEBA test simulation, the main observations are properly predicted within the blockage. Temperatures are lower in blocked rod sleeves than in unblocked rod claddings since the steam gap reduces the power transmitted by the heater rod to the sleeve. In the core case, the model predicts the opposite result. Within the blockage, ballooned rod temperatures are higher than non-ballooned rod ones. We show by sensitivity studies that these behaviour difference between FEBA rods

  19. Thermal hydraulic study of a corium molten pool

    International Nuclear Information System (INIS)

    Pigny, S.; Grand, D.; Seiler, J.M.; Durin, M.

    1993-01-01

    The thermohydraulic behaviour of a mass of molten core is investigated, in the frame of PWR severe accidents studies. The corium may be located in the vessel lower head or in an external core-catcher. It is assumed to be present in the container instantaneously. Its motion is described by one velocity field. It may be homogeneous or made of two stratified fluids. The residual power is assumed to be constant and uniform in the UO 2 phase. The radiative losses and the external water-cooling are taken into account. The thermal resistance of a peripheral crust is considered. The influence of the crust on the pool geometry may be studied. The wall behaviour is analysed by a conduction calculation. The interest of a sacrificial layer is underlined, so as the necessity of a multicomponent multiphase model to study the behaviour of a core catcher. It is also concluded that some experiments are needed for code validation about volume heated natural convection and multiphase flows. (author). 14 figs., 3 refs

  20. Archaeal Nitrification in Hot Springs

    Science.gov (United States)

    Richter, A.; Daims, H.; Reigstad, L.; Wanek, W.; Wagner, M.; Schleper, C.

    2006-12-01

    Biological nitrification, i.e. the aerobic conversion of ammonia to nitrate via nitrite, is a major component of the global nitrogen cycle. Until recently, it was thought that the ability to aerobically oxidize ammonia was confined to bacteria of the phylum Proteobacteria. However, it has recently been shown that Archaea of the phylum Crenarchaeota are also capable of ammonia oxidation. As many Crenarchaeota are thermophilic or hyperthermophilic, and at least some of them are capable of ammonia oxidation we speculated on the existence of (hyper)thermophilic ammonia-oxidizing archaea (AOA). Using PCR primers specifically targeting the archaeal ammonia monooxygenase (amoA) gene, we were indeed able to confirm the presence of such organisms in several hot springs in Reykjadalur, Iceland. These hot springs exhibited temperatures well above 80 °C and pH values ranging from 2.0 to 4.5. To proof that nitrification actually took place under these extreme conditions, we measured gross nitrification rates by the isotope pool dilution method; we added 15N-labelled nitrate to the mud and followed the dilution of the label by nitrate production from ammonium either in situ (incubation in the hot spring) or under controlled conditions in the laboratory (at 80 °C). The nitrification rates in the hot springs ranged from 0.79 to 2.22 mg nitrate-N per L of mud and day. Controls, in which microorganisms were killed before the incubations, demonstrated that the nitrification was of biological origin. Addition of ammonium increased the gross nitrification rate approximately 3-fold, indicating that the nitrification was ammonium limited under the conditions used. Collectively, our study provides evidence that (1) AOA are present in hot springs and (2) that they are actively nitrifying. These findings have major implications for our understanding of nitrogen cycling of hot environments.

  1. Generalized multidemensional propagation velocity equations for pool-boiling superconducting windings

    International Nuclear Information System (INIS)

    Christensen, E.H.; O'Loughlin, J.M.

    1984-09-01

    Several finite difference, finite element detailed analyses of propagation velocities in up to three dimensions in pool-boiling windings have been conducted for different electromagnetic and cryogenic environments. Likewise, a few full scale simulated winding and magnet tests have measured propagation velocities. These velocity data have been correlated in terms of winding thermophysical parameters. This analysis expresses longitudinal and transverse propagation velocities in the form of power function regression equations for a wide variety of windings and electromagnetic and thermohydraulic environments. The generalized velocity equations are considered applicable to well-ventilated, monolithic conductor windings. These design equations are used piecewise in a gross finite difference mode as functions of field to predict the rate of normal zone growth during quench conditions. A further check of the validity of these predictions is available through total predicted quench durations correlated with actual quench durations of large magnets

  2. Pool gateway seal

    International Nuclear Information System (INIS)

    Starr, J.A.; Steinert, L.A.

    1983-01-01

    A device for sealing a gateway between interconnectable pools in a nuclear facility comprising a frame supporting a liquid impermeable sheet positioned in a u-shaped gateway between the pools. An inflatable tube carried in a channel in the periphery of the frame and adjoining the gateway provides a seal therebetween when inflated. A restraining arrangement on the bottom edge of the frame is releasably engagable with an adjacent portion of the gateway to restrict the movement of the frame in the u-shaped gateway upon inflation of the tube, thereby enhancing the seal. The impermeable sheet is formed of an elastomer and thus is conformable to a liquid permeable supportive wall upon application of liquid pressure to the side of the sheet opposite the wall

  3. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

    Roebert, G.A.

    1978-01-01

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  4. IBIS, FBR 3-D Steady-State and Kinetics with Thermohydraulic Feedback

    International Nuclear Information System (INIS)

    Konomura, Mamoru; Tada, Nobuo; Oka, Yoshiaki; An, Shigehiro

    1987-01-01

    1 - Description of program or function: The IBIS code performs steady state and kinetics calculations based on a three-dimensional nuclear diffusion kinetics with thermal hydraulic feedback. It can calculate the following values in hexagonal-Z geometry of a fast breeder reactor core through the progress of transient: (1) Net reactivity; (2) Total and group-wise delayed neutron fraction; (3) Group-wise delayed neutron precursor concentration; (4) Total power and energy; (5) Space dependent neutron flux in each energy group; (6) Space dependent temperature of each material; (7) Maximum temperature of each material and its location. 2 - Method of solution: The quasi-static method is adopted to solve the three-dimensional nuclear diffusion kinetics problem. The method is the same as employed in the code QX1. The shape function equation is solved with the finite difference treatment as used in the codes CITATION and HONEYCOMB. One-dimensional thermo-hydraulics is solved with a model similar to that given in the code SASLA. Sodium boiling can be taken into account. 3 - Restrictions on the complexity of the problem: The number of neutron energy groups is fixed to 3 groups in the present version of the code

  5. Thermo-hydraulic consequence of pressure suppression containment vessel during blowdown, 2

    International Nuclear Information System (INIS)

    Aya, Izuo; Nariai, Hideki; Kobayashi, Michiyuki

    1980-01-01

    As a part of the safety research works for the integral-type marine reactor, an analytical code SUPPAC-2V was developed to simulate the thermo-hydraulic consequence of a pressure suppression containment system during blowdown and the code was applied to the Model Experimental Facility of the Safety of Integral Type Marine Reactors (explained already in Part 1). SUPPAC-2V is much different from existing codes in the following points. A nonhomogeneous model for the gaseous region in the drywell, a new correlation for condensing heat transfer coefficient at drywell wall based on existing data and approximation of air bubbles in wetwell water by one dimensional bubble rising model are adopted in this code. In comparing calculational results with experimental results, values of predominant input parameters were evaluated and discussed. Moreover, the new code was applied also to the NSR-7 marine reactor, conceptually designed at the Shipbuilding Research Association in Japan, of which suppression system had been already analysed by CONTEMPT-PS. (author)

  6. Thermo-Hydraulic behaviour of dual-channel superconducting Cable-In-Conduit Conductors for ITER

    International Nuclear Information System (INIS)

    Renard, B.

    2006-09-01

    In an effort to optimise the cryogenics of large superconducting coils for fusion applications (ITER), dual channel Cable-In-Conduit Conductors (CICC) are designed with a central channel spiral to provide low hydraulic resistance and faster helium circulation. The qualitative and economic rationale of the conductor central channel is here justified to limit the superconductor temperature increase, but brings more complexity to the conductor cooling characteristics. The pressure drop of spirals is experimentally evaluated in nitrogen and water and an explicit hydraulic friction model is proposed. Temperatures in the cable must be quantified to guarantee superconductor margin during coil operation under heat disturbance and set adequate inlet temperature. Analytical one-dimensional thermal models, in steady state and in transient, allow to better understand the thermal coupling of CICC central and annular channels. The measurement of a heat transfer characteristic space and time constants provides cross-checking experimental estimations of the internal thermal homogenization. A simple explicit model of global inter-channel heat exchange coefficient is proposed. The risk of thermosyphon between the two channels is considered since vertical portions of fusion coils are subject to gravity. The new hydraulic model, heat exchange model and gravitational risk ratio allow the thermohydraulic improvement of CICC central spirals. (author)

  7. Inductive analysis of failure patterns and of their impact on thermohydraulic circuits of nuclear power plants

    International Nuclear Information System (INIS)

    Limnios, N.

    1983-01-01

    The APACHE code (Automatic Analysis of Failures of Hydraulic and Thermohydraulic Circuits more particularly of Water) situates in an important program of computer codes development in the field of studies on reliability and safety of systems in nuclear power plants. APACHE is an automatic generation code of failure pattern and of their effects. After a presentation of the theoretical basis, the methodological principles of the theory of networks are developed. Then, the model of the code is developed: model of individual behavior of each classical model component of normal behavior and model of failure pattern with specifications. The global model of hydraulic systems and the resolution systems are then developed. More particularly, some aspects of the theory of graphs, and the algorithms developed for the automatic construction of the equation systems and especially the algorithm of the research of meshes are presented. The computer aspect of the code and the programming of the code with its limits and some specifications are described. The practical aspect of utilization is finally presented [fr

  8. Development of gas-cooled fast reactor and its thermo-hydraulics

    International Nuclear Information System (INIS)

    Kawamura, Hiroshi

    1977-10-01

    Development, thermo-hydraulics and safety of GCFR are reviewed. The Development of Gas-Cooled Fast Reactor (GCFR) utilizes helium technology of HTGR and fuel technology of LMFBR. The breeding ratio of GCFR will be larger than that of LMFBR by about 0.2. Features of GCFR are a fuel with roughened surface to raise the heat transfer and vent system for the pressure equalization in the fuel rod. Helium as coolant of GCFR is chemically stable and stays in the single phase. So, there is no fuel-coolant interaction unlike the case of LMFBR. Since the helium must be pressurized, possibility of a depressurization accident is not negligible. In the United States, a 300MWe demonstration plant program is about to start; the collaboration with European countries is now quite active in this field. Though the development of GCFR started behind that of LMFBR, GCFR is equally promising as a fast breeder reactor. When realized, it will present possibility of a choice between these two. (auth.)

  9. The development of the thermohydraulic analysis code for the passive containment cooling system: PCCSAC

    International Nuclear Information System (INIS)

    Wang Jianyu; Zhang Shenru; Min Yuanyou

    1994-01-01

    To estimate the performance of the passive containment cooling system (PCCS) of the AC-600 nuclear power plant, the PCCSAC code is developed currently by the jointed efforts between Tsinghua University and NPIC. Different features on the passive behavior of the system and the main components of the containment are considered in the code which is needed by the further AC-600 R and D Program. With a brief description of the AC-600 passive containment cooling system and components, the main thermohydraulic models and numerical scheme used in the PCCSAC code are introduced and the selected results of the verification and the prediction for the performance of the AC-600 passive containment cooling system under LOCA and a steam line break accident are presented to preliminarily demonstrate the applicability and reliability of the PCCSAC model. The current PCCSAC model is conservative and a further 2-D PCCSAC version is under consideration in addition to provide the database for models from some tests associated with the components and systems unique to AC-600 nuclear power plant to meet the requirement of the more realistic modelization for the AC-600 passive containment cooling system. (author)

  10. A parametric thermohydraulic study an advanced pressurized light water reactor with a tight fuel rod lattice

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Hame, W.

    1982-12-01

    A parametric thermohydraulic study for an Advanced Pressurized Light Water Reactor (APWR) with a tight fuel rod lattice has been performed. The APWR improves the uranium utilisation. The APWR core should be placed in a modern German PWR plant. Within this study about 200 different reactors have been calculated. The tightening of the fuel rod lattice implies a decrease of the net electrical output of the plant, which is greater for the heterogeneous reactor than for the homogeneous reactor. APWR cores mean higher core pressure drops and higher water velocities in the core region. The cores tend to be shorter and the number of fuel rods to be higher than for the PWR. At the higher fuel rod pitch to diameter ratios (p/d) the DNB limitation is more stringent than the limitation on the fuel rod linear rating given by the necessity of reflooding after a reactor accident. The contrary is true for the lower p/d ratios. Subcooled boiling in the highest rated coolant channels occurs for the most of the calculated reactors. (orig.) [de

  11. Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies

    International Nuclear Information System (INIS)

    Sorokin, G.A.; Avdeev, E.F.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P.

    2000-01-01

    An investigation into the reactor core accident cooling, which are associated with the power grow up or switch off circulation pumps in the event of the protective equipment comes into action, results in the problem of liquid metal boiling heat transfer. Considerable study has been given over the last 30 years to alkaline metal boiling including researches of heat transfer, boiling patterns, hydraulic resistance, crisis of heat transfer, initial heating up, boiling onset and instability of boiling. The results of these investigations have shown that the process of liquid metal boiling has substantial features in comparison with water boiling. Mathematical modeling of two phase flows in fast reactor fuel subassemblies have been developed intensively. Significant success has been achieved in formulation of two phase flow through the pin bundle and in their numerical realization. Currently a set of codes for thermohydraulic analysis of two phase flows in fast reactor subassembly have been developed with 3D macrotransfer governing equations. These codes are used for analysis of boiling onset and liquid metals boiling in fuel subassemblies during loss-of-coolant accidents, of warming up of reactor core, of blockage of some part of flow cross section in fuel subassembly. (author)

  12. Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Moreira, Uebert G.; Dominguez, Dany S. [Universidade Estadual de Santa Cruz (UESC), Ilh´eus, BA (Brazil). Programa de P´os-Graduacao em Modelagem Computacional em Ciencia e Tecnologia; Mazaira, Leorlen Y.R.; Lira, Carlos A.B.O. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Hernandez, Carlos R.G., E-mail: uebert.gmoreira@gmail.com, E-mail: dsdominguez@gmail.com, E-mail: leored1984@gmail.com, E-mail: cabol@ufpe.br, E-mail: cgh@instec.cu [Instituto Superior de Tecnologas y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

  13. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Dominguez, Dany S.; Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G.; Lira, Carlos Alberto Brayner de Oliveira

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  14. Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

    International Nuclear Information System (INIS)

    Moreira, Uebert G.; Dominguez, Dany S.

    2017-01-01

    The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

  15. General and preliminary thermohydraulic, hydrogen and aerosol instrumentation plan for the Phebus Fp-project

    International Nuclear Information System (INIS)

    Hampel, G.; Poss, G.; Frohlich, H.K.

    1989-10-01

    The objective of the project was to draw up an instrumentation plan for the French core melting programme PHEBUS FP. This instrumentation plan essentially was to include proven and reliable instruments for recording various thermohydraulic, aerosol and hydrogen phenomena. The candidate measuring methods, which are known mainly from reactor safety programmes, have been described and examined for their usefulness in PHEBUS. Each method and instrument has been described in detail under various aspects such as measuring principle, measuring range, technical design, evaluation model, calibration procedure, accuracy, previous experience, commercial availability, etc. Special attention has been paid to the behaviour of the measuring transducers when exposed to radiation. First, the performance of the instruments was compared with the requirements of PHEBUS. The results of this comparison served as the basis for a measuring concept in tabular form, giving the locations of the measurements, the measuring tasks, and the number and kind of instruments that are recommended. Redundancy and cost-benefit aspects have been taken into account in qualitative terms

  16. Thermo-hydraulic analysis of the cool-down of the EDIPO test facility

    Science.gov (United States)

    Lewandowska, Monika; Bagnasco, Maurizio

    2011-09-01

    The first cool-down of the EDIPO (European DIPOle) test facility is foreseen to take place in 2011 by means of the existing 1.2 kW cryoplant at EPFL-CRPP Villigen. In this work, the thermo-hydraulic analysis of the EDIPO cool-down is performed in order both to assess the its duration and to optimize the procedure. The cool-down is driven by the helium flowing in both the outer cooling channel and in the windings connected hydraulically in parallel. We take into account limitations due to the pressure drop in the cooling circuit and the refrigerator capacity as well as heat conduction in the iron yoke. Two schemes of the hydraulic cooling circuit in the EDIPO windings are studied (coils connected in series and coils connected in parallel). The analysis is performed by means of an analytical model complemented by and numerical model. The results indicate that the cool-down to 5 K can be achieved in about 12 days.

  17. Fuel assembly storage pool

    International Nuclear Information System (INIS)

    Hiranuma, Hiroshi.

    1976-01-01

    Object: To remove limitation of the number of storage of fuel assemblies to increase the number of storage thereof so as to relatively reduce the water depth required for shielding radioactive rays. Structure: Fuel assembly storage rack containers for receiving a plurality of spent fuel assembly racks are stacked in multi-layer fashion within a storage pool filled with water for shielding radioactive rays and removing heat. (Furukawa, Y.)

  18. CERN Electronics Pool presentations

    CERN Multimedia

    2011-01-01

    The CERN Electronics Pool has organised a series of presentations in collaboration with oscilloscope manufacturers. The last one will take place according to the schedule below.   Time will be available at the end of the presentation to discuss your personal needs. The Agilent presentation had to be postponed and will be organised later. -     Lecroy: Thursday, 24 November 2011, in 530-R-030, 14:00 to 16:30.

  19. Investigation to radioactive contamination of pool water in IMEF

    International Nuclear Information System (INIS)

    Song, Ung Sup; Jung, Yang Hong; Lee, J. H.; Lee, H. K.

    2003-06-01

    The pool (3x6x10) in irradiated materials examination facility is usually used for the purpose of taking the specimen out of cask loaded into the pool, and carrying in/out the specimen to/ from the hot cell. Always, it must be cared for the water into the pool to be fine condition because all operation are worked with the naked eye during taking an irradiated materials out of the cask and plunging them in the bucket-elevator. In the aspects of the radioactive remained substances in the water must be controlled so that the amount of substances to be lower than the standard amount prescribed by RCA Korea Activity in a part of radioactive contamination control. In consequence, an expertness of status and a practical use of skill make possible the prevention of radioactive material's diffusion or the radioactive contamination of pool water and safety work

  20. Swimming Pools and Molluscum Contagiosum

    Science.gov (United States)

    ... Travelers’ Health: Smallpox & Other Orthopoxvirus-Associated Infections Poxvirus Swimming Pools Recommend on Facebook Tweet Share Compartir The ... often ask if molluscum virus can spread in swimming pools. There is also concern that it can ...

  1. Pool-type reactor

    International Nuclear Information System (INIS)

    Hopkins, S.R.

    1977-01-01

    This invention relates to a pool nuclear reactor fitted with a perfected system to raise the buckets into a vertical position at the bottom of a channel. This reactor has an inclined channel to guide a bucket containing a fuel assembly to introduce it into the reactor jacket or extract it therefrom and a damper at the bottom of the channel to stop the drop of the bucket. An upright vertically movable rod has a horizontally articulated arm with a hook. This can pivot to touch a radial lug on the bucket and pivot the bucket around its base in a vertical position, when the rod moves up [fr

  2. Thermo-hydraulic characterization of a self-pumping corrugated wall heat exchanger

    International Nuclear Information System (INIS)

    Schmidmayer, Kevin; Kumar, Prashant; Lavieille, Pascal; Miscevic, Marc; Topin, Frédéric

    2017-01-01

    Compactness, efficiency and thermal control of the heat exchanger are of critical significance for many electronic industry applications. In this view, a new concept of heat exchanger at millimeter scale is proposed and numerically studied. It consists in dynamically deforming at least one of its walls by a progressive wave in order to create an active corrugated channel. Systematic studies were performed in single-phase flow on the different deformation parameters that allow obtaining the thermo-hydraulic characteristics of the system. It has been observed the dynamic wall deformation induces a significant pumping effect. Intensification of heat transfer remains very important even for highly degraded waveforms although the pumping efficiency is reduced in this case. The mechanical power applied on the upper wall to deform it dynamically is linked to the wave shape, amplitude, frequency and outlet-inlet pressure difference. The overall performance of the proposed system has been evaluated and compared to existing static channels. The performance of the proposed heat exchanger evolved in two steps for a given wall deformation. It declines slightly up to a critical value of mechanical power applied on the wall. When this critical value is exceeded, it deteriorates significantly, reaching the performance of existing conventional systems. - Highlights: • A new concept of heat exchanger within channel at millimeter scale is proposed. • Upper wall is deformed dynamically by applying external mechanical power. • Pumping effect is observed and is linked to the wave shape, amplitude and frequency. • Efficient proposed system in low Reynolds number range. • Overall performance is significantly high compared to static corrugated and straight channels.

  3. Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

    International Nuclear Information System (INIS)

    Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

    2003-01-01

    During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- ε turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential

  4. The Leuze mineral water swimming pool - purposefully optimized energy utilization. Mineralbad Leuze: Sinnvoll optimierte Energienutzung

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1989-04-01

    The mineral-water swimming pool in Stuttgart-Bad Cannstatt is fed by mineral springs. The author reports on the design and energy technology used in this indoor swimming pool (photographs), pool hall (feed and exhaust air), treatment basin, showers, locker rooms (air throughput rate, feed and exhaust air management), cafeteria, kitchen, gymnastics and technical services rooms, toilets, chemicals storage room, cooling system and heat pump (heat recovery from drained pool water up to 50%). District heating steam (18 bar) is used for heat supply (reducing station). The author comments on the temperature levels required for different heating cycles (hot-pool hall, hot-water basin, skylight heating, space heating) and on thermal output requirements (kW). (HWJ).

  5. Flow and heat transfer thermohydraulic modelisation during the reflooding phase of a P.W.R.'s core

    International Nuclear Information System (INIS)

    Raymond, Patrick

    1978-04-01

    Some generalities about L.O.C.A. are first recalled. The French experimental studies about Emergency Core Cooling System are briefly described. The different heat transfer mechanisms to take into account, according to the flow pattern in the dry zone, and the correlations or methods to calculate them, are defined. Then the Thermohydraulic code computer: FLIRA, which describe the reflooding phase, and a modelisation taking into account the different flow patterns are setted. A first interpretation of ERSEC experiments with a tubular test section shows that it is possible, with this modelisation and some classical heat transfer correlations, to describe the reflooding phase. [fr

  6. 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)

  7. Thermohydraulic characteristics analysis of natural convective cooling mode on the steady state condition of upgraded JRR-3 core, using COOLOD-N code

    International Nuclear Information System (INIS)

    Kaminaga, Masanori; Watanabe, Shukichi; Ando, Hiroei; Sudo, Yukio; Ikawa, Hiromasa.

    1987-03-01

    This report describes the results of the steady state thermohydraulic analysis of upgraded JRR-3 core under natural convective cooling mode, using COOLOD-N code. In the code, function to calculate flow-rate under natural convective cooling mode, and a heat transfer package have been newly added to the COOLOD code which has been developed in JAERI. And this report describes outline of the COOLOD-N code. The results of analysis show that the thermohydraulics of upgraded JRR-3 core, under natural convective cooling mode have enough margine to ONB temperature, DNB heat flux and occurance of blisters in fuel meats, which are design criterion of upgraded JRR-3. (author)

  8. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

    2004-01-01

    Requirements of neutron, thermohydraulic and safety analysis calculation are very important because of issuing new version of SAR for DNRR, research on construction of new research reactor and nuclear power plant. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor has been done in the frame work of research theme in the year 2002-2003. The purposes of the research are maintaining safety operation of the DNRR and enhancement of man power and calculation and safety analysis tool potential. (author)

  9. Mathematical modelling and simulation of the thermal performance of a solar heated indoor swimming pool

    Directory of Open Access Journals (Sweden)

    Mančić Marko V.

    2014-01-01

    Full Text Available Buildings with indoor swimming pools have a large energy footprint. The source of major energy loss is the swimming pool hall where air humidity is increased by evaporation from the pool water surface. This increases energy consumption for heating and ventilation of the pool hall, fresh water supply loss and heat demand for pool water heating. In this paper, a mathematical model of the swimming pool was made to assess energy demands of an indoor swimming pool building. The mathematical model of the swimming pool is used with the created multi-zone building model in TRNSYS software to determine pool hall energy demand and pool losses. Energy loss for pool water and pool hall heating and ventilation are analyzed for different target pool water and air temperatures. The simulation showed that pool water heating accounts for around 22%, whereas heating and ventilation of the pool hall for around 60% of the total pool hall heat demand. With a change of preset controller air and water temperatures in simulations, evaporation loss was in the range 46-54% of the total pool losses. A solar thermal sanitary hot water system was modelled and simulated to analyze it's potential for energy savings of the presented demand side model. The simulation showed that up to 87% of water heating demands could be met by the solar thermal system, while avoiding stagnation. [Projekat Ministarstva nauke Republike Srbije, br. III 42006: Research and development of energy and environmentally highly effective polygeneration systems based on using renewable energy sources

  10. Solar 'hot spots' are still hot

    Science.gov (United States)

    Bai, Taeil

    1990-01-01

    Longitude distributions of solar flares are not random but show evidence for active zones (or hot spots) where flares are concentrated. According to a previous study, two hot spots in the northern hemisphere, which rotate with a synodic period of about 26.72 days, produced the majority of major flares, during solar cycles 20 and 21. The more prominent of these two hot spots is found to be still active during the rising part of cycle 22, producing the majority of northern hemisphere major flares. The synodic rotation period of this hot spot is 26.727 + or - 0.007 days. There is also evidence for hot spots in the southern hemisphere. Two hot spots separated by 180 deg are found to rotate with a period of 29.407 days, with one of them having persisted in the same locations during cycles 19-22 and the other, during cycles 20-22.

  11. Solar hot spots are still hot

    International Nuclear Information System (INIS)

    Bai, T.

    1990-01-01

    Longitude distributions of solar flares are not random but show evidence for active zones (or hot spots) where flares are concentrated. According to a previous study, two hot spots in the northern hemisphere, which rotate with a synodic period of about 26.72 days, produced the majority of major flares, during solar cycles 20 and 21. The more prominent of these two hot spots is found to be still active during the rising part of cycle 22, producing the majority of northern hemisphere major flares. The synodic rotation period of this hot spot is 26.727 + or - 0.007 days. There is also evidence for hot spots in the southern hemisphere. Two hot spots separated by 180 deg are found to rotate with a period of 29.407 days, with one of them having persisted in the same locations during cycles 19-22 and the other, during cycles 20-22. 14 refs

  12. Application of solar roof shallow pool at individual residental buildings

    OpenAIRE

    Gavrilović Dragan J.

    2011-01-01

    The paper discusses the possibility of applying shallow roof pools of water on the basis of passive solar water capture functioning as thermal batteries and thermal "regulators" in a "hot - cold" mode with individual residential buildings. With this application, the utilization of the existing functionality of the building roof area would improve and open up the possibility of achieving better overall bio-climate individual object. By using this system, a flat roof impassable "terrace" ...

  13. Experimental Investigation of Thermohydraulic Performance of a Rectangular Solar Air Heater Duct Equipped with V-Shaped Perforated Blocks

    Directory of Open Access Journals (Sweden)

    Tabish Alam

    2014-01-01

    Full Text Available This paper presents the thermohydraulic performance of rectangular solar air heater duct equipped with V-shaped rectangular perforated blocks attached to the heated surface. The V-shaped perforated blocks are tested for downstream (V-down to the air flow at Reynolds number from 2000 to 20000. The perforated blocks have relative pitch ratio (P/e from 4 to 12, relative blockage height ratio (e/H from 0.4 to 1.0, and open area ration from 5% to 25% at a fixed value of angle of attack of 60∘ in a rectangular duct having duct aspect ratio (W/H of 12. Thermohydraulic performance is compared at different geometrical parameters of V-shaped perforated blocks for equal pumping power which shows that maximum performance is observed at a relative pitch of 8, relative rib height of 0.8, and open area ration of 20%. It is also observed that the performance of V-shaped perforated blocks was better than transverse-perforated blocks.

  14. Observation and control system of the thermohydraulic assays laboratory; Sistema de observacion y control del laboratorio de ensayos termohidraulicos

    Energy Technology Data Exchange (ETDEWEB)

    Santome, D; Hualde, R

    1991-12-31

    The Thermohydraulic Assays Laboratory (L.E.T.) is an installation whose purpose will be the components testing and the CAREM-25 reactor thermohydraulic processes operation dynamics. This plant is located at Pilcaniyeu, province of Rio Negro. Part of the tests which will be carried out consist in the use of different control strategies. The control of the systems by digital processors (control by software) has been decided to proceed with a maximum flexibility and capacity to make changes in the algorithms. This work describes the design and implementation of a digital control system to command the three circuits of the installation. (Author). [Espanol] El Laboratorio de Ensayos Termohidraulicos (L.E.T.) es una instalacion cuyo objeto sera el ensayo de componentes y de la dinamica de operacion de los procesos termohidraulicos del reactor CAREM-25. Esta planta esta localizada en Pilcaniyeu, provincia de Rio Negro. Parte de las pruebas que se efectuaran en el L.E.T. consisten en el empleo de distintas estrategias de control. Para disponer de una maxima flexibilidad y capacidad de efectuar cambios en los algoritmos, se decidio realizar el control de los sistemas por medio de procesadores digitales (control por software). Este trabajo consistio en el diseno e implementacion de un sistema de control digital distribuido para el comando de los tres circuitos con que cuenta la instalacion. (Autor).

  15. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

    International Nuclear Information System (INIS)

    March-Leuba, J.; Rey, J.M.

    1992-01-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of 'unexpected' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a 'new and improved' state of the art has emerged recently. (authors). 6 figs., 57 refs., 1 appendix

  16. Computational fluid dynamic model for thermohydraulic calculation for the steady-state of the real scale HTR-1

    Energy Technology Data Exchange (ETDEWEB)

    Gamez, Abel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy Y.; Gonzalez, Daniel; Garcia, Carlos, E-mail: agamezgmf@gmail.com, E-mail: leored1984@gmail.com, E-mail: jrosales@instec.cu, E-mail: lcastro@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgr@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Oliveira, Carlos B. de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil)

    2015-07-01

    The high temperature gas cooled reactor (HTGR) is one of candidates of next generation of nuclear reactor according to IAEA report 2013. Evaluation of thermohydraulic performance and an experimental comparison results were proposed to the international research community. In this article, the tree dimensional CFD thermohydraulic modelation of steady state of HTR-10 modular reactor, using ANSYS CFX v14.0, has been done. Code-to-code and Code-to-experiment benchmark analyses, related to the testing program of the HTR-10 plant including steady state temperature distribution with the reactor at full power, were developed. The 3D real scale representation of reflector zone and fluid path flow inner and outer reflector blocks and cold helium cavity were carried out. The porous medium model was used to simulate the core zone in the reactor. The power distribution of the initial core published by IAEA-TECDOC-1694 obtained by Chief Scientific Investigators (CSIs) from China was used as heat sources in the core zone. (author)

  17. Microbial diversity in acidic thermal pools in the Uzon Caldera, Kamchatka.

    Science.gov (United States)

    Mardanov, Andrey V; Gumerov, Vadim M; Beletsky, Alexey V; Ravin, Nikolai V

    2018-01-01

    Microbial communities of four acidic thermal pools in the Uzon Caldera, Kamchatka, Russia, were studied using amplification and pyrosequencing of 16S rRNA gene fragments. The sites differed in temperature and pH: 1805 (60 °C, pH 3.7), 1810 (90 °C, pH 4.1), 1818 (80 °C, pH 3.5), and 1807 (86 °C, pH 5.6). Archaea of the order Sulfolobales were present among the dominant groups in all four pools. Acidilobales dominated in pool 1818 but were a minor fraction at the higher temperature in pool 1810. Uncultivated Archaea of the Hot Thaumarchaeota-related clade were present in significant quantities in pools 1805 and 1807, but they were not abundant in pools 1810 and 1818, where high temperatures were combined with low pH. Nanoarchaeota were present in all pools, but were more abundant in pools 1810 and 1818. A similar abundance pattern was observed for Halobacteriales. Thermophilic Bacteria were less diverse and were mostly represented by aerobic hydrogen- and sulfur-oxidizers of the phylum Aquificae and sulfur-oxidising Proteobacteria of the genus Acidithiobacillus. Thus we showed that extremely acidic hot pools contain diverse microbial communities comprising different metabolic groups of prokaryotes, including putative lithoautotrophs using energy sources of volcanic origin, and various facultative and obligate heterotrophs.

  18. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors

    International Nuclear Information System (INIS)

    Vega R, A. K.; Espinosa P, G.; Gomez T, A. M.

    2016-09-01

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  19. Effect of boundary conditions on thermohydraulic behavior of clay buffer used in nuclear waste repository

    International Nuclear Information System (INIS)

    Arul Peter, A.; Murugesan, K.; Mamidi, Ganesh; Sharma, Umesh Kumar; Sharma, D. Akanshu; Arora, Puneet

    2010-01-01

    barrier will be subjected to in real situations. When either the temperature or heat flux is increased at the canister side, the saturation corresponding to that side decreases because of increase in temperature and pressure gradients. The comparisons of the results obtained from the present numerical simulation with the available experimental results shows that the present model can simulate the thermo-hydraulic behaviour of unsaturated porous media according to the physics underlying the problem. (author)

  20. Neutronics and thermohydraulics of the reactor C.E.N.E. Pt. 1

    International Nuclear Information System (INIS)

    Caro, R.; Ahnert, C.; Esteban Naudin, A.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-01-01

    The analysis of neutronics (both statics and kinetics), of the 10 Mwt swimming pool reactor C.E.N.E. is included. A short description of the theoretical model used, along with the theoretical versus experimental cheking, carried out, whenever possible, with the reactors JEN-1 and JEN-2 of Junta de Energia Nuclear, is given in each of these chapters. (author) [es

  1. Study of mixed convection in sodium pool

    International Nuclear Information System (INIS)

    Wang Zhou; Chen Yan

    1995-01-01

    The mixed convection phenomena in the sodium pool of fast reactor have been studied systematically by the two dimensional modeling method. A generalized concept of circumferential line in the cylindrical coordinates was proposed to overcome the three dimensional effect induced by the pool geometry in an analysis of two dimensional modeling. A method of sub-step in time was developed for solving the turbulent equations. The treatments on the boundary condition for the auxiliary velocity field have been proposed, and the explanation of allowing the flow function method to be used in the flow field in presence of a mass source term was given. As examples of verification, the experiments were conducted with water flow in a rectangular cavity. The results from theoretical analysis were applied to the numerical computation for the mixed convection in the cavity. The mechanism of stratified flow in the cavity was studied. A numerical calculation was carried out for the mixed convection in hot plenum of a typical fast reactor

  2. Hot tub folliculitis

    Science.gov (United States)

    ... survives in hot tubs, especially tubs made of wood. Symptoms The first symptom of hot tub folliculitis ... may help prevent the problem. Images Hair follicle anatomy References D'Agata E. Pseudomonas aeruginosa and other ...

  3. Measurements in large JP-4 pool fires

    International Nuclear Information System (INIS)

    Keltner, N.R.; Kent, L.A.; Schneider, M.E.

    1987-01-01

    Over the past four years, Sandia National Laboratories has conducted a number of large pool fire tests to evaluate the design of radioactive material (RAM) shipping containers. Some of these tests have been designed to define the thermal environment and some have been used for certification testing. In each test there have been a number of fire diagnostic measurements. The simplest sets of diagnostics have involved measurements of temperature at several elevations on arrays of towers, measurements of hot wall heat flux with small calorimeters suspended from the towers, the average fuel recession rate, and the wind speed and direction. The most complex sets of diagnostics have included the above and in various tests added radiometers in the lower flame zone, centerline velocity measurements at a number of elevations, radiometers and calorimeters at the fuel surface, large cylindrical and flat plate calorimeters, infrared imaging, time resolved fuel recession rates, and a variety of soot particle concentration and size measurements made in the plume with a tethered balloon and an instrumented airplane. All of the large fires have been conducted in a 9.1 m by 18.3 m pool using JP-4 as the fuel. Typical duration is one-half hour. Covering all of the results is beyond the scope of a single paper. Conditionally sampled temperature and velocity measurements from one fire will be presented; for this fire, a 20 cm layer of fuel was floated on 61 cm of water. Pool surface heat flux, fuel recession rate data, and smoke emission data from a second fire are given. Because the wind has a strong effect on the temperature and velocity measurements, conditional sampling has been used to try to obtain data during periods of low winds. 10 refs., 3 figs

  4. Model of large pool fires

    Energy Technology Data Exchange (ETDEWEB)

    Fay, J.A. [Department of Mechanical Engineering, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)]. E-mail: jfay@mit.edu

    2006-08-21

    A two zone entrainment model of pool fires is proposed to depict the fluid flow and flame properties of the fire. Consisting of combustion and plume zones, it provides a consistent scheme for developing non-dimensional scaling parameters for correlating and extrapolating pool fire visible flame length, flame tilt, surface emissive power, and fuel evaporation rate. The model is extended to include grey gas thermal radiation from soot particles in the flame zone, accounting for emission and absorption in both optically thin and thick regions. A model of convective heat transfer from the combustion zone to the liquid fuel pool, and from a water substrate to cryogenic fuel pools spreading on water, provides evaporation rates for both adiabatic and non-adiabatic fires. The model is tested against field measurements of large scale pool fires, principally of LNG, and is generally in agreement with experimental values of all variables.

  5. Model of large pool fires

    International Nuclear Information System (INIS)

    Fay, J.A.

    2006-01-01

    A two zone entrainment model of pool fires is proposed to depict the fluid flow and flame properties of the fire. Consisting of combustion and plume zones, it provides a consistent scheme for developing non-dimensional scaling parameters for correlating and extrapolating pool fire visible flame length, flame tilt, surface emissive power, and fuel evaporation rate. The model is extended to include grey gas thermal radiation from soot particles in the flame zone, accounting for emission and absorption in both optically thin and thick regions. A model of convective heat transfer from the combustion zone to the liquid fuel pool, and from a water substrate to cryogenic fuel pools spreading on water, provides evaporation rates for both adiabatic and non-adiabatic fires. The model is tested against field measurements of large scale pool fires, principally of LNG, and is generally in agreement with experimental values of all variables

  6. Design of hydrotherapy exercise pools.

    Science.gov (United States)

    Edlich, R F; Abidin, M R; Becker, D G; Pavlovich, L J; Dang, M T

    1988-01-01

    Several hydrotherapy pools have been designed specifically for a variety of aquatic exercise. Aqua-Ark positions the exerciser in the center of the pool for deep-water exercise. Aqua-Trex is a shallow underwater treadmill system for water walking or jogging. Swim-Ex generates an adjustable laminar flow that permits swimming without turning. Musculoskeletal conditioning can be accomplished in the above-ground Arjo shallow-water exercise pool. A hydrotherapy pool also can be custom designed for musculoskeletal conditioning in its shallow part and cardiovascular conditioning in a deeper portion of the pool. Regardless of the type of exercise, there is general agreement that the specific exercise conducted in water requires significantly more energy expenditure than when the same exercise is performed on land.

  7. Modelling Hot Air Balloons.

    Science.gov (United States)

    Brimicombe, M. W.

    1991-01-01

    A macroscopic way of modeling hot air balloons using a Newtonian approach is presented. Misleading examples using a car tire and the concept of hot air rising are discussed. Pressure gradient changes in the atmosphere are used to explain how hot air balloons work. (KR)

  8. Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I

    International Nuclear Information System (INIS)

    Caro, R.; Ahnert, C.; Naudin, A. E.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-01-01

    In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

  9. Gene Pools and the Genetic Architecture of Domesticated Cowpea

    Directory of Open Access Journals (Sweden)

    Bao-Lam Huynh

    2013-11-01

    Full Text Available Cowpea [ (L. Walp.] is a major tropical legume crop grown in warm to hot areas throughout the world and especially important to the people of sub-Saharan Africa where the crop was domesticated. To date, relatively little is understood about its domestication origins and patterns of genetic variation. In this study, a worldwide collection of cowpea landraces and African ancestral wild cowpea was genotyped with more than 1200 single nucleotide polymorphism markers. Bayesian inference revealed the presence of two major gene pools in cultivated cowpea in Africa. Landraces from gene pool 1 are mostly distributed in western Africa while the majority of gene pool 2 are located in eastern Africa. Each gene pool is most closely related to wild cowpea in the same geographic region, indicating divergent domestication processes leading to the formation of two gene pools. The total genetic variation within landraces from countries outside Africa was slightly greater than within African landraces. Accessions from Asia and Europe were more related to those from western Africa while accessions from the Americas appeared more closely related to those from eastern Africa. This delineation of cowpea germplasm into groups of genetic relatedness will be valuable for guiding introgression efforts in breeding programs and for improving the efficiency of germplasm management.

  10. Thermo-hydraulic test of the moderator cell of liquid hydrogen cold neutron source for the Budapest research reactor

    International Nuclear Information System (INIS)

    Grosz, Tamas; Rosta, Laszlo; Hargitai, Tibor; Mityukhlyaev, V.A.; Serebrov, A.P.; Zaharov, A.A.

    1999-01-01

    Thermo-hydraulic experiment was carried out in order to test performance of the direct cooled liquid hydrogen moderator cell to be installed at the research reactor of the Budapest Neutron Center. Two electric hearers up to 300 W each imitated the nuclear heat release in the liquid hydrogen as well as in construction material. The test moderator cell was also equipped with temperature gauges to measure the hydrogen temperature at different positions as well as the inlet and outlet temperature of cooling he gas. The hydrogen pressure in the connected buffer volume was also controlled. At 140 w expected total heat load the moderator cell was filled with liquid hydrogen within 4 hours. The heat load and hydrogen pressure characteristics of the moderator cell are also presented. (author)

  11. Thermo-hydraulic performance of solar air heater having multiple v-shaped rib roughness on absorber plates

    Directory of Open Access Journals (Sweden)

    Dhananjay Kumar

    2018-03-01

    Full Text Available This paper presents the performance analysis of the effect of geometrical parameters having multiple v-shaped rib roughness on the airflow side of the absorber plates. Mathematical approach and solution procedure for the analysis of such a solar air heater has been developed theoretically and MATLAB code generated for the solution of the mathematical equations. The effect of parameters such as flow Reynolds number and Relative roughness height on the thermohydraulic performance have been examined and compared with the conventional flat plate solar air heater. A substantial improvement in thermal efficiency of roughened solar air heater as compared to smooth one due to appreciable enhancement in heat transfer coefficient. The enhancement in heat transfer coefficient is also accompanied by a considerable enhancement in pumping power requirement due to the increase in friction factor.

  12. Analyses and computer code developments for accident-induced thermohydraulic transients in water-cooled nuclear reactor systems

    International Nuclear Information System (INIS)

    Wulff, W.

    1977-01-01

    A review is presented on the development of analyses and computer codes for the prediction of thermohydraulic transients in nuclear reactor systems. Models for the dynamics of two-phase mixtures are summarized. Principles of process, reactor component and reactor system modeling are presented, as well as the verification of these models by comparing predicted results with experimental data. Codes of major importance are described, which have recently been developed or are presently under development. The characteristics of these codes are presented in terms of governing equations, solution techniques and code structure. Current efforts and problems of code verification are discussed. A summary is presented of advances which are necessary for reducing the conservatism currently implied in reactor hydraulics codes for safety assessment

  13. Mechanistic insights into a hydrate contribution to the Paleocene-Eocene carbon cycle perturbation from coupled thermohydraulic simulations

    Science.gov (United States)

    Minshull, T. A.; Marín-Moreno, H.; Armstrong McKay, D. I.; Wilson, P. A.

    2016-08-01

    During the Paleocene-Eocene Thermal Maximum (PETM), the carbon isotopic signature (δ13C) of surface carbon-bearing phases decreased abruptly by at least 2.5 to 3.0‰. This carbon isotope excursion (CIE) has been attributed to widespread methane hydrate dissociation in response to rapid ocean warming. We ran a thermohydraulic modeling code to simulate hydrate dissociation due to ocean warming for various PETM scenarios. Our results show that hydrate dissociation in response to such warming can be rapid but suggest that methane release to the ocean is modest and delayed by hundreds to thousands of years after the onset of dissociation, limiting the potential for positive feedback from emission-induced warming. In all of our simulations at least half of the dissociated hydrate methane remains beneath the seabed, suggesting that the pre-PETM hydrate inventory needed to account for all of the CIE is at least double that required for isotopic mass balance.

  14. Sensitiveness Analysis of Neutronic Parameters Due to Uncertainty in Thermo-hydraulic parameters on CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Serra, Oscar

    2000-01-01

    Some studies were done about the effect of the uncertainty in the values of several thermo-hydraulic parameters on the core behaviour of the CAREM-25 reactor.By using the chain codes CITVAP-THERMIT and the perturbation the reference states, it was found that concerning to the total power, the effects were not very important, but were much bigger for the pressure.Furthermore were hardly significant in the presence of any perturbation on the void fraction calculation and the fuel temperature.The reactivity and the power peaking factor had highly important changes in the case of the coolant flow.We conclude that the use of this procedure is adequate and useful to our purpose

  15. ASCOT-1: a computer program for analyzing the thermo-hydraulic behavior in a PWR core during a LOCA

    International Nuclear Information System (INIS)

    Kobayashi, Kensuke; Sato, Kazuo

    1978-09-01

    A digital computer code ASCOT-1 has been developed to analyze the thermo-hydraulic behavior in a PWR core during a loss-of-coolant accident. The core is assumed to be axi-symmetric two-dimensional and the conservation laws are solved by the method of characteristics. For the temperature response of representative fuels of the concentric annular subregions into which the core is divided, the heat conduction equations are solved by the explicit method with the averaged flow conditions decided above. The boundary conditions at the upper and lower plenum are given as inputs. The program is of an adjustable dimension so there are no restrictions to the numbers of meshes. ASCOT-1 is written in FORTRAN-IV for FACOM230-75. (author)

  16. Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

    2006-07-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

  17. ENERGY STAR Certified Pool Pumps

    Data.gov (United States)

    U.S. Environmental Protection Agency — Certified models meet all ENERGY STAR requirements as listed in the Version 1.1 ENERGY STAR Program Requirements for Pool Pumps that are effective as of February 15,...

  18. Pooling and correlated neural activity

    Directory of Open Access Journals (Sweden)

    Robert Rosenbaum

    2010-04-01

    Full Text Available Correlations between spike trains can strongly modulate neuronal activity and affect the ability of neurons to encode information. Neurons integrate inputs from thousands of afferents. Similarly, a number of experimental techniques are designed to record pooled cell activity. We review and generalize a number of previous results that show how correlations between cells in a population can be amplified and distorted in signals that reflect their collective activity. The structure of the underlying neuronal response can significantly impact correlations between such pooled signals. Therefore care needs to be taken when interpreting pooled recordings, or modeling networks of cells that receive inputs from large presynaptic populations. We also show that the frequently observed runaway synchrony in feedforward chains is primarily due to the pooling of correlated inputs.

  19. Grundfoss: Chlorination of Swimming Pools

    DEFF Research Database (Denmark)

    Hjorth, Poul G.; Hogan, John; Andreassen, Viggo

    1998-01-01

    Grundfos asked for a model, describing the problem of mixing chemicals, being dosed into water systems, to be developed. The application of the model should be dedicated to dosing aqueous solution of chlorine into swimming pools.......Grundfos asked for a model, describing the problem of mixing chemicals, being dosed into water systems, to be developed. The application of the model should be dedicated to dosing aqueous solution of chlorine into swimming pools....

  20. Sustainability of common pool resources

    OpenAIRE

    Timilsina, Raja Rajendra; Kotani, Koji; Kamijo, Yoshio

    2017-01-01

    Sustainability has become a key issue in managing natural resources together with growing concerns for capitalism, environmental and resource problems. We hypothesize that the ongoing modernization of competitive societies, which we refer to as "capitalism," affects human nature for utilizing common pool resources, thus compromising sustainability. To test this hypothesis, we design and implement a set of dynamic common pool resource games and experiments in the following two types of Nepales...

  1. Water treatment in public swimming pools - reduction of energy consumption; Vandbehandling i svoemmebade - reduktion af energiforbrug

    Energy Technology Data Exchange (ETDEWEB)

    Hammerich, H.; Radisch, N. (Ramboell, Koege (Denmark)); Olesen, Jens Christian (Gladsaxe Sportscenter, Gladsaxe (Denmark)) (and others)

    2010-04-15

    Measurements were made in five public swimming baths, and energy savings were achieved using new filters, pumps, water treatment control depending on bather load, etc. In a 50 metre pool, electricity consumption for water treatment decreased by 50%, and in a hot-water/paddling pool, electricity consumption decreased by 30-40% while still maintaining satisfactory water quality - even during periods of heavy bather load. In another swimming bath, ventilation electricity consumption was reduced by 15%. The results will e.g. be used to revise the Danish executive order on swimming pools and water quality to allow bather load-dependent water circulation. (ln)

  2. Hot Surface Ignition

    OpenAIRE

    Tursyn, Yerbatyr; Goyal, Vikrant; Benhidjeb-Carayon, Alicia; Simmons, Richard; Meyer, Scott; Gore, Jay P.

    2015-01-01

    Undesirable hot surface ignition of flammable liquids is one of the hazards in ground and air transportation vehicles, which primarily occurs in the engine compartment. In order to evaluate the safety and sustainability of candidate replacement fuels with respect to hot surface ignition, a baseline low lead fuel (Avgas 100 LL) and four experimental unleaded aviation fuels recommended for reciprocating aviation engines were considered. In addition, hot surface ignition properties of the gas tu...

  3. Seismic analysis of large pools

    Energy Technology Data Exchange (ETDEWEB)

    Dong, R.G.; Tokarz, F.J.

    1976-11-17

    Large pools for storing spent, nuclear fuel elements are being proposed to augment present storage capacity. To preserve the ability to isolate portions of these pools, a modularization requirement appears desirable. The purpose of this project was to investigate the effects of modularization on earthquake resistance and to assess the adequacy of current design methods for seismic loads. After determining probable representative pool geometries, three rectangular pool configurations, all 240 x 16 ft and 40 ft deep, were examined. One was unmodularized; two were modularized into 80 x 40 ft cells in one case and 80 x 80 ft cells in the other. Both embedded and above-ground installations for a hard site and embedded installations for an intermediate hard site were studied. It was found that modularization was unfavorable in terms of reducing the total structural load attributable to dynamic effects, principally because one or more cells could be left unfilled. The walls of unfilled cells would be subjected to significantly higher loads than the walls of a filled, unmodularized pool. Generally, embedded installations were preferable to above-ground installations, and the hard site was superior to the intermediate hard site. It was determined that Housner's theory was adequate for calculating hydrodynamic effects on spent fuel storage pools. Current design methods for seismic loads were found to be satisfactory when results from these methods were compared with those from LUSH analyses. As a design method for dynamic soil pressure, we found the Mononobe-Okabe theory, coupled with correction factors as suggested by Seed, to be acceptable. The factors we recommend for spent fuel storage pools are tabulated.

  4. Seismic analysis of large pools

    International Nuclear Information System (INIS)

    Dong, R.G.; Tokarz, F.J.

    1976-01-01

    Large pools for storing spent, nuclear fuel elements are being proposed to augment present storage capacity. To preserve the ability to isolate portions of these pools, a modularization requirement appears desirable. The purpose of this project was to investigate the effects of modularization on earthquake resistance and to assess the adequacy of current design methods for seismic loads. After determining probable representative pool geometries, three rectangular pool configurations, all 240 x 16 ft and 40 ft deep, were examined. One was unmodularized; two were modularized into 80 x 40 ft cells in one case and 80 x 80 ft cells in the other. Both embedded and above-ground installations for a hard site and embedded installations for an intermediate hard site were studied. It was found that modularization was unfavorable in terms of reducing the total structural load attributable to dynamic effects, principally because one or more cells could be left unfilled. The walls of unfilled cells would be subjected to significantly higher loads than the walls of a filled, unmodularized pool. Generally, embedded installations were preferable to above-ground installations, and the hard site was superior to the intermediate hard site. It was determined that Housner's theory was adequate for calculating hydrodynamic effects on spent fuel storage pools. Current design methods for seismic loads were found to be satisfactory when results from these methods were compared with those from LUSH analyses. As a design method for dynamic soil pressure, we found the Mononobe-Okabe theory, coupled with correction factors as suggested by Seed, to be acceptable. The factors we recommend for spent fuel storage pools are tabulated

  5. Multi-physics modeling and numerical simulation of weld pool in GTA welding

    International Nuclear Information System (INIS)

    Nguyen, Minh-Chien

    2015-01-01

    In this work, we develop a 3D physical and numerical model of the GTA (Gas Tungsten Arc) welding process in order to predict, for given welding parameters, useful quantities for the designer of welded assembly: weld bead shape, fluid flow in the weld pool as well as thermal distribution in the work piece. The model is developed in the Cast3M (http://www-cast3m.cea.fr/) finite element software and takes into account the main physical phenomena acting in the work piece and particularly in the weld pool, subject to source terms modeling the arc part of the welding process. A steady solution of this model is thought for and involves the coupling of the nonlinear thermohydraulics and electromagnetic equations together with the displacement of the deformable free surface of the weld pool. A first step in the development consisted in modeling the electromagnetic phenomena with two different numerical methods, in comparing the numerical results obtained with those of the literature and in quantifying the importance of the Lorentz force and the Joule effect compared to the other mechanical and thermal sources by computing power balances. Then, in order to assess the predictive capability of the model, simulations of various welding configurations are performed: variation in the chemical composition of the material, of the welding speed, of the prescribed arc pressure and of the welding positions, which is a focus of this work, are studied. A good agreement is obtained between the results of our model and other experimental and numerical results of the literature. Eventually, a model accounting for metal filling is proposed and its results are discussed. Thus, our complete model can be seen as a solid foundation towards future totally-coupled 3D welding models including the arc and it will be included in WPROCESS the in-house CEA software dedicated to the numerical simulation of welding. (author) [fr

  6. Hot ductility testing and weld simulation tests

    International Nuclear Information System (INIS)

    Weber, G.; Schick, M.

    1999-01-01

    The objective of the project was to enhance the insight into the causes of intergranular cracks detected in austenitic circumferential welds at BWR pipes. The susceptibility of a variety of austenitic pipe materials to hot cracking during welding and in-service intergranular crack corrosion was examined. The assumption was cracking in the root area of the HAZ of a multiple-layer weld. Hot-ductility tests and weld simulation tests specifically designed for the project were performed with the austenitic LWR pipe materials 1.4553 (X6 CrNiNb 18 10 S), 1.4550 (X10 CrNiNb 18 9), 1.4533 (X6 CrNiTi 18 9, two weld pools), and a non-stabilized TP 304 (X5 CrNi 18 10). (orig./CB) [de

  7. Scintigraphic detection of gynaecological tumours by sup(99m)Tc-HSA blood pool imaging

    Energy Technology Data Exchange (ETDEWEB)

    Georgiou, E C; Arkas, A Z; Alevizaki, C C; Gnafakis, N [Evangelismos Hospital, Athens (Greece). Dept. of Endocrinology and Radioisotopes; Evangelismos Hospital, Athens (Greece). Gynaecology Clinic)

    1979-12-01

    Three cases of scintigraphic visualization of gynaecological tumours by sup(99m)Tc-HSA blood pool imaging are presented, i.e. one case of fibrothecoma ('cold' area visualization) and two cases of fibromyoma ('hot' area visualization). It is suggested that this technique may be useful in the preoperative evaluation of gynaecological tumours and their vascularity.

  8. Thermal, thermo-hydraulic and thermo-mechanic analysis for fuel elements of IEA-R1 reactor at 5MW

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Silva Macedo, L.V. da

    1989-01-01

    In connection with the on going conversion of IEA-R1 Research Reactor, operated by IPEN-CNEN/SP, from the use of highly enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel, steady-state thermal and thermo-hydraulic analysis of both existing HEU and proposed LEU cores under 2 MW operating conditions have been carried out. Keeping in mind the possibility of power upgrading, steady-state thermal, thermo-hydraulic and thermomechanical analysis of proposed LEU core under 5 MW operating conditions have also been carried out. The thermal and thermo-hydraulic analysis at 2 MW show that the conversion of the existing HEU core to be proposed LEU core will not change the reactor safety margins. Although the upgrading of the reactor power to 5 MW will result in safety margins lower than in case of 2MW, these will be still sufficient for optimum operation and safe behaviour. The thermomechanical analysis at 5 MW show that the thermal stresses induced in the fuel element will satisfy the design limits for mechanical strenght and elastic stability. (author) [pt

  9. Analysis of loads and fluid-structure interactions in a condensation pool

    Energy Technology Data Exchange (ETDEWEB)

    Timperi, A.; Paettikangas, T.; Niemi, J. [VTT, Technical Research Centre of Finland (Finland)

    2007-04-15

    A simplified direct contact condensation model was implemented into the Volume of Fluid model of the Fluent CFD code. Transient three-dimensional test runs for the POOLEX experiments, where steam is blown into a water pool were performed. The model was found to provide too small condensation rate for steam when compared to experiments. In addition, the calculated back and forth oscillation of the steam water interface was much smaller than in the experiments. The model was found to be numerically quite robust. The discrepancies of the simulation, such as the too small condensation rate, could be to some extent cured by making improvements into the condensation model. As an alternative estimation method of thermohydraulic loads in condensation pools, the SILA code based on potential flow theory, was taken into use. SILA solves the pressure distribution caused by oscillating bubbles in a pool, and is easier to use and more flexible than Method of Images studied earlier. SILA has been modified for pools without an inner cylinder and test simulations for the POOLEX water pool were performed. The MpCCI FSI coupling software employs an explicit coupling scheme, which results in numerical instability in the case of the POOLEX facility. In order to improve stability, ways for implementing an implicit coupling scheme with MpCCI were examined. It was found that such a scheme is difficult to achieve without access to the source codes. An implicit coupling scheme is expected to be available with MpCCI in forthcoming years. A method was developed which can be used for analysing two-way FSI problems realistically by using only one-way coupling of CFD and structural analysis codes. In the method, the mass of the fluid is accounted for in the structural motion by adding the fluid to the structural model as an acoustic medium. Validity of the method was examined with promising results mathematically by an order of magnitude analysis and by comparing numerical results with a full

  10. Plenum separator system for pool-type nuclear reactors

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1983-01-01

    This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

  11. 21 CFR 1250.89 - Swimming pools.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Swimming pools. 1250.89 Section 1250.89 Food and... SANITATION Sanitation Facilities and Conditions on Vessels § 1250.89 Swimming pools. (a) Fill and draw swimming pools shall not be installed or used. (b) Swimming pools of the recirculation type shall be...

  12. Patent pools: Intellectual property rights and competition.

    NARCIS (Netherlands)

    Rodriguez, V.F.

    2010-01-01

    Patent pools do not correct all problems associated with patent thickets. In this respect, patent pools might not stop the outsider problem from striking pools. Moreover, patent pools can be expensive to negotiate, can exclude patent holders with smaller numbers of patents or enable a group of major

  13. Simplified analysis of trasients in pool type liquid metal reactors

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1987-01-01

    The conceptual design of a liquid metal fast breeder reactor will require a great effort of development in several technical disciplines. One of them is the thermal-hydraulic design of the reactor and of the heat and fluid transport components inside the reactor vessel. A simplified model to calculate the maximum sodium temperatures is presented in this paper. This model can be used to optimize the layout of components inside the reactor vessel and was easily programmed in a small computer. Illustrative calculations of two transients of a typical hot pool type fast reactor are presented and compared with the results of other researchers. (author) [pt

  14. Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    {sub th} power and electricity generation with 100 kW{sub th} idle power. Consequently, KANUTER has the characteristics of a compact and lightweight system, excellent propellant efficiency, bimodal capability, and mission versatility as indicated in the reference design parameters. This thermo-hydraulic design analysis was carried out to estimate the optimum FWT of the unique SLHC fuel design in the core and thereby the maximum rocket performance. The FWT affects the mechanical strength of the SLHC fuel assembly as well as the thermo-hydraulic capability mainly depending on the heat transfer area of fuel. The thicker fuel wafer is mechanically strong with low pressure drop, while the thinner fuel wafer is thermally robust with less mechanical strength and higher shear stress in the core.

  15. Efficient pooling designs for library screening

    OpenAIRE

    Bruno, William J.; Knill, Emanuel; Balding, David J.; Bruce, D. C.; Doggett, N. A.; Sawhill, W. W.; Stallings, R. L.; Whittaker, Craig C.; Torney, David C.

    1994-01-01

    We describe efficient methods for screening clone libraries, based on pooling schemes which we call ``random $k$-sets designs''. In these designs, the pools in which any clone occurs are equally likely to be any possible selection of $k$ from the $v$ pools. The values of $k$ and $v$ can be chosen to optimize desirable properties. Random $k$-sets designs have substantial advantages over alternative pooling schemes: they are efficient, flexible, easy to specify, require fewer pools, and have er...

  16. EP BICYCLE POOL - VIGNETTES 2002

    CERN Multimedia

    EP-SMI Help Desk

    2002-01-01

    The vignettes (insurance certificates) for 2002 become obligatory from 1 June. If you have a bicycle from the EP Pool, please bring it to the EP-SMI Help Desk (Building 124) on any working day up to 31 May between 8h.30 - 12h.00 or 13h.30 - 17h.30. EP-SMI Help Desk

  17. Sculpting the Barnyard Gene Pool

    Science.gov (United States)

    Childers, Gina; Wolfe, Kim; Dupree, Alan; Young, Sheila; Caver, Jessica; Quintanilla, Ruby; Thornton, Laura

    2016-01-01

    Project-based learning (PBL) takes student engagement to a higher level through reflective collaboration, inquiry, critical thinking, problem solving, and personal relevance. This article explains how six high school teachers developed an interconnected, interdisciplinary STEM-focused PBL called "Sculpting the Barnyard Gene Pool." The…

  18. Thermo-Hydraulic Analysis of Heat Storage Filled with the Ceramic Bricks Dedicated to the Solar Air Heating System.

    Science.gov (United States)

    Nemś, Magdalena; Nemś, Artur; Kasperski, Jacek; Pomorski, Michał

    2017-08-12

    This article presents the results of a study into a packed bed filled with ceramic bricks. The designed storage installation is supposed to become part of a heating system installed in a single-family house and eventually to be integrated with a concentrated solar collector adapted to climate conditions in Poland. The system's working medium is air. The investigated temperature ranges and air volume flow rates in the ceramic bed were dictated by the planned integration with a solar air heater. Designing a packed bed of sufficient parameters first required a mathematical model to be constructed and heat exchange to be analyzed, since heat accumulation is a complex process influenced by a number of material properties. The cases discussed in the literature are based on differing assumptions and different formulas are used in calculations. This article offers a comparison of various mathematical models and of system operating parameters obtained from these models. The primary focus is on the Nusselt number. Furthermore, in the article, the thermo-hydraulic efficiency of the investigated packed bed is presented. This part is based on a relationship used in solar air collectors with internal storage.

  19. Thermo-hydraulic Analysis of a Water-cooled Printed Circuit Heat Exchanger in a Small-scale Nitrogen Loop

    International Nuclear Information System (INIS)

    Kim, Chan Soo; Hong, Sung Deok; Kim, Min Hwan; Shim, Jaesool; Lee, Gyung Dong

    2013-01-01

    The development of high-temperature heat exchangers is very important because of its higher operation temperature and pressure than those of common light water reactors and industrial process plants. In particular, the intermediate heat exchanger is a key-challenged high temperature component in a Very High Temperature gas-cooled Reactor (VHTR). A printed circuit heat exchanger is one of the candidates for an intermediate heat exchanger in a VHTR. The printed circuit heat exchanger (PCHE) was developed and commercialized by HEATRIC. The compactness is better than any other heat exchanger types, because its core matrices are fabricated by diffusion bonding with photo-chemically etched micro-channels. Various tests and analysis have been performed to verify the performance of PCHE. The thermal stress analysis of the high temperature PCHE is necessary to endure the extremely operation condition of IHX. In this study, the thermo-hydraulic analysis for the laboratory-scale PCHE is performed to provide the input data for the boundary conditions of a structural analysis. The results from the first-principal calculation are compared with those from computational fluid dynamics code analysis. COMSOL 4.3a analysis is successfully performed at the uniform pressure drop condition in a set of flow channel stacks. The heat-exchanged region concentrated to the nitrogen inlet cause the uniform mass velocity distribution in the channels, therefore there is little difference between two analytical results

  20. Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

    Energy Technology Data Exchange (ETDEWEB)

    Arndt, S.A.

    1997-07-01

    The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for code use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities.

  1. IAEA coordinated research programme on heat transfer behavior and thermo-hydraulics code testing for super critical water cooled reactors

    International Nuclear Information System (INIS)

    Bilbao y Leon, Sama; Aksan, Nusret

    2009-01-01

    One of the key roles of the IAEA is to foster the collaboration among Member States on the development of advances in technology for advanced nuclear power plants. There is high international interest, both in developing and industrialized countries, in innovative supercritical water-cooled reactors (SCWRs), primarily because such concepts will achieve high thermal efficiencies (44-45%) and promise improved economic competitiveness utilizing and building upon the recent developments for highly efficient fossil power plants. The SCWR has been selected as one of the promising concepts for development by the Generation-IV International Forum. Following the advice of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and TWG-HWR), with the feedback from the Gen-IV SCWR Steering Committee, and in coordination with the OECD-NEA, IAEA has recently started a Coordinated Research Programme (CRP) in the areas of heat transfer behaviour and testing of thermo-hydraulic computer methods for Supercritical Water-Cooled Reactors. The first Research Coordination Meeting (RCM) of the CRP was held at the IAEA Headquarters, in Vienna, Austria in July 2008. This paper summarizes the current status of the CRP, including the Integrated Research Plan and the general schedule for the CRP. (author)

  2. Thermo-Hydraulic Analysis of Heat Storage Filled with the Ceramic Bricks Dedicated to the Solar Air Heating System

    Science.gov (United States)

    Nemś, Magdalena; Nemś, Artur; Kasperski, Jacek; Pomorski, Michał

    2017-01-01

    This article presents the results of a study into a packed bed filled with ceramic bricks. The designed storage installation is supposed to become part of a heating system installed in a single-family house and eventually to be integrated with a concentrated solar collector adapted to climate conditions in Poland. The system’s working medium is air. The investigated temperature ranges and air volume flow rates in the ceramic bed were dictated by the planned integration with a solar air heater. Designing a packed bed of sufficient parameters first required a mathematical model to be constructed and heat exchange to be analyzed, since heat accumulation is a complex process influenced by a number of material properties. The cases discussed in the literature are based on differing assumptions and different formulas are used in calculations. This article offers a comparison of various mathematical models and of system operating parameters obtained from these models. The primary focus is on the Nusselt number. Furthermore, in the article, the thermo-hydraulic efficiency of the investigated packed bed is presented. This part is based on a relationship used in solar air collectors with internal storage. PMID:28805703

  3. 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)

  4. Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

    International Nuclear Information System (INIS)

    Arndt, S.A.

    1997-01-01

    The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for code use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities

  5. Parallelization of TWOPORFLOW, a Cartesian Grid based Two-phase Porous Media Code for Transient Thermo-hydraulic Simulations

    Science.gov (United States)

    Trost, Nico; Jiménez, Javier; Imke, Uwe; Sanchez, Victor

    2014-06-01

    TWOPORFLOW is a thermo-hydraulic code based on a porous media approach to simulate single- and two-phase flow including boiling. It is under development at the Institute for Neutron Physics and Reactor Technology (INR) at KIT. The code features a 3D transient solution of the mass, momentum and energy conservation equations for two inter-penetrating fluids with a semi-implicit continuous Eulerian type solver. The application domain of TWOPORFLOW includes the flow in standard porous media and in structured porous media such as micro-channels and cores of nuclear power plants. In the latter case, the fluid domain is coupled to a fuel rod model, describing the heat flow inside the solid structure. In this work, detailed profiling tools have been utilized to determine the optimization potential of TWOPORFLOW. As a result, bottle-necks were identified and reduced in the most feasible way, leading for instance to an optimization of the water-steam property computation. Furthermore, an OpenMP implementation addressing the routines in charge of inter-phase momentum-, energy- and mass-coupling delivered good performance together with a high scalability on shared memory architectures. In contrast to that, the approach for distributed memory systems was to solve sub-problems resulting by the decomposition of the initial Cartesian geometry. Thread communication for the sub-problem boundary updates was accomplished by the Message Passing Interface (MPI) standard.

  6. Hot Weather Tips

    Science.gov (United States)

    ... the person plenty of water and fruit or vegetable juice even if they say they’re not thirsty. No alcohol, coffee or tea. Seek medical help if you suspect dehydration. Light meals: Avoid hot, heavy meals and don’ ...

  7. China's 'Hot Money' Problems

    National Research Council Canada - National Science Library

    Martin, Michael F; Morrison, Wayne M

    2008-01-01

    .... The recent large inflow of financial capital into China, commonly referred to as "hot money," has led some economists to warn that such flows may have a destabilizing effect on China's economy...

  8. Phenomenology and thermo-hydraulic stability of the CAREM-25 reactor: Evaluation of subcooled boiling effect

    International Nuclear Information System (INIS)

    Acuna, F.M.; Marcel, C.P.; Zanocco, P.G.; Delmastro, D.F.

    2012-01-01

    In this article the phenomenology present in self/pressurized, integral, natural circulation, low thermodynamic quality nuclear reactors similar to CAREM-25 is investigated. In particular, analytical relations for the mass flow rate, the core mean enthalpy and the location of the two phase boundary are derived in terms of the so-called natural variables of the system: the nuclear power, the condensation power and the system pressure. In addition, some consequences of the flashing phenomenon in the reactor thermal-hydraulics are discussed emphasizing those affecting the reactor stability. The reactor stability performance was studied by using the HUARPE code which is a low diffusive code. The stability results obtained by neglecting the subcooled effect in the system are presented in the so-called the stability maps in which the results are presented for a wide range of conditions. The stability effect caused by the presence of subcooled boiling in the reactor core was also examined. In order to investigate such a consequence, the code was slightly modified such that the predicted vapor profile in the hot leg is similar to that estimated by RELAP system code at steady state conditions. The simple implemented algorithm allows varying a free parameter with which a broad number of cases can be studied. This is important since the subcooled boiling predictions generally have large uncertainties and therefore to cover a large number of situations is desired to derive confident conclusions. The results show the existence of vapor created by means of subcooled boiling enhances the system stability for a wide range of conditions. For this reason from this preliminary investigation, it is concluded neglecting the subcooled effect in CAREM-25 stability studies is a conservative criterion (author))

  9. "Ripples" in an Aluminum Pool?

    Science.gov (United States)

    Rohr, James; Wang, Si-Yin; Nesterenko, Vitali F.

    2018-05-01

    Our motivation for this article is for students to realize that opportunities for discovery are all around them. Discoveries that can still puzzle present day researchers. Here we explore an observation by a middle school student concerning the production of what appears to be water-like "ripples" produced in aluminum foil when placed between two colliding spheres. We both applaud and explore the student's reasoning that the ripples were formed in a melted aluminum pool.

  10. Fire preparedness measures in buildings with hot laboratories

    International Nuclear Information System (INIS)

    Oberlaender, B.C.

    2003-01-01

    Important hot laboratory safety issues are the general design/construction of the building with respect to fire, fire prevention, fire protection, administrative controls, and risk assessment. Within the network of the European Working Group Hot Laboratories and Remote Handling items concerning 'fire preparedness measures in hot laboratories' were screened and studied. Two questionnaires were sent to European hot laboratories; the first in November 2002 on 'fire preparedness measures, fire detection and fire suppression/extinguishing in lead shielded cells, concrete shielded cells' and the second in June 2003 on 'Fire preparedness measures in buildings with hot laboratories'. The questionnaires were filled in by a total of ten hot laboratories in seven European countries. On request of participants the answers were evaluated and 'anonymised' for presentation and discussion at the plenary meeting. The answers showed that many European hot laboratories are implementing improvements to their fire protection programmes to comply with more stringent requirements of the national authorities. The recommendations ('International guidelines for the fire protection of Nuclear Power Plants') given by the insurance pools are followed up with national variations. An ISO standard (ISO 17873) is in progress giving criteria for the design and the operation of ventilation systems as well as fire hazard management in nuclear installations others than reactors

  11. Sustainability of common pool resources.

    Science.gov (United States)

    Timilsina, Raja Rajendra; Kotani, Koji; Kamijo, Yoshio

    2017-01-01

    Sustainability has become a key issue in managing natural resources together with growing concerns for capitalism, environmental and resource problems. We hypothesize that the ongoing modernization of competitive societies, which we refer to as "capitalism," affects human nature for utilizing common pool resources, thus compromising sustainability. To test this hypothesis, we design and implement a set of dynamic common pool resource games and experiments in the following two types of Nepalese areas: (i) rural (non-capitalistic) and (ii) urban (capitalistic) areas. We find that a proportion of prosocial individuals in urban areas is lower than that in rural areas, and urban residents deplete resources more quickly than rural residents. The composition of proself and prosocial individuals in a group and the degree of capitalism are crucial in that an increase in prosocial members in a group and the rural dummy positively affect resource sustainability by 65% and 63%, respectively. Overall, this paper shows that when societies move toward more capitalistic environments, the sustainability of common pool resources tends to decrease with the changes in individual preferences, social norms, customs and views to others through human interactions. This result implies that individuals may be losing their coordination abilities for social dilemmas of resource sustainability in capitalistic societies.

  12. Flashing oscillation in pool water

    International Nuclear Information System (INIS)

    Takamasa, Tomoji; Kondo, Koichi; Hazuku, Tatsuya

    1996-01-01

    This paper presents an experimental study of high-pressure saturated water discharging into the pool water. The purpose of the experiment is to clarify the phenomena that occur in blow-down of high-pressure saturated water from the pressure vessel into the water-filled containment in the case of a wall-crack accident or a LOCA in an advanced reactor. The results revealed that a flashing oscillation (FO) occurs when high-pressure saturated water discharges into the pool water, under specified experimental settings. The range of the flashing oscillates between a point very close to and some distance from the vent hole. The pressures in the vent tube and pool water vary according to the flashing oscillation. The pressure oscillation and frequency of flashing position might be caused by the balancing action between the supply of saturated water, flashing at the control volume and its condensation on the steam-water interface. A linear analysis was conducted using a spherical flashing bubble model. The period of the flashing oscillation in the experiments can be explained by theoretical analysis

  13. Pressure supression pool thermal mixing

    International Nuclear Information System (INIS)

    Cook, D.H.

    1984-10-01

    A model is developed and verified to describe the thermal mixing that occurs in the pressure suppression pool (PSP) of a commercial BWR. The model is designed specifically for a Mark-I containment and is intended for use in severe accident sequence analyses. The model developed in this work produces space and time dependent temperature results throughout the PSP and is useful for evaluating the bulk PSP thermal mixing, the condensation effectiveness of the PSP, and the long-term containment integrity. The model is designed to accommodate single or multiple discharging T-quenchers, a PSP circumferential circulation induced by the residual heat removal system discharge, and the thermal stratification of the pool that occurs immediately after the relief valves close. The PSP thermal mixing is verified by comparing the model-predicted temperatures to experimental temperatures that were measured in an operating BWR suppression pool. The model is then used to investigate several PSP thermal mixing problems that include the time to saturate at full relief valve flow, the temperature response to a typical stuck open relief valve scenario, and the effect of operator rotation of the relief valve discharge point

  14. Pressure suppression pool thermal mixing

    International Nuclear Information System (INIS)

    Cook, D.H.

    1984-01-01

    A model is developed and verified to describe the thermal mixing that occurs in the pressure suppression pool (PSP) of a commercial BWR. The model is designed specifically for a Mark-I containment and is intended for use in severe accident sequence analyses. The model produces space and time dependent temperature results throughout the PSP and is useful for evaluating the bulk PSP thermal mixing, the condensation effectiveness of the PSP, and the long-term containment integrity. The model is designed to accommodate single or multiple discharging T-quenchers, a PSP circumferential circulation induced by the residual heat removal system discharge, and the thermal stratification of the pool that occurs immediately after the relief valves close. The PSP thermal mixing model is verified by comparing the model predicted temperatures to experimental temperatures that were measured in an operating BWR suppression pool. The model is then used to investigate several PSP thermal mixing problems that include the time to saturate at full relief valve flow, the temperature response to a typical stuck open relief valve scenario, and the effect of operator rotation of the relief valve discharge point

  15. Groundwater dependent pools in seasonal and permanent streams in the Clare Valley of South Australia

    Directory of Open Access Journals (Sweden)

    Erick Bestland

    2017-02-01

    In this Mediterranean climate with cool wet winters and dry hot summers strong salinity changes (up to 2.5 times due to seasonal cycles of wetting and drying were observed in surface water. Oxygen and hydrogen isotope values from pool sites showed strong evaporative enrichment during the dry season with up to 50% net evaporation calculated. Water isotopes from groundwater, however, cluster at the depleted end of the local meteoric water line and most do not show change despite significant seasonal salinity changes. Strontium isotope values and concentrations from the pools over the one year period do not define a mixing relationship. Instead, most pool sites have unchanging strontium isotope values despite the large seasonal change in salinity indicating strong evaporation of groundwater fed pools during this drought year.

  16. CDC Study Finds Fecal Contamination in Pools

    Science.gov (United States)

    ... Communication (404) 639-3286 CDC study finds fecal contamination in pools A study of public pools done ... The E. coli is a marker for fecal contamination. Finding a high percentage of E. coli-positive ...

  17. Energy flux of hot atoms

    International Nuclear Information System (INIS)

    Wotzak, G.P.; Kostin, M.D.

    1976-01-01

    The process in which hot atoms collide with thermal atoms of a gas, transfer kinetic energy to them, and produce additional hot atoms is investigated. A stochastic method is used to obtain numerical results for the spatial and time dependent energy flux of hot atoms in a gas. The results indicate that in hot atom systems a front followed by an intense energy flux of hot atoms may develop

  18. 31P-NMR differentiation between intracellular phosphate pools in Cosmarium (chlorophyta)

    International Nuclear Information System (INIS)

    Elgavish, A.; Elgavish, G.A.

    1980-01-01

    31 P nuclear magnetic resonance (NMR) spectroscopy of intact Cosmarium sp. cells is presented as a suitable tool for the differentiation of intracellular accumulation pools of polyphosphates. The cold trichloroacetic acid (TCA) insoluble fraction is shown to contain most of the total cellular phosphate in the phosphate rich Cosmarium cells. Moreover, evidence from a 31 P-NMR study and electron microscopic observations of cold TCA treated Cosmarium cells indicate that this fraction consists mostly of polyphosphates which seem to retain the native morphological structure observed in the untreated cells. The determination of orthophosphate in the hot water extract of Cosmarium cells did not measure the polyphosphate pools. Determination of total phosphorus content in the hot water extract rendered a value three times higher than the frequently used orthophosphate determination procedure. However, as revealed by the 31 P-NMR spectra and the chemical analyses of the extract and of the treated cells, even total phosphorus in the extract measured only 30% of the total cellular phosphorus. 31 P-NMR enabled the unequivocal chemical identification of the major phosphate compounds in the hot water extract (Surplus P) as orthophosphate and polyphosphates of about 10 phosphate units chainlength. More than 70% of the accumulation pool of polyphosphates was still in the cells after extraction. However, the electron microscopy study revealed that the native granular structure of polyphosphates had been destroyed by the hot water extraction procedure

  19. Analysis Thermo-hydraulic of trajectories related to procedures for operation of Emergency (POE). Application to the loss of a train of the DTH; Analisis termohidraulico de trayectorias vinculadas a Procedimientos de Operacion de emergencia (POE). Aplicacion a la perdida de un tren de RHR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Saez, F.; Martorell Alsina, S.; Carlos Alberola, A.; Villanueva Lopez, J. F.; Martorell Aygues, P.

    2012-07-01

    This work explores different possible sequences at the loss of a train of the DTH when the plant is lowering power. The study of the different possible trajectories has been done through the collapse tool and study thermo-hydraulic each of these paths is done by the code TRACE Thermo-hydraulic.

  20. Sol Duc Hot Springs feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    1981-12-01

    Sol Duc Springs is located in the Olympic National Park in western Washington state. Since the turn of the century, the area has served as a resort, offering hot mineral baths, lodge and overnight cabin accommodations. The Park Service, in conjunction with the concessionaire, is in the process of renovating the existing facilities, most of which are approximately 50 years old. The present renovation work consists of removing all of the existing cabins and replacing them with 36 new units. In addition, a new hot pool is planned to replace the existing one. This report explores the possibility of a more efficient use of the geothermal resource to accompany other planned improvements. It is important to note that the system outlined is based upon the resource development as it exists currently. That is, the geothermal source is considered to be: the two existing wells and the hot springs currently in use. In addition, every effort has been made to accommodate the priorities for utilization as set forth by the Park Service.

  1. Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

    International Nuclear Information System (INIS)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung

    2015-01-01

    The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW th power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H 2 corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the preliminary design

  2. Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW{sub th} power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H{sub 2} corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the

  3. HotRegion: a database of predicted hot spot clusters.

    Science.gov (United States)

    Cukuroglu, Engin; Gursoy, Attila; Keskin, Ozlem

    2012-01-01

    Hot spots are energetically important residues at protein interfaces and they are not randomly distributed across the interface but rather clustered. These clustered hot spots form hot regions. Hot regions are important for the stability of protein complexes, as well as providing specificity to binding sites. We propose a database called HotRegion, which provides the hot region information of the interfaces by using predicted hot spot residues, and structural properties of these interface residues such as pair potentials of interface residues, accessible surface area (ASA) and relative ASA values of interface residues of both monomer and complex forms of proteins. Also, the 3D visualization of the interface and interactions among hot spot residues are provided. HotRegion is accessible at http://prism.ccbb.ku.edu.tr/hotregion.

  4. Microbial hotspots and hot moments in soil

    Science.gov (United States)

    Kuzyakov, Yakov; Blagodatskaya, Evgenia

    2015-04-01

    increases in C stocks. Consequently, the intensification of fluxes is much stronger than the increase of pools. Maintenance of stoichiometric ratios by accelerated microbial growth in hotspots requires additional nutrients (e.g. N and P), causing their microbial mining from soil organic matter, i.e. priming effects. Consequently, priming effects are localized in microbial hotspots and are consequences of hot moments. Finally, we estimated the contribution of the hotspots to the whole soil profile and suggested that, irrespective of their volume, the hotspots are mainly responsible for the ecologically relevant processes in soil.

  5. 1968 Listing of Swimming Pool Equipment.

    Science.gov (United States)

    National Sanitation Foundation, Ann Arbor, MI. Testing Lab.

    An up-to-date listing of swimming pool equipment including--(1) companies authorized to display the National Sanitation Foundation seal of approval, (2) equipment listed as meeting NSF swimming pool equipment standards relating to diatomite type filters, (3) equipment listed as meeting NSF swimming pool equipment standard relating to sand type…

  6. Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors

    International Nuclear Information System (INIS)

    2002-06-01

    All prototype, demonstration and commercial liquid metal cooled fast reactors (LMFRs) have used liquid sodium as a coolant. Sodium cooled systems, operating at low pressure, are characterised by very large thermal margins relative to the coolant boiling temperature and a very low structural material corrosion rate. In spite of the negligible thermal energy stored in the liquid sodium available for release in case of leakage, there is some safety concern because of its chemical reactivity with respect to air and water. Lead, lead-bismuth or other alloys of lead, appear to eliminate these concerns because the chemical reactivity of these coolants with respect to air and water is very low. Some experts believe that conceptually, these systems could be attractive if high corrosion activity inherent in lead, long term materials compatibility and other problems will be resolved. Extensive research and development work is required to meet this goal. Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Federation in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. The need to exchange information on alternative fast reactor coolants was a major consideration in the recommendation by the Technical Working Group on Fast Reactors (TWGFRs) to collect, review and document the information on lead and lead-bismuth alloy coolants: technology, thermohydraulics, physical and chemical properties, as well as to make an assessment and comparison with respective sodium characteristics

  7. Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2000-05-01

    Worldwide there is considerable experience in nuclear power technology, especially in water cooled reactor technology. Of the operating plants, in September 1998, 346 were light water reactors (LWRs) totalling 306 GW(e) and 29 were heavy water reactors (HWRs) totalling 15 GW(e). The accumulated experience and lessons learned from these plants are being incorporated into new advanced reactor designs. Utility requirements documents have been formulated to guide these design activities by incorporating this experience, and results from research and development programmes, with the aim of reducing costs and licensing uncertainties by establishing the technical bases for the new designs. Common goals for advanced designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. Large water cooled reactors with power outputs of 1300 MW(e) and above, which possess inherent safety characteristics (e.g. negative Doppler moderator temperature coefficients, and negative moderator void coefficient) and incorporate proven, active engineered systems to accomplish safety functions are being developed. Other designs with power outputs from, for example, 220 MW(e) up to about 1300 MW(e) which also possess inherent safety characteristics and which place more emphasis on utilization of passive safety systems are being developed. Passive systems are based on natural forces and phenomena such as natural convection and gravity, making safety functions less dependent on active systems and components like pumps and diesel generators. In some cases, further experimental tests for the thermohydraulic conditions of interest in advanced designs can provide improved understanding of the phenomena. Further, analytical methods to predict reactor thermohydraulic behaviour can be qualified for use by comparison with the experimental results. These activities should ultimately result in more economical designs. The

  8. Clinical blood pool MR Imaging

    Energy Technology Data Exchange (ETDEWEB)

    Leiner, Tim [Maastrich University Medical Center (Netherlands). Dept. of Radiology; Goyen, Martin [University Medical Center Hamburg-Eppendorf (Germany); Rohrer, Mathias [Bayer Schering Pharma AG, Berlin (Germany). European Business Unit Diagnostic Imaging; Schoenberg, Stefan O. (eds.) [University Hospital Mannheim Medical Faculty Mannheim - Heidelberg Univ. (Germany). Dept. of Clinical Radiology and Nuclear Medicine

    2008-07-01

    Clinical Blood Pool MR Imaging - This excellent treatise on Vasovist {sup registered} created by a team of exceptional faculty who are pioneers in MR Angiography covers the basic techniques, safety, efficacy, image processing and pharmaco-economic details to successfully implement a new level of MRA image quality with this new contrast agent. Martin Prince, Cornell University, New York The editors and authors have made groundbreaking contributions towards establishing MR angiography in various investigative settings, rendering it more precise and applying it for diverse indications. The work presented here is founded upon the extensive experience of the editors, as well as the broad range of experience from other scientific working groups. Maximilian Reiser, Ludwig Maximilian University, Munich Vasovist {sup registered} (Gadofosveset), worldwide the first blood pool agent, has only recently become available for clinical use, but has already gained wide acceptance as a tool to improve magnetic resonance angiography. This book presents the first in-depth introduction to the basic physicochemical aspects of the agent, the application of Vasovist {sup registered} in clinical MRA, as well as potential clinical applications beyond MRA and patient management-related aspects. The first part of the book explains basic and technical properties of the agent and the differences of Vasovist {sup registered} compared to currently available extracellular agents. The second part contains detailed chapters on safety and efficacy. In the third part the focus is on MR angiographic applications, and in the fourth part of the book potential clinical fields beyond MRA are explored. All clinical chapters feature ready-to-use clinical protocols and a series of take home messages that concisely summarize the current role of blood pool imaging for each specific indication. (orig.)

  9. Solar thermal space heating combined with swimming pool heating: A promising solution for southern Europe climates

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, M.J.; Neves, Ana [INETI/DER, Lisboa (Portugal)

    2006-07-01

    The system concept evaluation performed focused on systems that can provide hot water, space heating and swimming-pool heating, and are designed for application in southern climates specifically for single-family houses. Due to the climate characteristics of southern Europe, space heating is required only for a few months in the year. In this evaluation it was considered a six month period for space heating and, on the other six months, swimming pool heating was considered. This type of systems are applicable to a niche market of people who are building their houses as single-family houses and want also to take profit of the good climate conditions for the use of solar energy. It is common that the construction of a swimming pool is also planned and constructed. The evaluation is made considering as reference system a factory made with 4m{sup 2} collector area and 300 l storage tank. The system in evaluation offers extra service - space heating and swimming pool heating and is formed by a collector field and a combistore providing solar hot water preparation and space heating in the winter period and providing also swimming pool heating in the summer period. The evaluation made shows that in southern Europe climates this system will give extra service in comparison to the traditional solar systems used and can be economically interesting.

  10. Method of measuring a liquid pool volume

    Science.gov (United States)

    Garcia, G.V.; Carlson, N.M.; Donaldson, A.D.

    1991-03-19

    A method of measuring a molten metal liquid pool volume and in particular molten titanium liquid pools is disclosed, including the steps of (a) generating an ultrasonic wave at the surface of the molten metal liquid pool, (b) shining a light on the surface of a molten metal liquid pool, (c) detecting a change in the frequency of light, (d) detecting an ultrasonic wave echo at the surface of the molten metal liquid pool, and (e) computing the volume of the molten metal liquid. 3 figures.

  11. Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility; Estudo termohidraulico de um elemento combustivel tipo MTR visando a construcao de um dispositivo de irradiacao

    Energy Technology Data Exchange (ETDEWEB)

    Coragem, Helio Boemer de Oliveira

    1980-07-01

    A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

  12. Thermohydraulic status and component behavior in the PWR during the selected meltdown scenario station blackout (SBO); Thermohydraulisches Verhalten und Komponentenverhalten eines DWR bei ausgewaehltem Kernschmelzszenarium infolge Station Blackout (SBO). Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Band, Sebastian; Blaesius, Christoph; Scheuerer, Martina; Steinroetter, Thomas

    2017-09-15

    The report on the thermohydraulic status and component behavior in the PWR during the selected meltdown scenario station blackout (SBO) includes the following issues: status of science and technology on this topic, analysis of a high-pressure meltdown scenario using ATHLET-CD for a German PWR starting from the initiating event station blackout, three-dimensional computational fluid dynamic (CFD) analyses of the pressurizer coolant loop in a generic German PWR, evaluation of the thermohydraulic steam generator behavior and its effect on the involved primary circuit components.

  13. Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility; Estudo termohidraulico de um elemento combustivel tipo MTR visando a construcao de um dispositivo de irradiacao

    Energy Technology Data Exchange (ETDEWEB)

    Coragem, Helio Boemer de Oliveira

    1980-07-01

    A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

  14. Large-scale pool fires

    Directory of Open Access Journals (Sweden)

    Steinhaus Thomas

    2007-01-01

    Full Text Available A review of research into the burning behavior of large pool fires and fuel spill fires is presented. The features which distinguish such fires from smaller pool fires are mainly associated with the fire dynamics at low source Froude numbers and the radiative interaction with the fire source. In hydrocarbon fires, higher soot levels at increased diameters result in radiation blockage effects around the perimeter of large fire plumes; this yields lower emissive powers and a drastic reduction in the radiative loss fraction; whilst there are simplifying factors with these phenomena, arising from the fact that soot yield can saturate, there are other complications deriving from the intermittency of the behavior, with luminous regions of efficient combustion appearing randomly in the outer surface of the fire according the turbulent fluctuations in the fire plume. Knowledge of the fluid flow instabilities, which lead to the formation of large eddies, is also key to understanding the behavior of large-scale fires. Here modeling tools can be effectively exploited in order to investigate the fluid flow phenomena, including RANS- and LES-based computational fluid dynamics codes. The latter are well-suited to representation of the turbulent motions, but a number of challenges remain with their practical application. Massively-parallel computational resources are likely to be necessary in order to be able to adequately address the complex coupled phenomena to the level of detail that is necessary.

  15. Structural integrity assessment of HANARO pool cover

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo

    2001-11-01

    This report is for the seismic analysis and the structural integrity evaluation of HANARO Pool Cover in accordances with the requirement of the Technical Specification for Seismic Analysis of HANARO Pool Cover. For performing the seismic analysis and evaluating the structural integrity for HANARO Pool Cover, the finite element analysis model using ANSYS 5.7 was developed and the dynamic characteristics were analyzed. The seismic response spectrum analyses of HANARO Pool Cover under the design floor response spectrum loads of OBE and SSE were performed. The analysis results show that the stress values in HANARO Pool Cover for the seismic loads are within the ASME Code limits. It is also confirmed that the fatigue usage factor is less than 1.0. Therefore any damage on structural integrity is not expected when an HANARO Pool Cover is installed in the upper part of the reactor pool

  16. Multifragmentation of hot nuclei

    International Nuclear Information System (INIS)

    Tamain, B.

    1990-10-01

    It is difficult to deposit a large amount (∼ 1 Gev) of excitation energy into a nucleus. And if one wants to deposit large excitation energy values, the best way consists of shooting a given target nucleus with several nucleons, which can be achieved by using intermediate energy (10-100 MeV/nucleon) heavy ions. Such very excited objects were named hot nuclei. The study of hot nuclei has been undertaken only for 7 years because intermediate energy heavy ion facilities were not available before. The game is then to determine the decay properties of such nuclei, their limits of existence. Their study is connected with general properties of nuclear matter: namely its equation of state. Of special interest, is the onset of a new decay mechanism: multifragmentation, which is the non-sequential disassembly of a hot nucleus into several light nuclei (often called intermediate-mass fragments or IMF) or particles. This paper, shows how this mechanism can reflect fundamental properties of nuclear matter, but also how its experimental signature is difficult to establish. Multifragmentation has also been studied by using very energetic projectiles (protons and heavy ions) in the relativistic or ultra-relativistic region. The multifragmentation question of hot nuclei is far from being solved. One knows that IMF production increases when the excitation energy brought into a system is strongly increased, but very little is known about the mechanisms involved and a clear onset for multifragmentation is not established

  17. Utilizing hot electrons

    Energy Technology Data Exchange (ETDEWEB)

    Nozik, Arthur J.

    2018-03-01

    In current solar cells, any photon energy exceeding the semiconductor bandgap is lost before being collected, limiting the cell performance. Hot carrier solar cells could avoid these losses. Now, a detailed experimental study and analysis shows that this strategy could lead to an improvement of the photoconversion efficiency in practice.

  18. Mechanical shielded hot cell

    International Nuclear Information System (INIS)

    Higgy, H.R.; Abdel-Rassoul, A.A.

    1983-01-01

    A plan to erect a mechanical shielded hot cell in the process hall of the Radiochemical Laboratory at Inchas is described. The hot cell is designed for safe handling of spent fuel bundles, from the Inchas reactor, and for dismantling and cutting the fuel rods in preparation for subsequent treatment. The biological shielding allows for the safe handling of a total radioactivity level up to 10,000 MeV-Ci. The hot cell consists of an α-tight stainless-steel box, connected to a γ-shielded SAS, through an air-lock containing a movable carriage. The α-box is tightly connected with six dry-storage cavities for adequate storage of the spent fuel bundles. Both the α-box, with the dry-storage cavities, and the SAS are surrounded by 200-mm thick biological lead shielding. The α-box is equipped with two master-slave manipulators, a lead-glass window, a monorail crane and Padirac and Minirag systems. The SAS is equipped with a lead-glass window, tong manipulator, a shielded pit and a mechanism for the entry of the spent fuel bundle. The hot cell is served by adequate ventilation and monitoring systems. (author)

  19. Hot Deformation Behavior of Hot-Extruded AA7175 Through Hot Torsion Tests.

    Science.gov (United States)

    Lee, Se-Yeon; Jung, Taek-Kyun; Son, Hyeon-Woo; Kim, Sang-Wook; Son, Kwang-Tae; Choi, Ho-Joon; Oh, Sang-Ho; Lee, Ji-Woon; Hyun, Soong-Keun

    2018-03-01

    The hot deformation behavior of hot-extruded AA7175 was investigated with flow curves and processing maps through hot torsion tests. The flow curves and the deformed microstructures revealed that dynamic recrystallization (DRX) occurred in the hot-extruded AA7175 during hot working. The failure strain was highest at medium temperature. This was mainly influenced by the dynamic precipitation of fine rod-shaped MgZn2. The processing map determined the optimal deformation condition for the alloy during hot working.

  20. Experimental investigation of effect of flow attack angle on thermohydraulic performance of air flow in a rectangular channel with discrete V-pattern baffle on the heated plate

    Directory of Open Access Journals (Sweden)

    Raj Kumar

    2016-05-01

    Full Text Available In this work, the effect of angle of attack ( α a of the discrete V-pattern baffle on thermohydraulic performance of rectangular channel has been studied experimentally. The baffle wall was constantly heated and the other three walls of the channel were kept insulated. The experimentations were conducted to collect the data on Nusselt number ( N u b and friction factor ( f b by varying the Reynolds number (Re = 3000–21,000 and angle of attack ( α a from 30° to 70°, for the kept values of relative baffle height ( H b / H = 0 . 50 , relative pitch ratio ( P b / H = 1 . 0 , relative discrete width ( g w / H b = 1 . 5 and relative discrete distance ( D d / L v = 0 . 67 . As compared to the smooth wall, the V-pattern baffle roughened channel enhances the Nusselt number ( N u b and friction factor ( f b by 4.2 and 5.9 times, respectively. The present discrete V-pattern baffle shapes with angle of attack ( α a of 60° equivalent to flow Reynolds number of 3000 yields the greatest thermohydraulic performance. Discrete V-pattern baffle has improved thermal performance as compared to other baffle shapes’ rectangular channel.

  1. Experimental study of core thermohydraulics in fast reactors during transition from forced to natural circulation. Influence of inter-wrapper flow

    International Nuclear Information System (INIS)

    Kamide, H.; Hayashi, K.; Momoi, K.

    1997-01-01

    The evaluation of core thermohydraulics under natural circulation conditions is important to utilize inherent safety features of Fast Reactors. When heat exchangers of a decay heat removal system are operated in an upper plenum of reactor vessel, cold sodium is provided by the heat exchangers. Core-plenum interactions will occur during a natural circulation condition due to this cold sodium in the upper plenum, e.g., it can penetrate into gap regions between fuel subassemblies (inter-wrapper flow, IWF) and the flow may reverse in low power core channels. These interactions will significantly modify the flow and temperature distributions in the core. Sodium experiments were carried out to study these phenomena. In a test section, seven subassemblies are housed and connected to an upper plenum. The influences of core-plenum interactions on the core thermohydraulics were investigated under steady state conditions and also in transitions from forced to natural circulation. Cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop due to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core. (author)

  2. Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code

    International Nuclear Information System (INIS)

    Martin, L.; Saenz Tejada, P.

    1993-01-01

    When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

  3. Livermore pool-type reactor

    International Nuclear Information System (INIS)

    Mann, L.G.

    1977-01-01

    The Livermore Pool-Type Reactor (LPTR) has served a dual purpose since 1958--as an instrument for fundamental research and as a tool for measurement and calibration. Our early efforts centered on neutron-diffraction, fission, and capture gamma-ray studies. During the 1960's it was used for extensive calibration work associated with radiochemical and physical measurements on nuclear-explosive tests. Since 1970 the principal applications have been for trace-element measurements and radiation-damage studies. Today's research program is dominated by radiochemical studies of the shorter-lived fission products and by research on the mechanisms of radiation damage. Trace-element measurement for the National Uranium Resource Evaluation (NURE) program is the major measurement application today

  4. Cardiac blood pool emission tomography

    International Nuclear Information System (INIS)

    Itti, R.; Philippe, L.; Lorgeron, J.M.; Charbonnier, B.; Raynaud, P.; Brochier, M.

    1983-01-01

    After blood pool labeling using technetium-99m, a series of cardiac pictures is acquired during the rotation of a gamma-camera about the patient. Computer processing leads to reconstruction of various tomographic slices from the original planar projection. Electrocardiographic gating selects the different phases of the cardiac cycle. Individual slices through the left ventricular region are added in order to provide ''thick'' slices on which global and regional parameters of the left ventricular function can be determined. Due to the proportionality existing between count rates and labeled blood volumes, any geometrical model can be avoided. The delineation of regions of interest for count integration is made easier due to the absence of superimposition of structures; no correction for background is necessary. Tomography thus appears to be more consistent and more accurate than the classical methods using planar projections. In addition, right ventricular morphological and kinetic studies can be performed in the same conditions as for the left ventricle [fr

  5. Arc-weld pool interactions

    International Nuclear Information System (INIS)

    Glickstein, S.S.

    1978-08-01

    The mechanisms involved in arc-weld pool interactions are extremely complex and no complete theory is presently available to describe much of the phenomena observed during welding. For the past several years, experimental and analytical studies have been undertaken at the Bettis Atomic Power Laboratory to increase basic understanding of the gas tungsten arc welding process. These studies have included experimental spectral analysis of the arc in order to determine arc temperature and analytical modeling of the arc and weld puddle. The investigations have been directed toward determining the cause and effects of variations in the energy distribution incident upon the weldment. In addition, the effect of weld puddle distortion on weld penetration was investigated, and experimental and analytical studies of weld process variables have been undertaken to determine the effects of the variables upon weld penetration and configuration. A review of the results and analysis of these studies are presented

  6. Software Simulation of Hot Tearing

    DEFF Research Database (Denmark)

    Andersen, S.; Hansen, P.N.; Hattel, Jesper Henri

    1999-01-01

    The brittleness of a solidifying alloy in a temperature range near the solidus temperature has been recognised since the fifties as the mechanism responsible for hot tearing. Due to this brittlenes, the metal will crack under even small amounts of strain in that temperature range. We see these hot...... tears in castings close to hot centres, where the level of strain is often too high.Although the hot tearing mechanism is well understood, until now it has been difficult to do much to reduce the hot tearing tendency in a casting. In the seventies, good hot tearing criteria were developed by considering...... the solidification rate and the strain rate of the hot tear prone areas. But, until recently it was only possible to simulate the solidification rate, so that the criteria could not be used effectively.Today, with new software developments, it is possible to also simulate the strain rate in the hot tear prone areas...

  7. Hot Fuel Examination Facility (HFEF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Hot Fuel Examination Facility (HFEF) is one of the largest hot cells dedicated to radioactive materials research at Idaho National Laboratory (INL). The nation's...

  8. Hot subluminous star: HDE 283048

    International Nuclear Information System (INIS)

    Laget, M.; Vuillemin, A.; Parsons, S.B.; Henize, K.G.; Wray, J.D.

    1978-01-01

    The star HDE 283048, located at α = 3/sup h/50/sup m/.3, delta = +25 0 36', shows a strong ultraviolet continuum. Ground-based observations indicate a hot-dominated composite spectrum. Several lines of evidence suggest that the hot component is a hot subdwarf. 2 figures

  9. Hot chocolate effect

    International Nuclear Information System (INIS)

    Crawford, F.S.

    1982-01-01

    The ''hot chocolate effect'' was investigated quantitatively, using water. If a tall glass cylinder is filled nearly completely with water and tapped on the bottom with a softened mallet one can detect the lowest longitudinal mode of the water column, for which the height of the water column is one-quarter wavelength. If the cylinder is rapidly filled with hot tap water containing dissolved air the pitch of that mode may descend by nearly three octaves during the first few seconds as the air comes out of solution and forms bubbles. Then the pitch gradually rises as the bubbles float to the top. A simple theoretical expression for the pitch ratio is derived and compared with experiment. The agreement is good to within the 10% accuracy of the experiments

  10. Hot water reticulation

    Energy Technology Data Exchange (ETDEWEB)

    Fellows, S. K.

    1977-10-15

    Hot water reticulation (district heating) is an established method of energy supply within cities in many countries. It is based on the fact that heat can often be obtained cheaply in bulk, and that the resultant savings can, in suitable circumstances, justify the investment in a reticulation network of insulated pipes to distribute the heat to many consumers in the form of hot water or occasionally steam. The heat can be used by domestic, commercial, and industrial consumers for space heating and water heating, and by industries for process heat. The costs of supplying domestic consumers can be determined by considering an average residential area, but industrial and commercial consumers are so varied in their requirements that every proposal must be treated independently. Fixed costs, variable costs, total costs, and demand and resource constraints are discussed.

  11. The hot chocolate effect

    Science.gov (United States)

    Crawford, Frank S.

    1982-05-01

    The ''hot chocolate effect'' was investigated quantitatively, using water. If a tall glass cylinder is filled nearly completely with water and tapped on the bottom with a softened mallet one can detect the lowest longitudinal mode of the water column, for which the height of the water column is one-quarter wavelength. If the cylinder is rapidly filled with hot tap water containing dissolved air the pitch of that mode may descend by nearly three octaves during the first few seconds as the air comes out of solution and forms bubbles. Then the pitch gradually rises as the bubbles float to the top. A simple theoretical expression for the pitch ratio is derived and compared with experiment. The agreement is good to within the 10% accuracy of the experiments.

  12. Hot air balloon engine

    Energy Technology Data Exchange (ETDEWEB)

    Edmonds, Ian [Solartran Pty Ltd, 12 Lentara Street, Kenmore, Brisbane 4069 (Australia)

    2009-04-15

    This paper describes a solar powered reciprocating engine based on the use of a tethered hot air balloon fuelled by hot air from a glazed collector. The basic theory of the balloon engine is derived and used to predict the performance of engines in the 10 kW to 1 MW range. The engine can operate over several thousand metres altitude with thermal efficiencies higher than 5%. The engine thermal efficiency compares favorably with the efficiency of other engines, such as solar updraft towers, that also utilize the atmospheric temperature gradient but are limited by technical constraints to operate over a much lower altitude range. The increased efficiency allows the use of smaller area glazed collectors. Preliminary cost estimates suggest a lower $/W installation cost than equivalent power output tower engines. (author)

  13. The ''hot'' patella

    International Nuclear Information System (INIS)

    Kipper, M.S.; Alazraki, N.P.; Feiglin, D.H.

    1982-01-01

    Increased patellar uptake on bone scans is seen quite commonly but the possible or probable etiologies of this finding have not been previously well described. A review of 100 consecutive bone scans showed that the incidence of bilateral ''hot'' patellae is 15%. Identified etiologies include osteoarthritic degenerative disease (35%), fracture, possible metastatic disease, bursitis, Paget's disease, and osteomyelitis. The value of careful history, physical examination, and radiographs is stressed

  14. Hot nuclei and fragmentation

    International Nuclear Information System (INIS)

    Guerreau, D.

    1993-01-01

    A review is made of the present status concerning the production of nuclei above 5 MeV temperature. Considerable progress has been made recently on the understanding of the formation and the fate of such hot nuclei. It appears that the nucleus seems more stable against temperature than predicted by static calculations. However, the occurrence of multifragment production at high excitation energies is now well established. The various experimental features of the fragmentation process are discussed. (author) 59 refs., 12 figs

  15. 'Hot particle' intercomparison dosimetry

    International Nuclear Information System (INIS)

    Kaurin, D.G.L.; Baum, J.W.; Charles, M.W.; Darley, D.P.J.; Durham, J.S.; Scannell, M.J.; Soares, C.G.

    1996-01-01

    Dosimetry measurements of four 'hot particles' were made at different density thickness values using five different methods. The hot particles had maximum dimensions of 650 μm and maximum beta energies of 0.97, 046, 0.36, and 0.32 MeV. Absorbers were used to obtain the dose at different depths for each dosimeter. Measurements were made using exoelectron dosimeters, an extrapolation chamber, NE Extremity Tape Dosimeters (tm), Eberline RO-2 and RO-2A survey meters, and two sets of GafChromic (tm) dye film with each set read out at a different institution. From these results the dose was calculated averaged over 1 cm 2 of tissue at 18, 70, 125, and 400 μm depth. Comparisons of tissue-dose averaged over 1 cm 2 for 18, 70, and 125 μm depth based on interpolated measured values, were within 30% for the GafChromic (tm) dye film, extrapolation chamber, NE Extremity Tape Dosimeters (tm), and Eberline RO-2 and 2A (tm) survey meters except for the hot particle with 0.46 MeV maximum beta energy. The results for this source showed differences of up to 60%. The extrapolation chamber and NE Extremity Tape dosimeters under-responded for measurements at 400 μm by about a factor of 2 compared with the GafChromic dye films for two hot particles with maximum beta energy of 0.32 and 0.36 MeV which each emitted two 100% 1 MeV photons per disintegration. Tissue doses determined using exoelectron dosimeters were a factor of 2 to 5 less than those determined using other dosimeters, possibly due to failures of the equipment. (author)

  16. Factors driving semi-aquatic predator occurrence in traditional cattle drinking pools: conservation issues

    Directory of Open Access Journals (Sweden)

    Raoul Manenti

    2016-08-01

    Full Text Available In several cases, human impact on water bodies and on their freshwater communities is detrimental, but in some cases the human activity may favour and enhance the biodiversity of small water bodies, as traditional cattle drinking pools. Despite their small size, small water bodies may constitute hot spot of biodiversity often representing the only lentic aquatic biotope in landscapes where superficial water lacks or flows in lotic environments like creeks and streams. Predators are good indicators of biodiversity in ponds and give information of food chain web complexity. In particular, semi-aquatic predators like amphibians and dragonflies may account for a substantial percentage of energy flow between aquatic and terrestrial ecosystems. In this study, we evaluated the conservation value of traditional cattle drinking pools building by assessing the factors determining the occurrence and distribution of the semi-aquatic predators. From April to August 2015, we investigated 30 distinct pools recording several abiotic and biotic environmental variables. We detected 4 semi-aquatic predators: Salamandra salamandra larvae, Triturus carnifex, Aeshna sp. larvae and Libellula sp. larvae. Abiotic features played a major role in shaping the predator community that resulted linked to stable, with no dryness period, and large drinking pools. Invertebrate prey biomass was not particularly important, while vegetation cover and occurrence of unpalatable tadpoles were the most important biotic features of the pools. Our study provides novel evidence on the importance of cattle drinking pools management to preserve biodiversity especially in areas where traditional pastoral activity is disappearing.

  17. Solar Hot Water Heater

    Science.gov (United States)

    1978-01-01

    The solar panels pictured below, mounted on a Moscow, Idaho home, are part of a domestic hot water heating system capable of providing up to 100 percent of home or small business hot water needs. Produced by Lennox Industries Inc., Marshalltown, Iowa, the panels are commercial versions of a collector co-developed by NASA. In an effort to conserve energy, NASA has installed solar collectors at a number of its own facilities and is conducting research to develop the most efficient systems. Lewis Research Center teamed with Honeywell Inc., Minneapolis, Minnesota to develop the flat plate collector shown. Key to the collector's efficiency is black chrome coating on the plate developed for use on spacecraft solar cells, the coating prevents sun heat from "reradiating," or escaping outward. The design proved the most effective heat absorber among 23 different types of collectors evaluated in a Lewis test program. The Lennox solar domestic hot water heating system has three main components: the array of collectors, a "solar module" (blue unit pictured) and a conventional water heater. A fluid-ethylene glycol and water-is circulated through the collectors to absorb solar heat. The fluid is then piped to a double-walled jacket around a water tank within the solar module.

  18. Guide for decontaminating swimming pool at schools

    International Nuclear Information System (INIS)

    Matsuhashi, Shimpei; Kurikami, Hiroshi; Yasuda, Ryo; Takano, Takao; Seko, Noriaki; Naganawa, Hirochika; Kuroki, Ryota; Saegusa, Jun

    2012-07-01

    Because of TEPCO Fukushima Dai-ichi Nuclear Power Plant accident due to the Great East Japan Earthquake, a huge amount of radioactive materials was widely dispersed and precipitated into the environment. Swimming pools in Fukushima prefectures were contaminated with the radioactives. We JAEA carried out several demonstration tests to decontaminate the radioactives and discharge the pool water safely. We concluded the results obtained from the tests as 'Guide for decontaminating Swimming Pool at School' and released it quickly. Following this, we also released the guide in English. This manuscript, as an experimental report of the swimming pool water decontamination, is consisted from the guide in Japanese and English prepared. (author)

  19. Current operating practices of nuclear insurance pools

    International Nuclear Information System (INIS)

    O'Connell, J.M.

    1993-01-01

    This paper discusses the nuclear pooling system and co-operation between the pools, present practice and capacity, with a breakdown of the limits for third party liability and material damage. The author also describes the relationship between the pools and the nuclear operators (the policyholders), and concludes that the nuclear pools have been successful in serving the interests of their member companies, their policyholders and the governments as they have provided a stable insurance market by making available capacity in amounts that had never before been assembled and placed at risk in a single location. 2 tabs

  20. Guide for decontaminating swimming pool at schools

    Energy Technology Data Exchange (ETDEWEB)

    Matsuhashi, Shimpei; Kurikami, Hiroshi; Yasuda, Ryo; Takano, Takao; Seko, Noriaki; Naganawa, Hirochika; Kuroki, Ryota; Saegusa, Jun

    2012-07-15

    Because of TEPCO Fukushima Dai-ichi Nuclear Power Plant accident due to the Great East Japan Earthquake, a huge amount of radioactive materials was widely dispersed and precipitated into the environment. Swimming pools in Fukushima prefectures were contaminated with the radioactives. We JAEA carried out several demonstration tests to decontaminate the radioactives and discharge the pool water safely. We concluded the results obtained from the tests as 'Guide for decontaminating Swimming Pool at School' and released it quickly. Following this, we also released the guide in English. This manuscript, as an experimental report of the swimming pool water decontamination, is consisted from the guide in Japanese and English prepared. (author)

  1. Structure for nuclear fuel storage pools

    International Nuclear Information System (INIS)

    Ebata, Sakae; Nichiei, Shinji.

    1979-01-01

    Purpose: To enable leak detection in nuclear fuel storage pools, as well as prevent external leakages while keeping the strength of the constructional structures. Constitution: Protection plates are provided around pool linear plates and a leak reception is provided to the bottom. Leakages are detected by leak detecting pipeways and the external leakages are prevented by collecting them in a detection area provided in the intermediate layer. Since ferro-reinforcements at the bottom wall of the pool are disconnected by the protection plate making it impossible to form the constructional body, body hunches are provided to the bottom wall of the pool for processing the ferro-reinforcements. (Yoshino, Y.)

  2. Heat transfer from internally heated hemispherical pools

    International Nuclear Information System (INIS)

    Gabor, J.D.; Ellsion, P.G.; Cassulo, J.C.

    1980-01-01

    Experiments were conducted on heat transfer from internally heated ZnSO 4 -H 2 O pools to the walls of hemispherical containers. This experimental technique provides data for a heat transfer system that has to date been only theoretically treated. Three different sizes of copper hemispherical containers were used: 240, 280, 320 mm in diameter. The pool container served both as a heat transfer surface and as an electrode. The opposing electrode was a copper disk, 50 mm in diameter located at the top of the pool in the center. The top surface of the pool was open to the atmosphere

  3. Oxidation kinetics of corium pool

    Energy Technology Data Exchange (ETDEWEB)

    Sulatsky, A.A., E-mail: andrei314@mail.ru [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Smirnov, S.A. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (NIIEFA), St. Petersburg (Russian Federation); Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Kotova, S.Yu. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Fischer, M.; Hellmann, S. [AREVA NP GmbH, Erlangen (Germany); Tromm, W.; Miassoedov, A. [Forschungzentrum Karlsruhe (FZK), Karlsruhe (Germany); Bottomley, D. [EUROPÄISCHE KOMMISSION, Joint Research Centre Institut für Transurane (ITU), Karlsruhe (Germany); Piluso, P. [CEA Cadarache-DEN/DTN/STRI, St.Paul-lez-Durance (France); Barrachin, M. [Institut de Radioprotection et Sûreté Nucléaire, St.Paul-lez-Durance (France)

    2013-09-15

    Highlights: • The analysis of experimental data on molten corium oxidation was been carried out. • The analysis has revealed the main factors influencing the oxidation kinetics. • The analysis was used for developing a qualitative analytical model. • The numerical modeling has confirmed the results of experimental data analysis. -- Abstract: Experimental, theoretical and numerical studies of oxidation kinetics of an open surface corium pool have been reported. The experiments have been carried out within OECD MASCA program and ISTC METCOR, METCOR-P and EVAN projects. It has been shown that the melt oxidation is controlled by an oxidant supply to the melt free surface from the atmosphere, not by the reducer supply from the melt. The project experiments have not detected any input of the zirconium oxidation kinetics into the process chemistry. The completed analysis puts forward a simple analytical model, which gives an explanation of the main features of melt oxidation process. The numerical modeling results are in good agreement with experimental data and theoretical considerations.

  4. Oxidation kinetics of corium pool

    International Nuclear Information System (INIS)

    Sulatsky, A.A.; Smirnov, S.A.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Kotova, S.Yu.; Fischer, M.; Hellmann, S.; Tromm, W.; Miassoedov, A.; Bottomley, D.; Piluso, P.; Barrachin, M.

    2013-01-01

    Highlights: • The analysis of experimental data on molten corium oxidation was been carried out. • The analysis has revealed the main factors influencing the oxidation kinetics. • The analysis was used for developing a qualitative analytical model. • The numerical modeling has confirmed the results of experimental data analysis. -- Abstract: Experimental, theoretical and numerical studies of oxidation kinetics of an open surface corium pool have been reported. The experiments have been carried out within OECD MASCA program and ISTC METCOR, METCOR-P and EVAN projects. It has been shown that the melt oxidation is controlled by an oxidant supply to the melt free surface from the atmosphere, not by the reducer supply from the melt. The project experiments have not detected any input of the zirconium oxidation kinetics into the process chemistry. The completed analysis puts forward a simple analytical model, which gives an explanation of the main features of melt oxidation process. The numerical modeling results are in good agreement with experimental data and theoretical considerations

  5. Consideration on the 1 ton bucket elevator installed under water of pool in IMEF

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ung Sup; Lee, Jong Heon; Lee, Hong Gi; Choo, Yong Sun; Jung, Yang Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    The bucket elevator which can transfer examination capsule or fuel from pool to hot cell is installed under water of pool (3x6x10 m) in IMEF. A allowable load is 1 ton and the dimension of bucket is 25x25x150 cm. The upper and lower sides motion of bucket have about 63 degrees inclined duties with a chain driving system. A specialized made chain catches a rug of bucket with a roller sliding way between right and left guide rails which are fixed at inner of a square tube and moves to an upper and lower sides and it is made so that it is operated by sprocket wheel installed in a hot cell working table below. Sprocket wheel is executed to two steps of driving shaft by reduction geared motor installed at right outside of M1 hot cell. As for the starting operation, it is executed by push an operation button on a operating panel located at front of M1 hot cell.

  6. 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

  7. 13 CFR 120.1708 - Pool Certificates.

    Science.gov (United States)

    2010-01-01

    ... United States. (c) SBA purchase of a Loan Interest. SBA will determine whether to purchase a Loan... reserves the right to purchase a Loan Interest from a Pool at any time. (d) Self-liquidating. A Pool... market conditions and program experience, and will publish any such change in the Federal Register. (h...

  8. Compact fuel storage rack for fuel pools

    International Nuclear Information System (INIS)

    Parras, F.; Louvat, J.P.

    1986-01-01

    ETS LEMER and FRAMATOME propose a new compact storage rack. This rack permits a considerable increase of the storage capacity of cooling pools. A short description of the structure and the components is presented, to propose racks that are: . Inalterable, . Compact, . Insensitive to earthquakes. Installation in pools already in operation is simplified by their light structure and the bearing device [fr

  9. The warm pool in the Indian Ocean

    Digital Repository Service at National Institute of Oceanography (India)

    Vinayachandran, P.N.; Shetye, S.R.

    is larger and warmer, a peculiarity of the pool in the Indian Ocean is its seasonal variation. The surface area of the pool changes from 24 x 106 km2 in April to 8 x 106 km2 in September due to interaction with the southwest monsoon. The annual cycles of sea...

  10. Shedding Light on Dark Liquidity Pools

    NARCIS (Netherlands)

    Degryse, H.A.; van Achter, M.; Wuyts, G.

    2008-01-01

    This article reviews the recent developments on dark liquidity pools starting from the theoretical and empirical academic literature. The number of dark liquidity pools as well as their trading volume has grown substantially in the last couple of years. We highlight the incentives of providers as

  11. Seminar on experimental thermohydraulics

    International Nuclear Information System (INIS)

    Katsaounis, A.

    1979-06-01

    Considerations on reactor safety are made. Problems related to the project and assembling of facilities for experiments in steady state and transient conditions are discussed. The advantages of using model fluids are finally, analysed. (Author) [pt

  12. Big city consultants shut down our pool : a shocking community pool gets checked for stray voltage

    Energy Technology Data Exchange (ETDEWEB)

    Lynch, P. [Power Line Systems Engineering Inc., Markham, ON (Canada)

    2009-12-15

    This article discussed an investigation conducted at a community pool where swimmers complained of receiving electrical shocks both in the pool and on the pool's deck area. Electrical measurements taken at the pool revealed current flows from the pool water to various points around the deck area. Measured current flow in the pool area was 30 amps even when the main pool service breaker was opened to shut off power to the entire facility. Thirty amps of primary neutral current was then measured on the primary side aerial neutral in front of the pool. A 10 amp primary feeder from the pool joined up with the complex's primary neutral wire to increase the neutral current to 40 amps. The combined 40 amps current then returned to the secondary side of a nearby utility transformer substation. The study showed that the underground wet low-resistance grounded surface area of the pool was attracting the 30 amps of utility current from the surrounding ground area. The local utility disconnected the primary and secondary neutral interconnection at the pool's main 600-volt step-down transformer. The pool deck was removed in order to install additional copper bonding grounds. In order to avert serious injuries, many experts propose that all electric utilities should be required by law to reconfigure their power systems to prevent primary power neutral currents from entering private buildings. 1 tab., 2 figs.

  13. Ripples in a superconducting tidal pool

    CERN Document Server

    Hudson, E

    2003-01-01

    The behaviour of electrons in a metal is often compared to that of water in a pool. An empty pool is like a material that has all of its electrons removed. As electrons are 'poured' into the metal, they first occupy the lowest energies available - the bottom of the pool - and eventually fill up to the Fermi energy, the top of the pool. At this point we no longer discuss electrons but quasiparticles. These are electrons that have modified properties due to their interactions within the material. Waves in a pool can be excited, and their properties will depend on the depth of the water. Similarly in a metal, quasiparticles behave like waves that have a material-dependent dispersion relation between their energy and their wavevector, which specifies their direction and wavelength. This simple analogy also hints at an indirect method of measuring the dispersion relation of a metal, and hence the myriad of properties that depend on it. (U.K.)

  14. Emerging hot spot analysis

    DEFF Research Database (Denmark)

    Reinau, Kristian Hegner

    Traditionally, focus in the transport field, both politically and scientifically, has been on private cars and public transport. Freight transport has been a neglected topic. Recent years has seen an increased focus upon congestion as a core issue across Europe, resulting in a great need for know...... speed data for freight. Secondly, the analytical methods used, space-time cubes and emerging hot spot analysis, are also new in the freight transport field. The analysis thus estimates precisely how fast freight moves on the roads in Northern Jutland and how this has evolved over time....

  15. Progress in hot pressing

    International Nuclear Information System (INIS)

    Brodhag, C.; Thevenot, F.

    1988-01-01

    An experimental technique is described to study hot pressing of ceramics under conditions of controlled temperature and pressure during both the heating and final sintering stages. This method gives a better control of the final microstructure of the material. Transformation mechanisms can be studied during initial heating stage (impurity degasing, reaction, phase transformation, mechanical behavior of intergranular phase...) using computer control and graphical data representations. Some examples will be given for different systems studied in our laboratory: B (α, β, amorphous), B 12 O 2 (reaction of B + B 2 O 3 ), Si 3 N 4 ( + additives), TiN, Al 2 O 3 + AlON,ZrC

  16. Multipurpose reprocessing hot cell

    International Nuclear Information System (INIS)

    Fletcher, R.D.

    1975-01-01

    A multipurpose hot cell is being designed for use at the Idaho Chemical Processing Plant for handling future scheduled fuels that cannot be adequately handled by the existing facilities and equipment. In addition to providing considerable flexibility to handle a wide variety of fuel sizes up to 2,500 lb in weight the design will provide for remote maintenance or replacement of the in-cell equipment with a minimum of exposure to personnel and also provide process piping connections for custom processing of small quantities of fuel. (auth)

  17. Design and Construction of Pool Door for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Kwangsub; Lee, Sangjin; Choi, Jinbok; Oh, Jinho; Lee, Jongmin [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The pool door is a structure to isolate the reactor pool from the service pool for maintenance. The pool door is installed before the reactor pool is drained. The pool door consists of structural component and sealing component. The main structures of the pool door are stainless steel plates and side frames. The plates and frames are assembled by welded joints. Lug is welded at the top of the plate. The pool door is submerged in the pool water when it is used. Materials of the pool door should be resistive to corrosion and radiation. Stainless steel is used in structural components and air nozzle assemblies. Features of design and construction of the pool door for the research reactor are introduced. The pool door is designed to isolate the reactor pool for maintenance. Structural analysis is performed to evaluate the structural integrity during earthquake. Tests and inspections are also carried out during construction to identify the safety and function of the pool door.

  18. Design and Construction of Pool Door for Research Reactor

    International Nuclear Information System (INIS)

    Jung, Kwangsub; Lee, Sangjin; Choi, Jinbok; Oh, Jinho; Lee, Jongmin

    2016-01-01

    The pool door is a structure to isolate the reactor pool from the service pool for maintenance. The pool door is installed before the reactor pool is drained. The pool door consists of structural component and sealing component. The main structures of the pool door are stainless steel plates and side frames. The plates and frames are assembled by welded joints. Lug is welded at the top of the plate. The pool door is submerged in the pool water when it is used. Materials of the pool door should be resistive to corrosion and radiation. Stainless steel is used in structural components and air nozzle assemblies. Features of design and construction of the pool door for the research reactor are introduced. The pool door is designed to isolate the reactor pool for maintenance. Structural analysis is performed to evaluate the structural integrity during earthquake. Tests and inspections are also carried out during construction to identify the safety and function of the pool door

  19. Pool scrubbing models for iodine components

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, K [Battelle Ingenieurtechnik GmbH, Eschborn (Germany)

    1996-12-01

    Pool scrubbing is an important mechanism to retain radioactive fission products from being carried into the containment atmosphere or into the secondary piping system. A number of models and computer codes has been developed to predict the retention of aerosols and fission product vapours that are released from the core and injected into water pools of BWR and PWR type reactors during severe accidents. Important codes in this field are BUSCA, SPARC and SUPRA. The present paper summarizes the models for scrubbing of gaseous Iodine components in these codes, discusses the experimental validation, and gives an assessment of the state of knowledge reached and the open questions which persist. The retention of gaseous Iodine components is modelled by the various codes in a very heterogeneous manner. Differences show up in the chemical species considered, the treatment of mass transfer boundary layers on the gaseous and liquid sides, the gas-liquid interface geometry, calculation of equilibrium concentrations and numerical procedures. Especially important is the determination of the pool water pH value. This value is affected by basic aerosols deposited in the water, e.g. Cesium and Rubidium compounds. A consistent model requires a mass balance of these compounds in the pool, thus effectively coupling the pool scrubbing phenomena of aerosols and gaseous Iodine species. Since the water pool conditions are also affected by drainage flow of condensate water from different regions in the containment, and desorption of dissolved gases on the pool surface is determined by the gas concentrations above the pool, some basic limitations of specialized pool scrubbing codes are given. The paper draws conclusions about the necessity of coupling between containment thermal-hydraulics and pool scrubbing models, and proposes ways of further simulation model development in order to improve source term predictions. (author) 2 tabs., refs.

  20. Pool scrubbing models for iodine components

    International Nuclear Information System (INIS)

    Fischer, K.

    1996-01-01

    Pool scrubbing is an important mechanism to retain radioactive fission products from being carried into the containment atmosphere or into the secondary piping system. A number of models and computer codes has been developed to predict the retention of aerosols and fission product vapours that are released from the core and injected into water pools of BWR and PWR type reactors during severe accidents. Important codes in this field are BUSCA, SPARC and SUPRA. The present paper summarizes the models for scrubbing of gaseous Iodine components in these codes, discusses the experimental validation, and gives an assessment of the state of knowledge reached and the open questions which persist. The retention of gaseous Iodine components is modelled by the various codes in a very heterogeneous manner. Differences show up in the chemical species considered, the treatment of mass transfer boundary layers on the gaseous and liquid sides, the gas-liquid interface geometry, calculation of equilibrium concentrations and numerical procedures. Especially important is the determination of the pool water pH value. This value is affected by basic aerosols deposited in the water, e.g. Cesium and Rubidium compounds. A consistent model requires a mass balance of these compounds in the pool, thus effectively coupling the pool scrubbing phenomena of aerosols and gaseous Iodine species. Since the water pool conditions are also affected by drainage flow of condensate water from different regions in the containment, and desorption of dissolved gases on the pool surface is determined by the gas concentrations above the pool, some basic limitations of specialized pool scrubbing codes are given. The paper draws conclusions about the necessity of coupling between containment thermal-hydraulics and pool scrubbing models, and proposes ways of further simulation model development in order to improve source term predictions. (author) 2 tabs., refs

  1. The InterFrost benchmark of Thermo-Hydraulic codes for cold regions hydrology - first inter-comparison results

    Science.gov (United States)

    Grenier, Christophe; Roux, Nicolas; Anbergen, Hauke; Collier, Nathaniel; Costard, Francois; Ferrry, Michel; Frampton, Andrew; Frederick, Jennifer; Holmen, Johan; Jost, Anne; Kokh, Samuel; Kurylyk, Barret; McKenzie, Jeffrey; Molson, John; Orgogozo, Laurent; Rivière, Agnès; Rühaak, Wolfram; Selroos, Jan-Olof; Therrien, René; Vidstrand, Patrik

    2015-04-01

    The impacts of climate change in boreal regions has received considerable attention recently due to the warming trends that have been experienced in recent decades and are expected to intensify in the future. Large portions of these regions, corresponding to permafrost areas, are covered by water bodies (lakes, rivers) that interact with the surrounding permafrost. For example, the thermal state of the surrounding soil influences the energy and water budget of the surface water bodies. Also, these water bodies generate taliks (unfrozen zones below) that disturb the thermal regimes of permafrost and may play a key role in the context of climate change. Recent field studies and modeling exercises indicate that a fully coupled 2D or 3D Thermo-Hydraulic (TH) approach is required to understand and model the past and future evolution of landscapes, rivers, lakes and associated groundwater systems in a changing climate. However, there is presently a paucity of 3D numerical studies of permafrost thaw and associated hydrological changes, and the lack of study can be partly attributed to the difficulty in verifying multi-dimensional results produced by numerical models. Numerical approaches can only be validated against analytical solutions for a purely thermic 1D equation with phase change (e.g. Neumann, Lunardini). When it comes to the coupled TH system (coupling two highly non-linear equations), the only possible approach is to compare the results from different codes to provided test cases and/or to have controlled experiments for validation. Such inter-code comparisons can propel discussions to try to improve code performances. A benchmark exercise was initialized in 2014 with a kick-off meeting in Paris in November. Participants from USA, Canada, Germany, Sweden and France convened, representing altogether 13 simulation codes. The benchmark exercises consist of several test cases inspired by existing literature (e.g. McKenzie et al., 2007) as well as new ones. They

  2. Thermohydraulic modeling of the dry air passive containment cooling system process in the Westinghouse AP-600 ALWR

    Energy Technology Data Exchange (ETDEWEB)

    Harari, R; Weis, Y; Barnea, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    Following postulated events of a LOCA, the passive Containment Cooling System (PCCS) uses dry air to transfer the residual heat by natural circulation. The air flow path, designed between the steel reactor containment hot shell and the concrete shield building, creates an open thermosyphon. The purpose of this inherently safe process is to assure the long term steady-state cooling of the nuclear core after an emergency shutdown (authors).

  3. Thermohydraulics of a horizontal diphasic flow of superfluid helium; Thermo-hydraulique d'un ecoulement horizontal d'helium superfluide diphasique

    Energy Technology Data Exchange (ETDEWEB)

    Perraud, S

    2007-12-15

    This study aims at characterizing helium two phase flows, and to identify the dependence of their characteristics on various thermo-hydraulic parameters: vapour velocity, liquid height, vapour density, specificities of superfluidity. Both the engineer and the physicist's points of view are taken into consideration: the first one in terms of optimization of a particular cooling scheme based on a two-phase flow, and these second one in terms of more fundamental atomization-related questions. It has been shown that for velocities around 3 to 4 m/s, the liquid phase that was initially stratified undergoes an atomization through the presence of a drop haze carried by the vapor phase.This happens for superfluid helium as well as for normal helium without main differences on atomization.

  4. Residential solar hot water

    Energy Technology Data Exchange (ETDEWEB)

    1982-06-01

    This report examines the feasibility of using solar energy to preheat domestic water coming from the city supply at a temperature of approximately 4{degree}C. Four solar collectors totalling 7 m{sup 2} were installed on a support structure facing south at an angle of 60{degree} from the horizontal. The system worked most efficiently in the spring and early summer when the combination of long hours of sunshine, clean air and clear skies allowed for maximum availability of solar radiation. Performance dropped in late summer and fall mainly due to cloudier weather conditions. The average temperature in the storage tank over the 10 months of operation was 42{degree}C, ranging from a high of 83{degree}C in July to a low of 6{degree}C in November. The system provided a total of 7.1 GJ, which is approximately one-third the annual requirement for domestic hot water heating. At the present time domestic use of solar energy to heat water does not appear to be economically viable. High capital costs are the main problem. As a solar system with present day technology can only be expected to meet half to two-thirds of the hot water energy demand the savings are not sufficient for the system to pay for itself within a few years. 5 figs.

  5. Effects of surface shape on the geometry and surface topography of the melt pool in low-power density laser melting

    KAUST Repository

    Kim, Youngdeuk

    2011-04-15

    The quantitative correlations between workpiece volume and melt pool geometry, as well as the flow and thermal features of the melt pool are established. Thermocapillary convections in melt pool with a deformable free surface are investigated with respect to surface shape and laser intensity. When the contact angle between the tangent to the top surface and the vertical wall at the hot center is acute, the free surface flattens, compared with that of the initial free surface. Otherwise, the free surface forms a bowl-like shape with a deep crater and a low peripheral rim when the contact angle at the hot center is obtuse. Increasing the workpiece volume at a fixed laser intensity and a negative radial height gradient cause linear decreases in the geometric size and magnitude of flow and temperature of the melt pool. Conversely, linear increases are observed with a positive radial height gradient. © 2011 American Institute of Chemical Engineers (AIChE).

  6. Modifications in Compacted MX-80 Bentonite Due to Thermo-Hydraulic Treatment; Modificaciones en la Bentonita MX-80 Compactada Sometida a Tratamiento Termo-Hidraulico

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Espina, R.; Villar, M. V.

    2013-09-01

    The thermo-hydraulic tests reproduce the thermal and hydraulic conditions to which bentonite is subjected in the engineered barrier of a deep geological repository of radioactive waste. The results of thermo-hydraulic test TBT1500, which was running for approximately 1500 days, are presented. This is a continuation to the Technical Report Ciemat 1199, which presented results of test TBT500, performed under similar conditions but with duration of 500 days. In both tests the MX-80 bentonite was used with initial density and water content similar to those of the large-scale test TBT. The bentonite column was heated at the bottom at 140 degree centigrade and hydrated on top with deionized water. At the end of the test a sharp water content gradient was observed along the column, as well as an inverse dry density gradient. Hydration modified also the bentonite microstructure. Besides, an overall decrease of the smectite content with respect to the initial value took place, especially in the most hydrated areas where the percentage of interest ratified illite increased and in the longer test. On the other hand, the content of cristobalite, feldspars and calcite increased. Smectite dissolution processes (probably colloidal) occurred, particularly in the more hydrated areas and in the longer test. Due to the dissolution of low-solubility species and to the loss of exchangeable positions in the smectite, the content of soluble salts in the pore water increased with respect to the original one, especially in the longer test. The solubilized ions were transported; sodium, calcium, magnesium and sulphate having a similar mobility, which was in turn lower than that of potassium and chloride. The cationic exchange complex was also modified. (Author)

  7. 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)

  8. Swimming pool hydraulics and their significance for public pools. Bedeutung der Beckenhydraulik in oeffentlichen Schwimmbaedern

    Energy Technology Data Exchange (ETDEWEB)

    Gansloser, G

    1989-11-01

    The term of swimming pool hydraulics means the process of letting in and drawing off water to and from the pool while ensuring that no inadmissible water-borne contaminant concentrations will occur anywhere within the pool. Measurements were performed on a pool to study the significance of correct pool hydraulics. The author points out that a wrong water recirculation design will bring to nought the effects of an elaborate water treatment system; by contrast, poor pool water quality can be greatly improved by redesigning the pool water hydraulics approach. In principle, systems with with water inlet at one side and water outlet at the far side will fall short of hygienic requirements. (BWI).

  9. Spent fuel pool cleanup and stabilization

    International Nuclear Information System (INIS)

    Miller, R.L.

    1987-06-01

    Each of the plutonium production reactors at Hanford had a large water-filled spent fuel pool to provide interim storage of irradiated fuel while awaiting shipment to the separation facilities. After cessation of reactor operations the fuel was removed from the pools and the water levels were drawn down to a 5- to 10-foot depth. The pools were maintained with the water to provide shielding and radiological control. What appeared to be a straightforward project to process the water, remove the sediments from the basin, and stabilize the contamination on the floors and walls became a very complex and time consuming operation. The sediment characteristics varied from pool to pool, the ion exchange system required modification, areas of hard-pack sediments were discovered on the floors, special arrangements to handle and package high dose rate items for shipment were required, and contract problems ensued with the subcontractor. The original schedule to complete the project from preliminary engineering to final stabilization of the pools was 15 months. The actual time required was about 25 months. The original cost estimate to perform the work was $2,651,000. The actual cost of the project was $5,120,000, which included $150,000 for payment of claims to the subcontractor. This paper summarizes the experiences associated with the cleanup and radiological stabilization of the 100-B, -C, -D, and -DR spent fuel pools, and discusses a number of lessons learned items

  10. Hot springs in Hokuriku District

    Energy Technology Data Exchange (ETDEWEB)

    Sato, K. (Hot Springs Research Center, Japan)

    1971-01-01

    In the Hokuriku district including Toyama, Ishikawa, and Fukui Prefectures, hot springs of more than 25/sup 0/C were investigated. In the Toyama Prefecture, there are 14 hot springs which are located in an area from the Kurobe River to the Tateyama volcano and in the mountainous area in the southwest. In Ishikawa Prefecture there are 16 hot springs scattered in Hakusan and its vicinity, the Kaga mountains, and in the Noto peninsula. In northern Fukui Prefecture there are seven hot springs. The hot springs in Shirakawa in Gifu Prefecture are characterized as acid springs producing exhalations and H/sub 2/S. These are attributed to the Quaternary volcanoes. The hot springs of Wakura, Katayamazu, and Awara in Ishikawa Prefecture are characterized by a high Cl content which is related to Tertiary andesite. The hot springs of Daishoji, Yamanaka, Yamashiro, Kuritsu, Tatsunokuchi, Yuwaku, and Yunotani are characterized by a low HCO/sub 3/ content. The Ca and SO/sub 4/ content decreases from east to west, and the Na and Cl content increases from west to east. These fluctuations are related to the Tertiary tuff and rhyolite. The hot springs of Kuronagi, Kinshu, and Babadani, located along the Kurobe River are characterized by low levels of dissolved components and high CO/sub 2/ and HCO/sub 3/ content. These trends are related to late Paleozoic granite. Hot springs resources are considered to be connected to geothermal resources. Ten tables, graphs, and maps are provided.

  11. Multifunctional Hot Structure Heat Shield

    Data.gov (United States)

    National Aeronautics and Space Administration — This project is performing preliminary development of a Multifunctional Hot Structure (HOST) heat shield for planetary entry. Results of this development will...

  12. Distributed Technologies in a Data Pool

    Science.gov (United States)

    Keiser, K.; Conover, H.; Graves, S.; He, Y.; Regner, K.; Smith, M.

    2004-12-01

    A Data Pool is an on-line repository providing interactive and programmatic access to data products through a variety of services. The University of Alabama in Huntsville has developed and deployed such a Data Pool in conjunction with the DISCOVER project, a collaboration with NASA and Remote Sensing Systems. DISCOVER provides long-term ocean and climate data from a variety of passive microwave satellite instruments, including such products as sea-surface temperature and wind, air temperature, atmospheric water vapor, cloud water and rain rate. The Data Pool provides multiple methods to access and visualize these products, including conventional HTTP and FTP access, as well as data services that provide for enhanced usability and interoperability, such as GridFTP, OPeNDAP, OpenGIS-compliant web mapping and coverage services, and custom subsetting and packaging services. This paper will focus on the distributed service technologies used in the Data Pool, which spans heterogeneous machines at multiple locations. For example, in order to provide seamless access to data at multiple sites, the Data Pool provides catalog services for all data products at the various data server locations. Under development is an automated metadata generation tool that crawls the online data repositories regularly to dynamically update the Data Pool catalog with information about newly generated data files. For efficient handling of data orders across distributed repositories, the Data Pool also implements distributed data processing services on the file servers where the data resides. Ontologies are planned to support automated service chaining for custom user requests. The UAH Data Pool is based on a configurable technology framework that integrates distributed data services with a web interface and a set of centralized database services for catalogs and order tracking. While this instantiation of the Data Pool was implemented to meet the needs of the DISCOVER project, the framework was

  13. A Pool of Distant Galaxies

    Science.gov (United States)

    2008-11-01

    Anyone who has wondered what it might be like to dive into a pool of millions of distant galaxies of different shapes and colours, will enjoy the latest image released by ESO. Obtained in part with the Very Large Telescope, the image is the deepest ground-based U-band image of the Universe ever obtained. It contains more than 27 million pixels and is the result of 55 hours of observations with the VIMOS instrument. A Sea of Galaxies ESO PR Photo 39/08 A Pool of Distant Galaxies This uniquely beautiful patchwork image, with its myriad of brightly coloured galaxies, shows the Chandra Deep Field South (CDF-S), arguably the most observed and best studied region in the entire sky. The CDF-S is one of the two regions selected as part of the Great Observatories Origins Deep Survey (GOODS), an effort of the worldwide astronomical community that unites the deepest observations from ground- and space-based facilities at all wavelengths from X-ray to radio. Its primary purpose is to provide astronomers with the most sensitive census of the distant Universe to assist in their study of the formation and evolution of galaxies. The new image released by ESO combines data obtained with the VIMOS instrument in the U- and R-bands, as well as data obtained in the B-band with the Wide-Field Imager (WFI) attached to the 2.2 m MPG/ESO telescope at La Silla, in the framework of the GABODS survey. The newly released U-band image - the result of 40 hours of staring at the same region of the sky and just made ready by the GOODS team - is the deepest image ever taken from the ground in this wavelength domain. At these depths, the sky is almost completely covered by galaxies, each one, like our own galaxy, the Milky Way, home of hundreds of billions of stars. Galaxies were detected that are a billion times fainter than the unaided eye can see and over a range of colours not directly observable by the eye. This deep image has been essential to the discovery of a large number of new galaxies

  14. Modal analysis of pool door in water tank

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Soo; Jeong, Kyeong Hoon; Park, Chan Gook; Koo, In Soo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    A pool door is installed at the chase of the pool gate by means of an overhead crane in the building of a research reactor. The principal function of the pool door, which is located between the reactor pool and service pool, is to separate the reactor pool from the service pool for the maintenance and/or the removal of the equipment either in the reactor pool or service pool. The pool door consists of stainless steel plates supported by structural steel frames and sealing components. The pool door is equipped with double inflatable gaskets. The configuration of the pool door is shown in Figure 1. The FEM analysis and theoretical calculation by the formula were performed to evaluate the natural frequency for the pool door in the water. The results from the two methods were compared.

  15. The Role of Nuclear Insurance Pools

    International Nuclear Information System (INIS)

    Reitsma, S. M. S.

    2006-01-01

    Since fifty years insurers respond to the need of both governments and the electricity industry to provide financial protection to cover the perils presented by the use of nuclear energy for peaceful purposes. This paper aims at explaining what difficulties had to be solved in order to enable insurers to provide such protection, that as a solution to these difficulties Nuclear Insurance Pools were formed, how such pools operate and what security they provide. Thereby not only a number of universal principles underlying nuclear pool insurance will be explained, but also some differences in the characteristics of such insurance per group of countries. (author)

  16. Laser surveillance systems for fuel storage pools

    International Nuclear Information System (INIS)

    Boeck, H.

    1985-06-01

    A Laser Surveillance System (LASSY) as a new safeguards device has been developed under the IAEA research contract No. 3458/RB at the Atominstitut Wien using earlier results by S. Fiarman. This system is designed to act as a sheet of light covering spent fuel assemblies in spent fuel storage pools. When movement of assemblies takes place, LASSY detects and locates the position of the movement in the pool and when interrogated, presents a list of pool positions and times of movement to the safeguards inspector. A complete prototype system was developed and built. Full scale tests showed the principal working capabilities of a LASSY underwater

  17. Convection in molten pool created by a concentrated energy flux on a solid metal target

    International Nuclear Information System (INIS)

    Dikshit, B.; Zende, G. R.; Bhatia, M. S.; Suri, B. M.

    2009-01-01

    During surface evaporation of metals by use of a concentrated energy flux such as electron beam or lasers, a liquid metal pool having a very high temperature gradient is formed around the hot zone created by the beam. Due to temperature dependence of surface tension, density, and depression of the evaporating surface caused by back pressure of the emitted vapor in this molten pool, a strong convective current sets in the molten pool. A proposition is made that this convection may pass through three different stages during increase in the electron beam power depending upon dominance of the various driving forces. To confirm this, convective heat transfer is quantified in terms of dimensionless Nusselt number and its evolution with power is studied in an experiment using aluminum, copper, and zirconium as targets. These experimentally determined values are also compared to the theoretical values predicted by earlier researchers to test the validity of their assumptions and to know about the type of flow in the melt pool. Thus, conclusion about the physical characteristics of flow in the molten pool of metals could be drawn by considering the roles of surface tension and curvature of the evaporating surface on the evolution of convective heat transfer.

  18. 10 CFR 36.63 - Pool water purity.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Pool water purity. 36.63 Section 36.63 Energy NUCLEAR... § 36.63 Pool water purity. (a) Pool water purification system must be run sufficiently to maintain the conductivity of the pool water below 20 microsiemens per centimeter under normal circumstances. If pool water...

  19. Hot Gas Halos in Galaxies

    Science.gov (United States)

    Mulchaey, John

    Most galaxy formation models predict that massive low-redshift disk galaxies are embedded in extended hot halos of externally accreted gas. Such gas appears necessary to maintain ongoing star formation in isolated spirals like the Milky Way. To explain the large population of red galaxies in rich groups and clusters, most galaxy evolution models assume that these hot gas halos are stripped completely when a galaxy enters a denser environment. This simple model has been remarkably successful at reproducing many observed properties of galaxies. Although theoretical arguments suggest hot gas halos are an important component in galaxies, we know very little about this gas from an observational standpoint. In fact, previous observations have failed to detect soft X-ray emission from such halos in disk galaxies. Furthermore, the assumption that hot gas halos are stripped completely when a galaxy enters a group or cluster has not been verified. We propose to combine proprietary and archival XMM-Newton observations of galaxies in the field, groups and clusters to study how hot gas halos are impacted by environment. Our proposed program has three components: 1) The deepest search to date for a hot gas halo in a quiescent spiral galaxy. A detection will confirm a basic tenet of disk galaxy formation models, whereas a non-detection will seriously challenge these models and impose new constraints on the growth mode and feedback history of disk galaxies. 2) A detailed study of the hot gas halos properties of field early-type galaxies. As environmental processes such as stripping are not expected to be important in the field, a study of hot gas halos in this environment will allow us to better understand how feedback and other internal processes impact hot gas halos. 3) A study of hot gas halos in the outskirts of groups and clusters. By comparing observations with our suite of simulations we can begin to understand what role the stripping of hot gas halos plays in galaxy

  20. The Nuclear Insurance Pools: Operations and Covers

    International Nuclear Information System (INIS)

    Tetley, M.

    2008-01-01

    Nuclear insurance pools have provided insurance for the nuclear industry for over fifty years and it is fair to say that the development of civil nuclear power would not have been possible without the support of the commercial insurance market. The unknown risks presented by the nascent nuclear power industry in the 1950s required a leap of faith by insurers who developed specialist pooled insurance capacity to ensure adequate capacity to back up the operators' compensation obligations. Since then, nuclear insurance pools have evolved to become comprehensive suppliers of most types of insurance for nuclear plant globally. This paper will outline the structure, development, products and current operations of nuclear insurance pools.(author)

  1. Ingestion of swimming pool water by recreational

    Data.gov (United States)

    U.S. Environmental Protection Agency — Swimming pool water ingestion data. This dataset is associated with the following publication: Dufour, A., L. Wymer, M. Magnuson, T. Behymer, and R. Cantu. Ingestion...

  2. Nuclear Insurance Pools: Worldwide Practice and Development

    International Nuclear Information System (INIS)

    Reitsma, S. M. S.

    1998-01-01

    The development of nuclear installations to produce electricity led to the establishment of Nuclear Insurance Pools and the introduction of international Conventions on Third Party Liability. Nuclear Pools offer both Third Party Liability insurance, reflecting the Conventions' principles, and other insurance products. They are market-wide, providing a facility for participation by insurers who could not otherwise write the insurance for the particularly sensitive nuclear risk. All acceptances are for the net retention of each Member without recourse to individual reinsurance protection. Common account reinsurance is arranged with other Nuclear Pools all over the world. Thus, a transparency is created, which ensures the highest degree of reinsurance security and imposes a known finite limit to each participating insurer's commitment. Therefore, Pool-members are prepared to make a greater commitment to nuclear risks than would be case where they felt uncertain as regards their total exposure following a significant loss. (author)

  3. AE/VCE Unconfirmed Vernal Pools

    Data.gov (United States)

    Vermont Center for Geographic Information — This dataset is derived from a project by the Vermont Center for Ecostudies(VCE) and Arrowwood Environmental(AE) to map vernal pools throughout the state of Vermont....

  4. AE/VCE Confirmed Vernal Pools

    Data.gov (United States)

    Vermont Center for Geographic Information — This dataset is derived from a project by the Vermont Center for Ecostudies(VCE) and Arrowwood Environmental(AE) to map vernal pools throughout the state of Vermont....

  5. Investigations in Marine Chemistry: Tide Pool Ecology.

    Science.gov (United States)

    Schlenker, Richard M.

    Students investigated the salinity of tide pools at different levels in the intertidal zone. Data are analyzed collectively. Students graphed and discussed data. Included are suggestions for evaluation and further study. (Author)

  6. 7 CFR 1126.7 - Pool plant.

    Science.gov (United States)

    2010-01-01

    ... excluded from the supply plant's shipments in computing the plant's shipping percentage. (d) A plant... part, or the plant has automatic pooling status under the other Federal order; and (7) That portion of...

  7. 7 CFR 1001.7 - Pool plant.

    Science.gov (United States)

    2010-01-01

    ... excluded from the supply plant's shipments in computing the percentages in paragraphs (c)(1) and (2) of... plant has automatic pooling status under the other Federal order. [64 FR 47954, Sept. 1, 1999, as...

  8. 7 CFR 1007.7 - Pool plant.

    Science.gov (United States)

    2010-01-01

    ...-upon use other than Class I shall be excluded from the supply plant's shipments in computing the plant... under the order in this part, or such plant has automatic pooling status under such other order. (h) Any...

  9. Microbial quality of a marine tidal pool

    CSIR Research Space (South Africa)

    Genthe, Bettina

    1995-01-01

    Full Text Available In this study the source of microbial pollution to a tidal pool was investigated. Both adjacent seawater which could contribute to possible faecal pollution and potential direct bather pollution were studied. The microbial quality of the marine...

  10. Robotic cleaning of a spent fuel pool

    International Nuclear Information System (INIS)

    Roman, H.T.; Marian, F.A.; Silverman, E.B.; Barkley, V.P.

    1987-01-01

    Spent fuel pools at nuclear power plants are not cleaned routinely, other than by purifying the water that they contain. Yet, debris can collect on the bottom of a pool and should be removed prior to fuel transfer. At Public Service Electric and Gas Company's Hope Creek Nuclear Power Plant, a submersible mobile robot - ARD Corporation's SCAVENGER - was used to clean the bottom of the spent fuel pool prior to initial fuel loading. The robotic device was operated remotely (as opposed to autonomously) with a simple forward/reverse control, and it cleaned 70-80% of the pool bottom. This paper reports that a simple cost-benefit analysis shows that the robotic device would be less expensive, on a per mission basis, than other cleaning alternatives, especially if it were used for other similar cleaning operations throughout the plant

  11. Experimental study on heat pipe heat removal capacity for passive cooling of spent fuel pool

    International Nuclear Information System (INIS)

    Xiong, Zhenqin; Wang, Minglu; Gu, Hanyang; Ye, Cheng

    2015-01-01

    Highlights: • A passively cooling SFP heat pipe with an 8.2 m high evaporator was tested. • Heat removed by the heat pipe is in the range of 3.1–16.8 kW. • The heat transfer coefficient of the evaporator is 214–414 W/m 2 /K. • The heat pipe performance is sensitive to the hot water temperature. - Abstract: A loop-type heat pipe system uses natural flow with no electrically driven components. Therefore, such a system was proposed to passively cool spent fuel pools during accidents to improve nuclear power station safety especially for station blackouts such as those in Fukushima. The heat pipe used for a spent fuel pool is large due to the spent fuel pool size. An experimental heat pipe test loop was developed to estimate its heat removal capacity from the spent fuel pool during an accident. The 7.6 m high evaporator is heated by hot water flowing vertically down in an assistant tube with a 207-mm inner diameter. R134a was used as the potential heat pipe working fluid. The liquid R134a level was 3.6 m. The tests were performed for water velocities from 0.7 to 2.1 × 10 −2 m/s with water temperatures from 50 to 90 °C and air velocities from 0.5 m/s to 2.5 m/s. The results indicate significant heat is removed by the heat pipe under conditions that may occur in the spent fuel pool

  12. Pooling strategies for St Petersburg gamblers

    OpenAIRE

    Csörgö, Sandor; Simons, Gordon

    2006-01-01

    Peter offers to play exactly one St Petersburg game with each of [math] players, Paul [math] , [math] , Paul [math] , whose conceivable pooling strategies are described by all possible probability distributions [math] . Comparing infinite expectations, we characterize among all [math] those admissible strategies for which the pooled winnings, each distributed as [math] , yield a finite added value for each and every one of Paul [math] , [math] , Paul [math] in comparison with their individual...

  13. Welding pool measurement using thermal array sensor

    Science.gov (United States)

    Cho, Chia-Hung; Hsieh, Yi-Chen; Chen, Hsin-Yi

    2015-08-01

    Selective laser melting (SLM) is an additive manufacturing (AM) technology that uses a high-power laser beam to melt metal powder in chamber of inert gas. The process starts by slicing the 3D CAD data as a digital information source into layers to create a 2D image of each layer. Melting pool was formed by using laser irradiation on metal powders which then solidified to consolidated structure. In a selective laser melting process, the variation of melt pool affects the yield of a printed three-dimensional product. For three dimensional parts, the border conditions of the conductive heat transport have a very large influence on the melt pool dimensions. Therefore, melting pool is an important behavior that affects the final quality of the 3D object. To meet the temperature and geometry of the melting pool for monitoring in additive manufacturing technology. In this paper, we proposed the temperature sensing system which is composed of infrared photodiode, high speed camera, band-pass filter, dichroic beam splitter and focus lens. Since the infrared photodiode and high speed camera look at the process through the 2D galvanometer scanner and f-theta lens, the temperature sensing system can be used to observe the melting pool at any time, regardless of the movement of the laser spot. In order to obtain a wide temperature detecting range, 500 °C to 2500 °C, the radiation from the melting pool to be measured is filtered into a plurality of radiation portions, and since the intensity ratio distribution of the radiation portions is calculated by using black-body radiation. The experimental result shows that the system is suitable for melting pool to measure temperature.

  14. Hot cell verification facility update

    International Nuclear Information System (INIS)

    Titzler, P.A.; Moffett, S.D.; Lerch, R.E.

    1985-01-01

    The Hot Cell Verification Facility (HCVF) provides a prototypic hot cell mockup to check equipment for functional and remote operation, and provides actual hands-on training for operators. The facility arrangement is flexible and assists in solving potential problems in a nonradioactive environment. HCVF has been in operation for six years, and the facility is a part of the Hanford Engineering Development Laboratory

  15. Numerical modelling of methanol liquid pool fires

    Science.gov (United States)

    Prasad, Kuldeep; Li, Chiping; Kailasanath, K.; Ndubizu, Chuka; Ananth, Ramagopal; Tatem, P. A.

    1999-12-01

    The focus of this paper is on numerical modelling of methanol liquid pool fires. A mathematical model is first developed to describe the evaporation and burning of a two-dimensional or axisymmetric pool containing pure liquid methanol. Then, the complete set of unsteady, compressible Navier-Stokes equations for reactive flows are solved in the gas phase to describe the convection of the fuel gases away from the pool surface, diffusion of the gases into the surrounding air and the oxidation of the fuel into product species. Heat transfer into the liquid pool and the metal container through conduction, convection and radiation are modelled by solving a modified form of the energy equation. Clausius-Clapeyron relationships are invoked to model the evaporation rate of a two-dimensional pool of pure liquid methanol. The governing equations along with appropriate boundary and interface conditions are solved using the flux-corrected transport algorithm. Numerical results exhibit a flame structure that compares well with experimental observations. Temperature profiles and burning rates were found to compare favourably with experimental data from single- and three-compartment laboratory burners. The model predicts a puffing frequency of approximately 12 Hz for a 1 cm diameter methanol pool in the absence of any air co-flow. It is also observed that increasing the air co-flow velocity helps in stabilizing the diffusion flame, by pushing the vortical structures away from the flame region.

  16. Recent advances in probabilistic species pool delineations

    Directory of Open Access Journals (Sweden)

    Dirk Nikolaus Karger

    2016-07-01

    Full Text Available A species pool is the set of species that could potentially colonize and establish within a community. It has been a commonly used concept in biogeography since the early days of MacArthur and Wilson’s work on Island Biogeography. Despite their simple and appealing definition, an operational application of species pools is bundled with a multitude of problems, which have often resulted in arbitrary decisions and workarounds when defining species pools. Two recently published papers address the operational problems of species pool delineations, and show ways of delineating them in a probabilistic fashion. In both papers, species pools were delineated using a process-based, mechanistical approach, which opens the door for a multitude of new applications in biogeography. Such applications include detecting the hidden signature of biotic interactions, disentangling the geographical structure of community assembly processes, and incorporating a temporal extent into species pools. Although similar in their conclusions, both ‘probabilistic approaches’ differ in their implementation and definitions. Here I give a brief overview of the differences and similarities of both approaches, and identify the challenges and advantages in their application.

  17. Sodium aerosol formation in an argon flow over hot sodium

    International Nuclear Information System (INIS)

    Clement, C.F.; Dolias, M.J.; UKAEA Atomic Energy Research Establishment, Harwell. Thermal Hydraulics Div.)

    1987-01-01

    Vapour evaporation, which partly forms aerosol, occurs when a cold gas flows over a hot liquid. A previous well-mixed model is extended to predict the final vapour plus aerosol content of such a flow in terms of its initial and final temperatures. The predictions are compared to results of the Copacabana II experiment in which argon passed over a sodium pool. Agreement is obtained for the final sodium density at moderate flow rates, and physical reasons are given as to why deviations occur at low and high flow rates. (author)

  18. Mental disease-related emergency admissions attributable to hot temperatures.

    Science.gov (United States)

    Lee, Suji; Lee, Hwanhee; Myung, Woojae; Kim, E Jin; Kim, Ho

    2018-03-01

    The association between high temperature and mental disease has been the focus of several studies worldwide. However, no studies have focused on the mental disease burden attributable to hot temperature. Here, we aim to quantify the risk attributed to hot temperatures based on the exposure-lag-response relationship between temperature and mental diseases. From data on daily temperature and emergency admissions (EA) for mental diseases collected from 6 major cities (Seoul, Incheon, Daejeon, Daegu, Busan, and Gwangju in South Korea) over a period of 11years (2003-2013), we estimated temperature-disease associations using a distributed lag non-linear model, and we pooled the data by city through multivariate meta-analysis. Cumulative relative risk and attributable risks were calculated for extreme hot temperatures, defined as the 99th percentile relative to the 50th percentile of temperatures. The strongest association between mental disease and high temperature was seen within a period of 0-4days of high temperature exposure. Our results reveal that 14.6% of EA for mental disease were due to extreme hot temperatures, and the elderly were more susceptible (19.1%). Specific mental diseases, including anxiety, dementia, schizophrenia, and depression, also showed significant risk attributed to hot temperatures. Of all EA for anxiety, 31.6% were attributed to extremely hot temperatures. High temperature was responsible for an attributable risk for mental disease, and the burden was higher in the elderly. This finding has important implications for designing appropriate public health policies to minimize the impact of high temperature on mental health. Copyright © 2017 Elsevier B.V. All rights reserved.

  19. Field observations and management strategy for hot spring wastewater in Wulai area, Taiwan.

    Science.gov (United States)

    Lin, J Y; Chen, C F; Lei, F R; Hsieh, C D

    2010-01-01

    Hot springs are important centers for recreation and tourism. However, the pollution that may potentially be caused by hot spring wastewater has rarely been discussed. More than half of Taiwan's hot springs are located in areas where the water quality of water bodies is to be protected, and untreated wastewater could pollute the receiving water bodies. In this study, we investigate hot spring wastewater in the Wulai area, one of Taiwan's famous hot spring resorts. Used water from five hot spring hotels was sampled and ten sampling events were carried out to evaluate the changes in the quality of used water in different seasons, at different periods of the week, and from different types of hotels. The concentrations of different pollutants in hot spring wastewater were found to exhibit wide variations, as follows: COD, 10-250 mg/L; SS, N.D.-93 mg/L; NH(3)-N, 0.01-1.93 mg/L; TP, 0.01-0.45 mg/L; and E. coli, 10-27,500 CFU/100 mL. The quality of hot spring wastewater depends on the operation of public pools, because this affects the frequency of supplementary fresh water and the outflow volume. Two management strategies, namely, onsite treatment systems and individually packaged treatment equipment, are considered, and a multi-objective optimization model is used to determine the optimal strategy.

  20. A study of some radioprotection apparatuses used in the case of pool reactors

    International Nuclear Information System (INIS)

    Robien, E. de; Choudens, H. de; Delpuech, J.

    1965-01-01

    Various problems of radioprotection concerning swimming-pool reactors in Grenoble have led us to study adequate solutions: a) The automatic verification of the staff-radioactivity when coming out of Melusine or Siloe has been realized thanks to a βγ gate which is insensitive to the ambient background in the reactor-hall; b) The automatic verification of the contamination of the shoes of the agents working in these reactors has been realized with a dedicated device; c) The necessity to measure precisely γ doses with the help of an autonomous apparatus has led to the making of a plastic-scintillator γ dosimeter; d) The obligation to forbid the opening of doors in some places where there might be a great intensity of radiation, has led us to make doors open according to the intensity of radiation inside the rooms; e) The releases of radioactive iodine have been measured with activated charcoal cartridges that surround a scintillator connected with a unique channel selector; f) Finally the control of reactor safety rod fall in case of a radioactive accident has been secured by a chain whose detector is a chamber immersed in the swimming-pool, which offers, in the particular case of the hot thickness swimming-pool reactor a double advantage: first it enables us to regulate the upper hot water layer, second to get free of transitory radiations which appear in the reactor hall as the experimental apparatuses are taken out from the core. (authors) [fr

  1. Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC; Desarrollo de un modelo del NSSS de un reactor PWR con el codigo termo-hidraulico GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Garcia-Torano, I.; Jimenez, G.

    2013-07-01

    The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

  2. Hot testing of coke

    Energy Technology Data Exchange (ETDEWEB)

    Balon, I D

    1976-07-01

    Earlier investigations failed to take full account of the factors affecting coke behavior within the blast furnace. An apparatus was accordingly developed for testing coke, based on a cyclone furnace where the sample could be held in a flow of hot oxidizing gases, simulating conditions in the blast furnace hearth. The results are said to be suitable for comprehensive assessment of the coke, including abrasive strength and its rate of gasification in a flow of carbon dioxide. Coke of size 6-10 mm tested at 1,100/sup 0/C in an atmosphere of oxidizing gases close to those obtaining in the blast furnace hearth, indicated that destruction and total gasification of the coke occurs after 5 minutes for a weak coke and 8 minutes for strong coke, depending on the physico-chemical and physico-mechanical properties of the particular coke. When samples were treated for a fixed period (3 minutes), the amount of coke remaining, and the percentage over 6 mm varied between 22 and 40 and between 4 and 7 percent respectively.

  3. Hot Hydrogen Test Facility

    International Nuclear Information System (INIS)

    W. David Swank

    2007-01-01

    The core in a nuclear thermal rocket will operate at high temperatures and in hydrogen. One of the important parameters in evaluating the performance of a nuclear thermal rocket is specific impulse, ISp. This quantity is proportional to the square root of the propellant's absolute temperature and inversely proportional to square root of its molecular weight. Therefore, high temperature hydrogen is a favored propellant of nuclear thermal rocket designers. Previous work has shown that one of the life-limiting phenomena for thermal rocket nuclear cores is mass loss of fuel to flowing hydrogen at high temperatures. The hot hydrogen test facility located at the Idaho National Lab (INL) is designed to test suitability of different core materials in 2500 C hydrogen flowing at 1500 liters per minute. The facility is intended to test non-uranium containing materials and therefore is particularly suited for testing potential cladding and coating materials. In this first installment the facility is described. Automated Data acquisition, flow and temperature control, vessel compatibility with various core geometries and overall capabilities are discussed

  4. HOT STARS WITH HOT JUPITERS HAVE HIGH OBLIQUITIES

    International Nuclear Information System (INIS)

    Winn, Joshua N.; Albrecht, Simon; Fabrycky, Daniel; Johnson, John Asher

    2010-01-01

    We show that stars with transiting planets for which the stellar obliquity is large are preferentially hot (T eff > 6250 K). This could explain why small obliquities were observed in the earliest measurements, which focused on relatively cool stars drawn from Doppler surveys, as opposed to hotter stars that emerged more recently from transit surveys. The observed trend could be due to differences in planet formation and migration around stars of varying mass. Alternatively, we speculate that hot-Jupiter systems begin with a wide range of obliquities, but the photospheres of cool stars realign with the orbits due to tidal dissipation in their convective zones, while hot stars cannot realign because of their thinner convective zones. This in turn would suggest that hot Jupiters originate from few-body gravitational dynamics and that disk migration plays at most a supporting role.

  5. Theory of hot particle stability

    International Nuclear Information System (INIS)

    Berk, H.L.; Wong, H.V.; Tsang, K.T.

    1986-10-01

    The investigation of stabilization of hot particle drift reversed systems to low frequency modes has been extended to arbitrary hot beta, β/sub H/ for systems that have unfavorable field line curvature. We consider steep profile equilibria where the thickness of the pressure drop, Δ, is less than plasma radius, r/sub p/. The analysis describes layer modes which have mΔ/r/sub p/ 2/3. When robust stability conditions are fulfilled, the hot particles will have their axial bounce frequency less than their grad-B drift frequency. This allows for a low bounce frequency expansion to describe the axial dependence of the magnetic compressional response

  6. Hot workability of aluminium alloys

    International Nuclear Information System (INIS)

    Yoo, Yeon Chul; Oh, Kyung Jin

    1986-01-01

    Hot Workability of aluminium alloys, 2024, 6061 and 7075, has been studied by hot torsion tests at temperatures from 320 to 515 deg C and at strain rates from 1.26 x 10 -3 to 5.71 x 10 -3 sec -1 . Hot working condition of these aluminium alloys was determined quantitatively from the constitutive equations obtained from flow stress curves in torsion. Experimental data of the logarith of the Zener-Hollomonn parameter showed good linear relationships to the logarith of sinh(ασ-bar)

  7. Self-formed waterfall plunge pools in homogeneous rock

    Science.gov (United States)

    Scheingross, Joel S.; Lo, Daniel Y.; Lamb, Michael P.

    2017-01-01

    Waterfalls are ubiquitous, and their upstream propagation can set the pace of landscape evolution, yet no experimental studies have examined waterfall plunge pool erosion in homogeneous rock. We performed laboratory experiments, using synthetic foam as a bedrock simulant, to produce self-formed waterfall plunge pools via particle impact abrasion. Plunge pool vertical incision exceeded lateral erosion by approximately tenfold until pools deepened to the point that the supplied sediment could not be evacuated and deposition armored the pool bedrock floor. Lateral erosion of plunge pool sidewalls continued after sediment deposition, but primarily at the downstream pool wall, which might lead to undermining of the plunge pool lip, sediment evacuation, and continued vertical pool floor incision in natural streams. Undercutting of the upstream pool wall was absent, and our results suggest that vertical drilling of successive plunge pools is a more efficient waterfall retreat mechanism than the classic model of headwall undercutting and collapse in homogeneous rock.

  8. Structure of pool in reactor building

    International Nuclear Information System (INIS)

    Yokoyama, Shigeki.

    1997-01-01

    Shielding walls made of iron-reinforced concrete having a metal liner including two body walls rigidly combined to the upper surface of a reactor container are disposed at least to one of an equipment pool or spent fuel storage pool in a reactor building. A rack for temporarily placing an upper lattice plate is detachably attached at least above one of a steam dryer or a gas/liquid separator temporarily placed in the temporary pool, and the height from the bottom portion to the upper end of the shielding wall is determined based on the height of an upper lattice plate temporary placed on the rack and the water depth required for shielding radiation from the upper lattice plate. An operator's exposure on the operation floor can be reduced by the shielding wall, and radiation dose from the spent fuels is reduced. The increase of the height of a pool guarder enhances bending resistance as a ceiling. In addition, the total height of them is made identical with the depth of the spent fuel storage pool thereby enabling to increase storage area for spent fuels. (N.H.)

  9. Cooling device for reactor suppression pool

    International Nuclear Information System (INIS)

    Togasaki, Susumu; Kato, Kiyoshi.

    1994-01-01

    In a cooling device of a reactor suppression pool, when a temperature of pool water is abnormally increased and a heat absorbing portion is heated by, for example, occurrence of an accident, coolants are sent to the outside of the reactor container to actuates a thermally operating portion by the heat energy of coolants and drive heat exchanging fluids of a secondary cooling system. If the heat exchanging fluids are sent to a cooling portion, the coolants are cooled and returned to the heat absorbing portion of the suppression pool water. If the heat absorbing portion is heat pipes, the coolants are evaporated by heat absorbed from the suppression pool water, steams are sent to the thermally operating portion, then coolants are liquefied and caused to return to the heat absorbing portion. If the thermal operation portion is a gas turbine, the gas turbine is operated by the coolants, and it is converted to a rotational force to drive heat exchanging fluids by pumps. By constituting the cooling portion with a condensator, the coolants are condensed and liquefied and returned to the heat absorbing portion of the suppression pool water. (N.H.)

  10. Condensation in a two-phase pool

    International Nuclear Information System (INIS)

    Duffey, R.B.; Hughes, E.D.

    1991-01-01

    We consider the case of vapor condensation in a liquid pool, when the heat transfer is controlled by heat losses through the walls. The analysis is based on drift flux theory for phase separation in the pool, and determines the two-phase mixture height for the pool. To our knowledge this is the first analytical treatment of this classic problem that gives an explicit result, previous work having established the result for the evaporative case. From conservation of mass and energy in a one-dimensional steady flow, together with a void relation between the liquid and vapor fluxes, we determine the increase in the mixture level from the base level of the pool. It can be seen that the thermal and hydrodynamic influences are separable. Thus, the thermal influence of the wall heat transfer appears through its effect on the condensing length L*, so that at high condensation rates the pool is all liquid, and at low rates overflows (the level swell or foaming effect). Similarly, the phase separation effect hydrodynamically determines the height via the relative velocity of the mixture to the entering flux. We examine some practical applications of this result to level swell in condensing flows, and also examine some limits in ideal cases

  11. 3. Workshop for IAEA ICSP on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents. Presentations

    International Nuclear Information System (INIS)

    2012-04-01

    Most advanced nuclear power plant designs adopted several kinds of passive systems. Natural circulation is used as a key driving force for many passive systems and even for core heat removal during normal operation such as NuScale, CAREM, ESBWR and Indian AHWR designs. Simulation of natural circulation phenomena is very challenging since the driving force of it is weak compared to forced circulation and involves a coupling between primary system and containment for integral type reactor. The IAEA ICSP (International Collaborative Standard Problem) on 'Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents' was proposed within the CRP on 'Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems that utilize Natural Circulation'. Oregon State University (OSU) of USA offered to host this ICSP. This ICSP plans to conduct the following experiments and blind/open simulations with system codes: 1. Quasi-steady state operation with different core power levels: Conduct quasi-steady state operation with step-wise increase of core power level in order to observe single phase natural circulation flow according to power level. The experimental facility and operating conditions for an integral PWR will be used. 2. Thermo-hydraulic Coupling between Primary system and Containment: Conduct a loss of feedwater transient with subsequent ADS blowdown and long term cooling to determine the progression of a loss of feedwater transient by natural circulation through primary and containment systems. These tests would examine the blowdown phase as well as the long term cooling using sump natural circulation by coupling the primary to containment systems. This data could be used for the evaluation of system codes to determine if they model specific phenomena in an accurate manner. OSU completed planned two ICSP tests in July 2011 and real initial and boundary conditions measured from the

  12. 13 CFR 120.1706 - Pool Originator's retained interest in Pool.

    Science.gov (United States)

    2010-01-01

    ... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Pool Originator's retained interest in Pool. 120.1706 Section 120.1706 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Establishment of SBA Secondary Market Guarantee Program for First Lien Position 504 Loan...

  13. Strategies for chemically healthy public swimming pools

    DEFF Research Database (Denmark)

    Hansen, Kamilla Marie Speht

    spreading of pathogens between swimmers because of its residual disinfection effect. In addition to potential contamination of pathogenic microorganisms, swimming pool water is polluted by organic matter deposited from the bathers such as saliva, urine, sweat, hair and personal care products. Since chlorine...... is a strong oxidant it oxidizes the organic matter in the pool water and forms disinfection byproducts (DBPs). More than 100 different DBPs have been identified. Some of these have been found to be genotoxic and may pose an increased cancer risk for the bathers. The aim of this thesis was to give an overview...... of the strategies which can be used to achieve microbiological safe water with low levels of DBPs to ensure healthy environment for bathers. There are different approaches to achieve healthy environment in public swimming pools which in this thesis are divided into three strategies: alternatives to chlorination...

  14. Hot Hydrogen Heat Source Development

    Data.gov (United States)

    National Aeronautics and Space Administration — The purpose of this project is to develop a  hot hydrogen heat source that would produce  a high temperature hydrogen flow which would be comparable to that produced...

  15. The decay of hot nuclei

    International Nuclear Information System (INIS)

    Moretto, L.G.; Wozniak, G.J.

    1988-11-01

    The formation of hot compound nuclei in intermediate-energy heavy ion reactions is discussed. The statistical decay of such compound nuclei is responsible for the abundant emission of complex fragments and high energy gamma rays. 43 refs., 23 figs

  16. Mercury content in Hot Springs

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, R

    1974-01-01

    A method of determination of mercury in hot spring waters by flameless atomic absorption spectrophotometry is described. Further, the mercury content and the chemical behavior of the elementary mercury in hot springs are described. Sulfide and iodide ions interfered with the determination of mercury by the reduction-vapor phase technique. These interferences could, however, be minimized by the addition of potassium permanganate. Waters collected from 55 hot springs were found to contain up to 26.0 ppb mercury. High concentrations of mercury have been found in waters from Shimoburo Springs, Aomori (10.0 ppb), Osorezan Springs, Aomori (1.3 approximately 18.8 ppb), Gosyogake Springs, Akita (26.0 ppb), Manza Springs, Gunma (0.30 approximately 19.5 ppb) and Kusatu Springs, Gunma (1.70 approximately 4.50 ppb). These hot springs were acid waters containing a relatively high quantity of chloride or sulfate.

  17. Do scientists trace hot topics?

    Science.gov (United States)

    Wei, Tian; Li, Menghui; Wu, Chensheng; Yan, Xiao-Yong; Fan, Ying; Di, Zengru; Wu, Jinshan

    2013-01-01

    Do scientists follow hot topics in their scientific investigations? In this paper, by performing analysis to papers published in the American Physical Society (APS) Physical Review journals, it is found that papers are more likely to be attracted by hot fields, where the hotness of a field is measured by the number of papers belonging to the field. This indicates that scientists generally do follow hot topics. However, there are qualitative differences among scientists from various countries, among research works regarding different number of authors, different number of affiliations and different number of references. These observations could be valuable for policy makers when deciding research funding and also for individual researchers when searching for scientific projects.

  18. Effect of pool turbulence on direct contact condensation at a steam/water interface

    International Nuclear Information System (INIS)

    Jackson, J.D.; Zhao, C.L.; Doerffer, S.; Byrne, J.E.; Falaki, H.

    2000-01-01

    Measurements of direct contact condensation beat transfer have been made for the case where the process takes place at the horizontal interface between saturated steam and a pool of water in a vertical cylindrical test section. A submerged vertical jet of subcooled water was injected upwards on the axis to promote the condensation and water was withdrawn at the same rate from the bottom of the pool. In conjunction with the above study, measurements of the turbulent velocity fluctuations just below a free surface produced by the injection of a vertical submerged jet have been measured using hot film anemometry on an isothermal air-water test facility of similar geometry for similar flow conditions at ambient temperature. A correlation is proposed in terms of a Stanton number based on turbulent velocity fluctuation near the interface on the liquid-side. Our results are in good agreement with those of others for similar configurations when compared in terms of condensation Stanton number. (author)

  19. Thermal analyses for the rack design with spent fuel pool during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Yeh, C-L.; Chen, Y-S.; Chen, B-Y., E-mail: clinyeh@iner.gov.tw, E-mail: yschen@iner.gov.tw, E-mail: onepicemine@iner.gov.tw [Inst. of Nuclear Energy Research, Taoyuan County, Taiwan (China); Tseng, Y-S., E-mail: ystseng@mx.nthu.edu.tw [National Tsing Hua Univ., Engineering and System Science, Hsinchu, Taiwan (China); Wei, W-C., E-mail: hn150456@iner.gov.tw [Inst. of Nuclear Energy Research, Taoyuan County, Taiwan (China)

    2014-07-01

    Alternative fuel arrangements separating the latest fuels discharge from the reactor core are proposed, such as the 1x4 configuration in which the hot assembly is surrounded by 4 assemblies with much lower decay heat. For the rack design in the BWR spent fuel pool design, the lateral flow is eliminated by solid walls. In this study, cooling enhancement of splitting fuel rack is investigated using Computational Fluid Dynamics (CFD). The fuels in the pool are modeled by porous medium. Separating the fuel rack by a distance of 10 cm can lower the peak cladding temperature and the natural convection between the fuels and then earns more response time for the site people to implement necessary mitigation actions. (author)

  20. Zooplankton at deep Red Sea brine pools

    KAUST Repository

    Kaartvedt, Stein

    2016-03-02

    The deep-sea anoxic brines of the Red Sea comprise unique, complex and extreme habitats. These environments are too harsh for metazoans, while the brine–seawater interface harbors dense microbial populations. We investigated the adjacent pelagic fauna at two brine pools using net tows, video records from a remotely operated vehicle and submerged echosounders. Waters just above the brine pool of Atlantis II Deep (2000 m depth) appeared depleted of macrofauna. In contrast, the fauna appeared to be enriched at the Kebrit Deep brine–seawater interface (1466 m).

  1. CFD aided analysis of a scaled down model of the Brazilian Multipurpose Reactor (RMB) pool

    International Nuclear Information System (INIS)

    Schweizer, Fernando L.A.; Lima, Claubia P.B.; Costa, Antonella L.; Veloso, Maria A.F.

    2013-01-01

    Research reactors are commonly built inside deep pools that provide radiological and thermal protection and easy access to its core. Reactors with thermal power in the order of MW usually use an auxiliary thermal-hydraulic circuit at the top of its pool to create a purified hot water layer (HWL). Thermal-hydraulic analysis of the flow configuration in the pool and HWL is paramount to insure radiological protection. A useful tool for these analyses is the application of CFD (Computational Fluid Dynamics). To obtain satisfactory results using CFD it is necessary the verification and validation of the CFD numerical model. Verification is divided in code and solution verifications. In the first one establishes the correctness of the CFD code implementation and in the former estimates the numerical accuracy of a particular calculation. Validation is performed through comparison of numerical and experimental results. This paper presents a dimensional analysis of the RMB (Brazilian Multipurpose Reactor) pool to determine a scaled down experimental installation able to aid in the HWL numerical investigation. Two CFD models were created one with the same dimensions and boundary conditions of the reactor prototype and the other with 1/10 proportion size and boundary conditions set to achieve the same inertial and buoyant forces proportions represented by Froude Number between the two models. Results comparing the HWL thickness show consistence between the prototype and the scaled down model behavior. (author)

  2. Numerical modeling of sodium fire – Part II: Pool combustion and combined spray and pool combustion

    International Nuclear Information System (INIS)

    Sathiah, Pratap; Roelofs, Ferry

    2014-01-01

    Highlights: • A CFD based method is proposed for the simulation of sodium pool combustion. • A sodium evaporation based model is proposed to model sodium pool evaporation. • The proposed method is validated against sodium pool experiments of Newman and Payne. • The results obtained using the proposed method are in good agreement with the experiments. - Abstract: The risk of sodium-air reaction has received considerable attention after the sodium-fire accident in Monju reactor. The fires resulting from the sodium-air reaction can be detrimental to the safety of a sodium fast reactor. Therefore, predicting the consequences of a sodium fire is important from a safety point of view. A computational method based on CFD is proposed here to simulate sodium pool fire and understand its characteristics. The method solves the Favre-averaged Navier-Stokes equation and uses a non-premixed mixture fraction based combustion model. The mass transfer of sodium vapor from the pool surface to the flame is obtained using a sodium evaporation model. The proposed method is then validated against well-known sodium pool experiments of Newman and Payne. The flame temperature and location predicted by the model are in good agreement with experiments. Furthermore, the trends of the mean burning rate with initial pool temperature and oxygen concentration are captured well. Additionally, parametric studies have been performed to understand the effects of pool diameter and initial air temperature on the mean burning rate. Furthermore, the sodium spray and sodium pool combustion models are combined to simulate simultaneous spray and pool combustion. Simulations were performed to demonstrate that the combined code could be applied to simulate this. Once sufficiently validated, the present code can be used for safety evaluation of a sodium fast reactor

  3. Numerical modeling of sodium fire – Part II: Pool combustion and combined spray and pool combustion

    Energy Technology Data Exchange (ETDEWEB)

    Sathiah, Pratap, E-mail: pratap.sathiah78@gmail.com [Shell Global Solutions Ltd., Brabazon House, Concord Business Park, Threapwood Road, Manchester M220RR (United Kingdom); Roelofs, Ferry, E-mail: roelofs@nrg.eu [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755ZG Petten (Netherlands)

    2014-10-15

    Highlights: • A CFD based method is proposed for the simulation of sodium pool combustion. • A sodium evaporation based model is proposed to model sodium pool evaporation. • The proposed method is validated against sodium pool experiments of Newman and Payne. • The results obtained using the proposed method are in good agreement with the experiments. - Abstract: The risk of sodium-air reaction has received considerable attention after the sodium-fire accident in Monju reactor. The fires resulting from the sodium-air reaction can be detrimental to the safety of a sodium fast reactor. Therefore, predicting the consequences of a sodium fire is important from a safety point of view. A computational method based on CFD is proposed here to simulate sodium pool fire and understand its characteristics. The method solves the Favre-averaged Navier-Stokes equation and uses a non-premixed mixture fraction based combustion model. The mass transfer of sodium vapor from the pool surface to the flame is obtained using a sodium evaporation model. The proposed method is then validated against well-known sodium pool experiments of Newman and Payne. The flame temperature and location predicted by the model are in good agreement with experiments. Furthermore, the trends of the mean burning rate with initial pool temperature and oxygen concentration are captured well. Additionally, parametric studies have been performed to understand the effects of pool diameter and initial air temperature on the mean burning rate. Furthermore, the sodium spray and sodium pool combustion models are combined to simulate simultaneous spray and pool combustion. Simulations were performed to demonstrate that the combined code could be applied to simulate this. Once sufficiently validated, the present code can be used for safety evaluation of a sodium fast reactor.

  4. 13 CFR 120.1705 - Pool formation requirements.

    Science.gov (United States)

    2010-01-01

    ... requirements. SBA may adjust the Pool characteristics periodically based on program experience and market... a Pool involving a Pool Loan it does not own, it must purchase the Loan Interest it proposes to pool... purchase the Loan Interest and take it into inventory or settle the purchase of the Loan Interest through...

  5. Delineating probabilistic species pools in ecology and biogeography

    OpenAIRE

    Karger, Dirk Nikolaus; Cord, Anna F; Kessler, Michael; Kreft, Holger; Kühn, Ingolf; Pompe, Sven; Sandel, Brody; Sarmento Cabral, Juliano; Smith, Adam B; Svenning, Jens-Christian; Tuomisto, Hanna; Weigelt, Patrick; Wesche, Karsten

    2016-01-01

    Aim To provide a mechanistic and probabilistic framework for defining the species pool based on species-specific probabilities of dispersal, environmental suitability and biotic interactions within a specific temporal extent, and to show how probabilistic species pools can help disentangle the geographical structure of different community assembly processes. Innovation Probabilistic species pools provide an improved species pool definition based on probabilities in conjuncti...

  6. A strategy for optimizing item-pool management

    NARCIS (Netherlands)

    Ariel, A.; van der Linden, Willem J.; Veldkamp, Bernard P.

    2006-01-01

    Item-pool management requires a balancing act between the input of new items into the pool and the output of tests assembled from it. A strategy for optimizing item-pool management is presented that is based on the idea of a periodic update of an optimal blueprint for the item pool to tune item

  7. The radionuclides of primary coolant in HANARO and the recent activities performed to reduce the radioactivity or reactor pool water

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minjin [HANARO Research Reactor Centre, Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1998-10-01

    In HANARO reactor, there have been activities to identify the principal radionuclides and to quantify them under the normal operation. The purposes of such activities were to establish the measure by which we can reduce the radioactivity of the reactor pool water and detect, in early stage, the abnormal symptoms due to the leakage of radioactive materials from the irradiation sample or the damage of the nuclear fuel, etc. The typical radionuclides produced by the activation of reactor coolant are N{sup 16} and Ar{sup 41}. The radionuclides produced by the activation of the core structural material consist of Na{sup 24}, Mn{sup 56}, and W{sup 187}. Of the various radionuclides, governing the radiation level at the pool surface are Na{sup 24}, Ar{sup 41}, Mn{sup 58}, and W{sup 187}. By establishing the hot water layer system on the pool surface, we expected that the radionuclides such as Ar{sup 41} and Mn{sup 56} whose half-life are relatively short could be removed to a certain extent. Since the content of radioactivity of Na{sup 24} occupies about 60% of the total radioactivity, we assumed that the total radiation level would be greatly reduced if we could decrease the radiation level of Na{sup 24}. However the actual radiation level has not been reduced as much as we expected. Therefore, some experiments have been carried out to find the actual causes afterwards. What we learned through the experiments are that any disturbance in reactor pool water layer causes increase of the pool surface radiation level and even if we maintain the hot water layer well, reactor shutdown will be very much likely to happen once the hot water layer is disturbed. (author)

  8. Thermo-hydraulic instability of natural circulation BWRs at low pressure star-up. Experimental estimation of instability region with test facility considering scaling law

    International Nuclear Information System (INIS)

    Inada, F.; Furuya, M.; Yasuo, A.; Tabata, H.; Yoshioka, Y.; Kim, H.T.

    1995-01-01

    In natural circulation BWRs developed for advanced light water reactors with simplified passive safety systems, thermo-hydraulic stability should be confirmed especially at low pressure start-up. In this paper, nondimensional parameters to estimate the hydrodynamic stability to reactors at low pressure start-up were obtained by transformation of the basic equations of drift-flux model in the two-phase region into nondimensional form. A test facility based on these parameters was then constructed. The height of the test facility is 70% of SBWR and many nondimensional test facility parameters are almost the same as those of the reactor. Reactor stability was estimated experimentally. Stability maps below 0.5MPa were obtained on the heat flux - channel inlet subcooling place. It was found that there were two stability boundaries, between which the flow became unstable. Flow was stable in the high and low channel inlet subcooling regions. Typical conditions of SBWR at low pressure start-up were noted in the high channel inlet subcooling stable region. The heat flux at typical SBWR start-up was about one fifth that of the stability boundary. Though some nondimensional parameters of the test facility did not exactly agree with those of SBWR, it was suggested that the flow in SBWR was stable below 0.5MPa because of the large margin. (author)

  9. Thermo-hydraulic modelling of the South East Gas Pipeline System - an integrated model; Modelagem termo-hidraulica do Sistema de Gasodutos do Sudeste : um modelo integrado

    Energy Technology Data Exchange (ETDEWEB)

    Vianna Neto, Armando M.; Santos, Arnaldo M.; Mercon, Eduardo G. [TRANSPETRO - PETROBRAS Transportes, Rio de Janeiro, RJ (Brazil)

    2003-07-01

    This paper presents the development of an integrated simulation model, for the numerical calculation of thermal-hydraulic behaviors in the Brazilian southeast onshore gas pipeline flow system, remotely operated by TRANSPETRO's Gas Pipeline Control Centre (CCG). In its final application, this model is supposed to provide simulated results at the closer range to reality, in order to improve gas pipeline simulation studies and evaluations for the system in question. Considering the fact that numerical thermo-hydraulic simulation becomes the CCG's most important tool to analyze the boundary conditions to adjust the mentioned gas flow system, this paper seeks and takes aim to the optimization of the following prime attributions of a gas pipeline control centre: verification of system behaviors, face to some unit maintenance stop or procedure, programmed or not, or to some new gas outlet or inlet connection to the system; daily operational compatibility analysis between programmed and realized gas volumes; gas technical expedition and delivery analysis. Finally, all this work was idealized and carried out within the one-phase flow domain (dry gas) (author)

  10. The applicability of CFD to simulate and study the mixing process and the thermo-hydraulic consequences of a main steam line break in PWR model

    Directory of Open Access Journals (Sweden)

    Farkas Istvan

    2017-01-01

    Full Text Available This paper focuses on the validation and applicability of CFD to simulate and analyze the thermo-hydraulic consequences of a main steam line break. Extensive validation data come from experiments performed using the Rossendorf coolant mixing model facility. For the calculation, the range of 9 to 12 million hexahe¬dral cells was constructed to capture all details in the interrogation domain in the system. The analysis was performed by running a time-dependent calculation, Detailed analyses were made at different cross-sections in the system to evaluate not only the value of the maximum and minimum temperature, but also the loca¬tion and the time at which it occurs during the transient which is considered to be indicator for the quality of mixing in the system. CFD and experimental results were qualitatively compared; mixing in the cold legs with emergency core cooling systems was overestimated. This could be explained by the sensitivity to the bound¬ary conditions. In the downcomer, the experiments displayed higher mixing: by our assumption this related to the dense measurement grid (they were not modelled. The temperature distribution in the core inlet plane agreed with the measurement results. Minor deviations were seen in the quantitative comparisons: the maximum temperature difference was 2ºC.

  11. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach; Simulacao termohidraulica do nucleo do reator nuclear HTR-10 com o uso da abordagem realistica CFD

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S.; Dominguez, Dany S., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil); Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas, La Habana (Cuba); Lira, Carlos Alberto Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  12. Feasibility study of the university of Utah TRIGA reactor power upgrade - part II: Thermohydraulics and heat transfer study in respect to cooling system requirements and design

    Directory of Open Access Journals (Sweden)

    Babitz Philip

    2013-01-01

    Full Text Available The thermodynamic conditions of the University of Utah's TRIGA Reactor were simulated using SolidWorks Flow Simulation, Ansys, Fluent and PARET-ANL. The models are developed for the reactor's currently maximum operating power of 90 kW, and a few higher power levels to analyze thermohydraulics and heat transfer aspects in determining a design basis for higher power including the cost estimate. It was found that the natural convection current becomes much more pronounced at higher power levels with vortex shedding also occurring. A departure from nucleate boiling analysis showed that while nucleate boiling begins near 210 kW it remains in this state and does not approach the critical heat flux at powers up to 500 kW. Based on these studies, two upgrades are proposed for extended operation and possibly higher reactor power level. Together with the findings from Part I studies, we conclude that increase of the reactor power is highly feasible yet dependable on its purpose and associated investments.

  13. Implementation of the optimization for the methodology of the neutronic calculation and thermo-hydraulic in IEA-R1 reactor

    International Nuclear Information System (INIS)

    Stefani, Giovanni Laranjo de; Conti, Thadeu das Neves; Fedorenko, Giuliana G.; Castro, Vinicius A.; Maio, Mireia F.; Santos, Thiago Augusto dos

    2011-01-01

    This work objective was to create a manager program that would automate the programs and computer codes in use for neutronic calculation and thermo-hydraulic in IEA-R1 reactor thus making the process for calculation of safety parameters and for configuration change up to 98% faster than that used in the reactor today. This process was tested in combination with the reactor operators and is being implemented by the quality department. The main codes and programs involved in the calculations of configuration change are Leopard, Hammier-Technion, Twodb, Citation and Cobra. Calculations of delayed neutron and criticality coefficients given in the process of safety parameters calculation are given by the Hammer-Technion and Citation in a process that involves about eleven repetitions so that it meets all the necessary conditions (such different temperatures of the moderator and fuel). The results are entirely consistent with the expected and absolutely the same as those given by manual process. Thus the work shows its reliability as well the advantage of saving time, once a process that could take up to four hours was turned in one that takes around five minutes when done in a home computer. Much of this advantage is due to the fact that were created subprograms to treat the output of each program used and transform them into the input of the other programs, removing from it the intermediate essential data for this to occur, thus avoiding also a possible human error by handling the various data supplied. (author)

  14. On closure strategy for 1-D thermohydraulics models and closure relationships of two-phase flows in simple and subchannel geometry for NPP accident conditions

    International Nuclear Information System (INIS)

    Kornienko, Y.; Kornienko, E.; Ninokata, H.

    2001-01-01

    One-dimensional mathematical models are extensively used in thermohydraulics assessment of Nuclear Power Plant (NPP) transients and accidents, because specifically 1-D system of the conservation laws allows to reduce computing time and required memory, especially in ''best estimate'' code calculations. This work is generalization of the well-known Zuber-Findley and Hancox-Nicoll methods for two-phase flow distribution parameters Cs taking into account the non-monotonous void fraction distribution in the transverse direction in terms of two superimposed monotonous profiles. The method is very useful in evaluating the saddle-shape void fraction profile effects. In this work two-phase flow distribution parameters Cs were developed for simple circular and rectangular pipes, and subchannel geometry in a rod bundle. Basic assumptions were power-mode approximations for describing the profiles of local volume flux density, phase velocity and temperature. The general analytical (quadrature) relationships for Cs were obtained and their 3-D illustrations are proposed. Also, we propose generalized formulation and simple approach to construct friction factor, heat and mass transfer coefficients within the gradient hypothesis and boundary layer assumptions. The contribution of momentum, heat and mass transfer as well as their sources and sinks in the channel cross-section are taken into account. In the same way, the friction factor, heat and mass transfer coefficients with the transversal and azimuthal variations being taken into account are proposed for subchannel geometry as well. (author)

  15. Public Swimming Pools | Florida Department of Health

    Science.gov (United States)

    Illness Disease Reporting and Surveillance Bureau of Public Health Laboratories Environmental Health Air Air Monitoring Carbon Monoxide Indoor Air Quality Mold Radon Water Aquatic Toxins Beach Water Quality purification, testing, treatment, and disinfection procedures. To ensure that the pool technicians are

  16. 28 CFR 540.64 - Press pools.

    Science.gov (United States)

    2010-07-01

    ... Judicial Administration BUREAU OF PRISONS, DEPARTMENT OF JUSTICE INSTITUTIONAL MANAGEMENT CONTACT WITH PERSONS IN THE COMMUNITY Contact With News Media § 540.64 Press pools. (a) The Warden may establish a... shall notify all news media representatives who have requested interviews or visits that have not been...

  17. Stein's neuronal model with pooled renewal input

    Czech Academy of Sciences Publication Activity Database

    Rajdl, K.; Lánský, Petr

    2015-01-01

    Roč. 109, č. 3 (2015), s. 389-399 ISSN 0340-1200 Institutional support: RVO:67985823 Keywords : Stein’s model * Poisson process * pooled renewal processes * first-passage time Subject RIV: BA - General Mathematics Impact factor: 1.611, year: 2015

  18. Phase analysis in gated blood pool tomography

    International Nuclear Information System (INIS)

    Nakajima, Kenichi; Bunko, Hisashi; Tada, Akira; Taki, Junichi; Nanbu, Ichiro

    1984-01-01

    Phase analysis of gated blood pool study has been applied to detect the site of accessory conduction pathway (ACP) in the Wolff-Parkinson-White (WPW) syndrome; however, there was a limitation to detect the precise location of ACP by phase analysis alone. In this study, we applied phase analysis to gated blood pool tomography using seven pin hole tomography (7PT) and gated emission computed tomography (GECT) in 21 patients with WPW syndrome and 3 normal subjects. In 17 patients, the sites of ACPs were confirmed by epicardial mapping and the result of the surgical division of ACP. In 7PT, the site of ACP grossly agreed to the abnormal initial phase in phase image in 5 out of 6 patients with left cardiac type. In GECT, phase images were generated in short axial, vertical and horizontal long axial sections. In 8 out of 9 patients, the site of ACP was correctly identified by phase images, and in a patient who had two ACPs, initial phase corresponded to one of the two locations. Phase analysis of gated blood pool tomography has advantages for avoiding overlap of blood pools and for estimating three-dimensional propagation of the contraction, and can be a good adjunctive method in patients with WPW syndrome. (author)

  19. Swimming-pool piles; Piles piscines

    Energy Technology Data Exchange (ETDEWEB)

    Trioulaire, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    In France two swimming-pool piles, Melusine and Triton, have just been set in operation. The swimming-pool pile is the ideal research tool for neutron fluxes of the order of 10{sup 13}. This type of pile can be of immediate interest to many research centres, but its cost must be reduced and a break with tradition should be observed in its design. It would be an advantage: - to bury the swimming-pool; - to reject the experimental channel; - to concentrate the cooling circuit in the swimming-pool; - to carry out all manipulations in the water; - to double the core. (author) [French] En France, deux piles piscines, Melusine et Triton, viennent d'entrer en service. La pile piscine est l'outil de recherche ideal pour des flux de neutrons de l'ordre de 10{sup 13}. Ce type de pile peut interesser des maintenant de nombreux centres de recherches mais il faut reduire son prix de revient et rompre avec le conformisme de sa conception. Il y a avantage: - a enterrer la piscine; - a supprimer les canaux experimentaux; - a concentrer le circuit de refrigeration dans la piscine; - a effectuer toutes les manipulations dans l'eau; - a doubler le coeur. (auteur)

  20. Spectrum pooling in MnWave Networks

    DEFF Research Database (Denmark)

    Boccardi, Federico; Shokri-Ghadikolaei, Hossein; Fodor, Gabor

    2016-01-01

    Motivated by the specific characteristics of mmWave technologies, we discuss the possibility of an authorization regime that allows spectrum sharing between multiple operators, also referred to as spectrum pooling. In particular, considering user rate as the performance measure, we assess...

  1. Pooling Objects for Recognizing Scenes without Examples

    NARCIS (Netherlands)

    Kordumova, S.; Mensink, T.; Snoek, C.G.M.

    2016-01-01

    In this paper we aim to recognize scenes in images without using any scene images as training data. Different from attribute based approaches, we do not carefully select the training classes to match the unseen scene classes. Instead, we propose a pooling over ten thousand of off-the-shelf object

  2. The new pooled cohort equations risk calculator

    DEFF Research Database (Denmark)

    Preiss, David; Kristensen, Søren L

    2015-01-01

    disease and any measure of social deprivation. An early criticism of the Pooled Cohort Equations Risk Calculator has been its alleged overestimation of ASCVD risk which, if confirmed in the general population, is likely to result in statin therapy being prescribed to many individuals at lower risk than...

  3. DNA pooling strategies for categorical (ordinal) traits

    Science.gov (United States)

    Despite reduced genotyping costs in recent years, obtaining genotypes for all individuals in a population may still not be feasible when sample size is large. DNA pooling provides a useful alternative to determining genotype effects. Clustering algorithms allow for grouping of individuals (observati...

  4. Pooling ASR data for closely related languages

    CSIR Research Space (South Africa)

    Van Heerden, C

    2010-05-01

    Full Text Available We describe several experiments that were conducted to assess the viability of data pooling as a means to improve speech-recognition performance for under-resourced languages. Two groups of closely related languages from the Southern Bantu language...

  5. "Teisele poole" Agambeniga ja Agambenita / Ragne Nukk

    Index Scriptorium Estoniae

    Nukk, Ragne, 1984-

    2012-01-01

    Tallinna Fotokuu rahvusvahelise fotonäituse "Teisele poole" ideeliste lähtekohtade tõlgendamisest. Autor kõrvutab Adam Budaki kuraatornäituse aluseks olnud Giorgio Agambeni esseed „Viimne kohtupäev“ näitusel eksponeeritud piltidega

  6. 7 CFR 1033.7 - Pool plant.

    Science.gov (United States)

    2010-01-01

    ... association and it fulfills the following conditions: (1) The aggregate monthly quantity supplied by all... or aseptically-processed fluid milk products. (c) A supply plant from which the quantity of bulk... operator of a supply plant located within the marketing area may include deliveries to pool distributing...

  7. Benthic assemblages of rock pools in northern Portugal: seasonal and between-pool variability

    Directory of Open Access Journals (Sweden)

    Iacopo Bertocci

    2012-11-01

    Full Text Available We investigated the seasonal (winter vs summer and within season and spatial (between-pool variability of benthic assemblages of rock pools at mid-intertidal level along the shore of Viana do Castelo (North Portugal. Physical traits of rock pools, including size, depth and position along the shore, were also compared between pools. While pools did not differ for any of the examined physical traits, results indicated a clear seasonal difference in the structure of assemblages, including a total of 49 macroalgal and 13 animal taxa. This finding was driven by six taxa that are more abundant in winter (the reef-forming polychaete Sabellaria alveolata, the articulated coralline algae Corallina spp., the brown alga Bifurcaria bifurcata, the encrusting coralline alga Lithophyllum incrustans, the red alga Chondracanthus acicularis and the grazing snails Gibbula spp. and four algal taxa that are more abundant in summer (the invasive brown Sargassum muticum, the green Ulva spp., the kelp Laminaria ochroleuca and the filamentous red Ceramium spp.. These data provide a new contribution to the knowledge of rock pool systems and have potential implications for monitoring programmes aimed at assessing ecological modifications related to natural and anthropogenic disturbances and for identifying processes responsible for the variability of rock pool assemblages.

  8. Uncertainty analysis for hot channel

    International Nuclear Information System (INIS)

    Panka, I.; Kereszturi, A.

    2006-01-01

    The fulfillment of the safety analysis acceptance criteria is usually evaluated by separate hot channel calculations using the results of neutronic or/and thermo hydraulic system calculations. In case of an ATWS event (inadvertent withdrawal of control assembly), according to the analysis, a number of fuel rods are experiencing DNB for a longer time and must be regarded as failed. Their number must be determined for a further evaluation of the radiological consequences. In the deterministic approach, the global power history must be multiplied by different hot channel factors (kx) taking into account the radial power peaking factors for each fuel pin. If DNB occurs it is necessary to perform a few number of hot channel calculations to determine the limiting kx leading just to DNB and fuel failure (the conservative DNBR limit is 1.33). Knowing the pin power distribution from the core design calculation, the number of failed fuel pins can be calculated. The above procedure can be performed by conservative assumptions (e.g. conservative input parameters in the hot channel calculations), as well. In case of hot channel uncertainty analysis, the relevant input parameters (k x, mass flow, inlet temperature of the coolant, pin average burnup, initial gap size, selection of power history influencing the gap conductance value) of hot channel calculations and the DNBR limit are varied considering the respective uncertainties. An uncertainty analysis methodology was elaborated combining the response surface method with the one sided tolerance limit method of Wilks. The results of deterministic and uncertainty hot channel calculations are compared regarding to the number of failed fuel rods, max. temperature of the clad surface and max. temperature of the fuel (Authors)

  9. Generalizing Pooling Functions in CNNs: Mixed, Gated, and Tree.

    Science.gov (United States)

    Lee, Chen-Yu; Gallagher, Patrick; Tu, Zhuowen

    2018-04-01

    In this paper, we seek to improve deep neural networks by generalizing the pooling operations that play a central role in the current architectures. We pursue a careful exploration of approaches to allow pooling to learn and to adapt to complex and variable patterns. The two primary directions lie in: (1) learning a pooling function via (two strategies of) combining of max and average pooling, and (2) learning a pooling function in the form of a tree-structured fusion of pooling filters that are themselves learned. In our experiments every generalized pooling operation we explore improves performance when used in place of average or max pooling. We experimentally demonstrate that the proposed pooling operations provide a boost in invariance properties relative to conventional pooling and set the state of the art on several widely adopted benchmark datasets. These benefits come with only a light increase in computational overhead during training (ranging from additional 5 to 15 percent in time complexity) and a very modest increase in the number of model parameters (e.g., additional 1, 9, and 27 parameters for mixed, gated, and 2-level tree pooling operators, respectively). To gain more insights about our proposed pooling methods, we also visualize the learned pooling masks and the embeddings of the internal feature responses for different pooling operations. Our proposed pooling operations are easy to implement and can be applied within various deep neural network architectures.

  10. Corium quench in deep pool mixing experiments

    International Nuclear Information System (INIS)

    Spencer, B.W.; McUmber, L.; Gregorash, D.; Aeschlimann, R.; Sienicki, J.J.

    1985-01-01

    The results of two recent corium-water thermal interaction (CWTI) tests are described in which a stream of molten corium was poured into a deep pool of water in order to determine the mixing behavior, the corium-to-water heat transfer rates, and the characteristic sizes of the quenched debris. The corium composition was 60% UO 2 , 16% ZrO 2 , and 24% stainless steel by weight; its initial temperature was 3080 K, approx.160 K above the oxide phase liquidus temperature. The corium pour stream was a single-phase 2.2 cm dia liquid column which entered the water pool in film boiling at approx.4 m/s. The water subcooling was 6 and 75C in the two tests. Test results showed that with low subcooling, rapid steam generation caused the pool to boil up into a high void fraction regime. In contrast, with large subcooling no net steam generation occurred, and the pool remained relatively quiescent. Breakup of the jet appeared to occur by surface stripping. In neither test was the breakup complete during transit through the 32 cm deep water pool, and molten corium channeled to the base where it formed a melt layer. The characteristic heat transfer rates measured 3.5 MJ/s and 2.7 MJ/s during the fall stage for small and large subcooling, respectively; during the initial stage of bed quench, the surface heat fluxes measured 2.4 MW/m 2 and 3.7 MW/m 2 , respectively. A small mass of particles was formed in each test, measuring typically 0.1 to 1 mm and 1 to 5 mm dia for the large and small subcooling conditions, respectively. 9 refs., 13 figs., 1 tab

  11. Suncatcher and cool pool. Project report

    Energy Technology Data Exchange (ETDEWEB)

    Hammond, J.

    1981-03-01

    The Suncatcher is a simple, conical solar concentrating device that captures light entering clerestory windows and directs it onto thermal storage elements at the back of a south facing living space. The cone shape and inclination are designed to capture low angle winter sunlight and to reflect away higher angle summer sunlight. It is found that winter radiation through a Suncatcher window is 40 to 50% higher than through an ordinary window, and that the average solar fraction is 59%. Water-filled steal culvert pipes used for thermal storage are found to undergo less stratification, and thus to be more effective, when located where sunlight strikes the bottom rather than the top. Five Suncatcher buildings are described. Designs are considered for 32/sup 0/, 40/sup 0/ and 48/sup 0/ north latitude, and as the latitude increases, the inclination angle of the cone should be lowered. The Cool Pool is an evaporating, shaded roof pond which thermosiphons cool water into water-filled columns within a building. Preliminary experiments indicate that the best shade design has unimpeded north sky view, good ventilation, complete summer shading, a low architectural profile, and low cost attic vent lowers work. Another series of experiments established the satisfactory performance of the Cool Pool on a test building using four water-filled cylinders, two cylinders, and two cylinders connected to the Cool Pool through a heat exchanger. Although an unshaded pool cools better at night than a shaded one, daytime heat gain far offsets this advantage. A vinyl waterbag heat exchanger was developed for use with the Cool Pool. (LEW)

  12. Statistical hot spot analysis of reactor cores

    International Nuclear Information System (INIS)

    Schaefer, H.

    1974-05-01

    This report is an introduction into statistical hot spot analysis. After the definition of the term 'hot spot' a statistical analysis is outlined. The mathematical method is presented, especially the formula concerning the probability of no hot spots in a reactor core is evaluated. A discussion with the boundary conditions of a statistical hot spot analysis is given (technological limits, nominal situation, uncertainties). The application of the hot spot analysis to the linear power of pellets and the temperature rise in cooling channels is demonstrated with respect to the test zone of KNK II. Basic values, such as probability of no hot spots, hot spot potential, expected hot spot diagram and cumulative distribution function of hot spots, are discussed. It is shown, that the risk of hot channels can be dispersed equally over all subassemblies by an adequate choice of the nominal temperature distribution in the core

  13. Hot dry rock heat mining

    International Nuclear Information System (INIS)

    Duchane, D.V.

    1992-01-01

    Geothermal energy utilizing fluids from natural sources is currently exploited on a commercial scale at sites around the world. A much greater geothermal resource exists, however, in the form of hot rock at depth which is essentially dry. This hot dry rock (HDR) resource is found almost everywhere, but the depth at which usefully high temperatures are reached varies from place to place. The technology to mine the thermal energy from HDR has been under development for a number of years. Using techniques adapted from the petroleum industry, water is pumped at high pressure down an injection well to a region of usefully hot rock. The pressure forces open natural joints to form a reservoir consisting of a small amount of water dispensed in a large volume of hot rock. This reservoir is tapped by second well located at some distance from the first, and the heated water is brought to the surface where its thermal energy is extracted. The same water is then recirculated to mine more heat. Economic studies have indicated that it may be possible to produce electricity at competitive prices today in regions where hot rock is found relatively close to the surface

  14. Hot Jupiters around M dwarfs

    Directory of Open Access Journals (Sweden)

    Murgas F.

    2013-04-01

    Full Text Available The WFCAM Transit Survey (WTS is a near-infrared transit survey running on the United Kingdom Infrared Telescope (UKIRT. We conduct Monte Carlo transit injection and detection simulations for short period (<10 day Jupiter-sized planets to characterize the sensitivity of the survey. We investigate the recovery rate as a function of period and magnitude in 2 hypothetical star-planet cases: M0–2 + hot Jupiter, M2–4 + hot Jupiter. We find that the WTS lightcurves are very sensitive to the presence of Jupiter-sized short-period transiting planets around M dwarfs. The non-detection of a hot-Jupiter around an M dwarf by the WFCAM Transit Survey allows us to place a firm upper limit of 1.9 per cent (at 95 per cent confidence on the planet occurrence rate.

  15. Promethus Hot Leg Piping Concept

    International Nuclear Information System (INIS)

    AM Girbik; PA Dilorenzo

    2006-01-01

    The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactor (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept

  16. Hot-pressing steatite bodies

    International Nuclear Information System (INIS)

    Aparicio Arroyo, E.

    1967-01-01

    Requirements for some special nuclear engineering ceramic shapes are: big size, impervious, dimensional accuracy and good mechanical and dielectric properties. Limitations of te conventional methods and advantages of te hot pressing techniques for the manufacturing of these shapes are discussed. Hot pressing characteristics of a certain steatite powder are studied. Occurrence of an optimum densification temperature just above the tale decomposition range is found. Experimental data show that the height/diameter ratio of the specimen has no effect on the sintering conditions. Increasing darkness from the graphite mould is detected above the optimum temperature. The hot-pressed steatite is compared with a fired dry-pressed sample of the same composition. (Author) 13 refs

  17. The Performance Evaluation of a Hot Water Layer using a Numerical Simulation

    International Nuclear Information System (INIS)

    Park, Jong Hark; Chae, Hee Taek; Kim, Heon Il; Jun, Byung Jin; Park, Cheol

    2009-01-01

    Most of all research reactors are immerged in the deep water pool to be a ultimate heat sink. At the neighbor of the reactor, some radio-active matters, such as Na-24, Ar-41, Mg-27, Al-28 and etc, may be generated by the neutron irradiation. Those radio-active isotopes may rise up to the pool water surface through the natural convection flow, which can make the radioactivity in the reactor hall rise high enough to concern about the health of people working in the reactor hall. When the irradiation test facilities are loaded or unloaded during a normal operation, the highly radio-activated primary coolant may flow out through the irradiation test holes on the top of the reactor. This also may be a main hazard source to make the working environment of the reactor hall bad. Making a hot water layer 1.5 ∼ 2.0 m thick at the top of reactor pool would be a good measure to resolve that problem. The hot water layer is formed by a thermal stratification of pool water, which can effectively suppress the ascending of the radio-active matters and primary coolant flowing out from the IR holes. In this study a performance evaluation of the hot water layer is conducted by a computational fluid dynamics technique. According to the results of the prediction the hot water layer is formed well about 1.5 m thick, and can suppress the flows containing radioactive matters ascending from the neighbor of the reactor

  18. Evaluating performance from spiral polyethylene tubes as solar collectors for heating swimming pools

    Energy Technology Data Exchange (ETDEWEB)

    Stefanelli, Anderson Thiago Pontes; Marchi Neto, Ismael de; Scalon, Vicente Luiz; Padilha, Alcides [UNESP, Universidade Estadual Paulista Julio de Mesquita Filho, Bauru, SP (Brazil). Dept. de Engenharia Mecanica], e-mails: scalon@feb.unesp.br, padilha@feb.unesp.br

    2010-07-01

    The solar energy is very common in the daily of citizens from different regions in world. Environmental questions and the consequent Development of renewable energy techniques were a decisive factor for expanding this market. Currently, the solar energy is present in many different devices: as direct conversion through photovoltaic panels as in solar domestic for hot water systems(SDHWS). Another common use is in the heating system for swimming pools, that could be utilized for therapeutic or comfort reasons. The main aspect that increments this use is the economy for operation of these systems. On the other hand, these systems need a high initial investment. Reducing this cost without reduction in collector efficiency using new materials and / or alternative projects is important target for new researches. Thus, this paper aims to analyze the efficiency of one of these alternative models for heating swimming pools. The conceptual device evaluated is a low cost model. It could be made from polyethylene tubes forming spiral heat exchangers. Analysis of the system is based on a dynamic model using differential equations system including solar collector and swimming pool. Experimental radiation and other environmental conditions in the region of Bauru-SP are used for analyse the dynamic behavior of the system. The simulations are based on analysis of three main parameters: number of collectors, the pump drive time and wall thickness of the collector of polyethylene. Based on these numerical tests one can conclude that this new model of solar collector for swimming pool has a better cost benefit ratio when superficial area is equal to 80% of pool area, pump operation is alternating with four minutes turned on and 28 turned off and the polyethylene wall thickness is 1.5 mm (author)

  19. Monopole transitions in hot nuclei

    International Nuclear Information System (INIS)

    Sujkowski, Z.

    1994-01-01

    Monopole transitions can be a signature of shape changing in a hot, pulsating nucleus (the low energy E0 mode) and/or a measure of the compressibility of finite nuclei (GMR, the breathing mode). Experimental information pertaining to GMR is reviewed. Recipes for deducing the incompressibility modules for infinite nuclear matter from data on GMR are discussed. Astrophysical implications are outlined. The first attempts at locating the GMR strength in moderately hot nuclei are described. Prospects for improving the experimental techniques to make an observation of this strength in selected nuclei unambiguous are discussed. (author). 46 refs, 8 figs

  20. Monopole transitions in hot nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Sujkowski, Z. [Soltan Inst. for Nuclear Studies, Otwock-Swierk (Poland)

    1994-12-31

    Monopole transitions can be a signature of shape changing in a hot, pulsating nucleus (the low energy E0 mode) and/or a measure of the compressibility of finite nuclei (GMR, the breathing mode). Experimental information pertaining to GMR is reviewed. Recipes for deducing the incompressibility modules for infinite nuclear matter from data on GMR are discussed. Astrophysical implications are outlined. The first attempts at locating the GMR strength in moderately hot nuclei are described. Prospects for improving the experimental techniques to make an observation of this strength in selected nuclei unambiguous are discussed. (author). 46 refs, 8 figs.

  1. Hot atom chemistry of sulphur

    International Nuclear Information System (INIS)

    Todorovski, D. S.; Koleva, D. P.

    1982-01-01

    An attempt to cover all papers dealing with the hot atom chemistry of sulpphur is made. Publications which: a) only touch the problem, b) contain some data, indirectly connected with sulphur hot atom chemistry, c) deal with 35 S-production from a chloride matrix, are included as well. The author's name and literature source are given in the original language, transcribed, when it is necessary, in latine. A number of primery and secondary documents have been used including Chemical Abstracts, INIS Atomindex, the bibliographies of A. Siuda and J.-P. Adloff for 1973 - 77, etc. (authors)

  2. Construction of concrete hot cells

    International Nuclear Information System (INIS)

    1981-12-01

    The standard is to be applied to rooms (hot cells) which are enclosed by a concrete shield and in which radioactive material is handled by remote control. The rooms may be in facilities for experimental purposes (e.g. development of fuel elements and materials or of chemical processes) or in facilities for production purposes (e.g. reprocessing of nuclear fuel or treatment of radioactive wastes). The standard is to give a design hasis for concrete hot cells and their installations which is to be applied by designers, constructors, future users and competent authorities as well as independent experts. (orig.) [de

  3. Construction of concrete hot cells

    International Nuclear Information System (INIS)

    1980-09-01

    The standard is to be applied to rooms (hot cells) which are enclosed by a concrete shield and in which radioactive material is handled by remote control. The rooms may be in facilities for experimental purposes (e.g. development of fuel elements and materials or of chemical processes) or in facilities for production purposes (e.g. reprocessing of nuclear fuel or treatment of radioactive wastes). The standard is to give a design basis for concrete hot cells and their installations which is to be applied by designers, constructors, future users and competent authorities as well as independent experts. (orig.) [de

  4. Hot-cell verification facility

    International Nuclear Information System (INIS)

    Eschenbaum, R.A.

    1981-01-01

    The Hot Cell Verification Facility (HCVF) was established as the test facility for the Fuels and Materials Examination Facility (FMEF) examination equipment. HCVF provides a prototypic hot cell environment to check the equipment for functional and remote operation. It also provides actual hands-on training for future FMEF Operators. In its two years of operation, HCVF has already provided data to make significant changes in items prior to final fabrication. It will also shorten the startup time in FMEF since the examination equipment will have been debugged and operated in HCVF

  5. Horizontal above-rack pool storage

    International Nuclear Information System (INIS)

    Moscardini, R.L.

    1993-08-01

    This report describes a unique method for storing spent, six year out of core, fuel at a prototypical PWR nuclear power station. The study describes a conceptual design, with favorable structural, thermal and criticality technical evaluations. However, economic considerations and licensing risks are judged to be less favorable. The concept study prescribes a fuel over fuel arrangement in an existing Spent Fuel Pool (SFP) with full maintenance of ALARA principles. This concept study is specific to a prototypical pool design, but may easily be projected to other nuclear facilities with other SFP conditions. For the prototypical PWR, the conceptual fuel bridge design will store over 200 additional fuel assemblies without significant facility modifications and for an indefinite time period

  6. Sodium pool fire model for CONACS code

    International Nuclear Information System (INIS)

    Yung, S.C.

    1982-01-01

    The modeling of sodium pool fires constitutes an important ingredient in conducting LMFBR accident analysis. Such modeling capability has recently come under scrutiny at Westinghouse Hanford Company (WHC) within the context of developing CONACS, the Containment Analysis Code System. One of the efforts in the CONACS program is to model various combustion processes anticipated to occur during postulated accident paths. This effort includes the selection or modification of an existing model and development of a new model if it clearly contributes to the program purpose. As part of this effort, a new sodium pool fire model has been developed that is directed at removing some of the deficiencies in the existing models, such as SOFIRE-II and FEUNA

  7. Pool Boiling CHF in Inclined Narrow Annuli

    International Nuclear Information System (INIS)

    Kang, Myeong Gie

    2010-01-01

    Pool boiling heat transfer has been studied extensively since it is frequently encountered in various heat transfer equipment. Recently, it has been widely investigated in nuclear power plants for application to the advanced light water reactors designs. Through the review on the published results it can be concluded that knowledge on the combined effects of the surface orientation and a confined space on pool boiling heat transfer is of great practical importance and also of great academic interest. Fujita et al. investigated pool boiling heat transfer, from boiling inception to the critical heat flux (CHF, q' CHF ), in a confined narrow space between heated and unheated parallel rectangular plates. They identified that both the confined space and the surface orientation changed heat transfer much. Kim and Suh changed the surface orientation angles of a downward heating rectangular channel having a narrow gap from the downward-facing position (180 .deg.) to the vertical position (90 .deg.). They observed that the CHF generally decreased as the inclination angle (θ ) increased. Yao and Chang studied pool boiling heat transfer in a confined heat transfer for vertical narrow annuli with closed bottoms. They observed that when the gap size ( s ) of the annulus was decreased the effect of space confinement to boiling heat transfer increased. The CHF was occurred at much lower value for the confined space comparing to the unconfined pool boiling. Pool boiling heat transfer in narrow horizontal annular crevices was studied by Hung and Yao. They concluded that the CHF decreased with decreasing gap size of the annuli and described the importance of the thin film evaporation to explain the lower CHF of narrow crevices. The effect of the inclination angle on the CHF on countercurrent boiling in an inclined uniformly heated tube with closed bottoms was also studied by Liu et al. They concluded that the CHF reduced with the inclination angle decrease. A study was carried out

  8. Review and assessment of pool scrubbing models

    International Nuclear Information System (INIS)

    Herranz, L.E.; Escudero, M.J.; Peyres, V.; Polo, J.; Lopez-Jimenez, J.

    1996-01-01

    Decontamination of fission products bearing bubbles as they through aqueous pools becomes a crucial phenomenon for source term evaluation of hypothetical risk dominant sequences of Light Water Reactors. In the present report a peer review and assessment of models encapsulated in SPARC andBUSCA codes is presented. Several aspects of pool scrubbing have been addressed: particle removal, fission product vapour retention and bubble hydrodynamics. Particular emphasis has been given to the close link between retention and hydrodynamics, from both modelling and experimental point of view. In addition, RHR and SGTR sequences were simulated with SPARC90 and BUSCA-AUG92 codes, and their results were compared with those obtained with MAAP 3.0B.As a result of this work, model capabilities and shortcomings have beenassessed and some areas susceptible of further research have been identified.(Author) 73 refs

  9. Review and assessment of pool scrubbing models

    International Nuclear Information System (INIS)

    Herranz, L.E.; Escudero, M.J.; Peyres, V.; Polo, J.; Lopez, J.

    1996-01-01

    Decontamination of fission products bearing bubbles as they pass through aqueous pools becomes a crucial phenomenon for source term evaluation of hypothetical risk dominant sequences of Light Water Reactors. In the present report a peer review and assessment of models encapsulated in SPARC and BUSCA codes is presented. Several aspects of pool scrubbing have been addressed: particle removal, fission product vapour retention and bubble hydrodynamics. Particular emphasis has been given to the close link between retention and hydrodynamics, from both modelling and experimental point of view. In addition, RHR and SGTR sequences were simulated with SPARC90 and BUSCA-AUG92 codes, and their results were compared with those obtained with MAAP 3.0B. As a result of this work, model capabilities and shortcomings have been assessed and some areas susceptible of further research have been identified. (Author) 73 refs

  10. Review and assessment of pool scrubbing models

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L.E.; Escudero, M.J.; Peyres, V.; Polo, J.; Lopez, J.

    1996-07-01

    Decontamination of fission products bearing bubbles as they pass through aqueous pools becomes a crucial phenomenon for source term evaluation of hypothetical risk dominant sequences of Light Water Reactors. In the present report a peer review and assessment of models encapsulated in SPARC and BUSCA codes is presented. Several aspects of pool scrubbing have been addressed: particle removal, fission product vapour retention and bubble hydrodynamics. Particular emphasis has been given to the close link between retention and hydrodynamics, from both modelling and experimental point of view. In addition, RHR and SGTR sequences were simulated with SPARC90 and BUSCA-AUG92 codes, and their results were compared with those obtained with MAAP 3.0B. As a result of this work, model capabilities and shortcomings have been assessed and some areas susceptible of further research have been identified. (Author) 73 refs.

  11. Pricing Electricity in Pools With Wind Producers

    DEFF Research Database (Denmark)

    Morales González, Juan Miguel; Conejo, A. J.; Kai Liu

    2012-01-01

    This paper considers an electricity pool that includes a significant number of wind producers and is cleared through a network-constrained auction, one day in advance and on an hourly basis. The hourly auction is formulated as a two-stage stochastic programming problem, where the first stage...... represents the clearing of the market and the second stage models the system operation under a number of plausible wind production realizations. This formulation co-optimizes energy and reserve, and allows deriving both pool energy prices and balancing energy prices. These prices result in both cost recovery...... for producers and revenue reconciliation. A case study of realistic size is used to illustrate the functioning of the proposed pricing scheme....

  12. Hot conditioning equipment conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Bradshaw, F.W., Westinghouse Hanford

    1996-08-06

    This report documents the conceptual design of the Hot Conditioning System Equipment. The Hot conditioning System will consist of two separate designs: the Hot Conditioning System Equipment; and the Hot Conditioning System Annex. The Hot Conditioning System Equipment Design includes the equipment such as ovens, vacuum pumps, inert gas delivery systems, etc.necessary to condition spent nuclear fuel currently in storage in the K Basins of the Hanford Site. The Hot Conditioning System Annex consists of the facility of house the Hot Conditioning System. The Hot Conditioning System will be housed in an annex to the Canister Storage Building. The Hot Conditioning System will consist of pits in the floor which contain ovens in which the spent nuclear will be conditioned prior to interim storage.

  13. Hot conditioning equipment conceptual design report

    International Nuclear Information System (INIS)

    Bradshaw, F.W.

    1996-01-01

    This report documents the conceptual design of the Hot Conditioning System Equipment. The Hot conditioning System will consist of two separate designs: the Hot Conditioning System Equipment; and the Hot Conditioning System Annex. The Hot Conditioning System Equipment Design includes the equipment such as ovens, vacuum pumps, inert gas delivery systems, etc.necessary to condition spent nuclear fuel currently in storage in the K Basins of the Hanford Site. The Hot Conditioning System Annex consists of the facility of house the Hot Conditioning System. The Hot Conditioning System will be housed in an annex to the Canister Storage Building. The Hot Conditioning System will consist of pits in the floor which contain ovens in which the spent nuclear will be conditioned prior to interim storage

  14. Pool type liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Guthrie, B.M.

    1978-08-01

    Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

  15. Condensation of vapor bubble in subcooled pool

    Science.gov (United States)

    Horiuchi, K.; Koiwa, Y.; Kaneko, T.; Ueno, I.

    2017-02-01

    We focus on condensation process of vapor bubble exposed to a pooled liquid of subcooled conditions. Two different geometries are employed in the present research; one is the evaporation on the heated surface, that is, subcooled pool boiling, and the other the injection of vapor into the subcooled pool. The test fluid is water, and all series of the experiments are conducted under the atmospheric pressure condition. The degree of subcooling is ranged from 10 to 40 K. Through the boiling experiment, unique phenomenon known as microbubble emission boiling (MEB) is introduced; this phenomenon realizes heat flux about 10 times higher than the critical heat flux. Condensation of the vapor bubble is the key phenomenon to supply ambient cold liquid to the heated surface. In order to understand the condensing process in the MEB, we prepare vapor in the vapor generator instead of the evaporation on the heated surface, and inject the vapor to expose the vapor bubble to the subcooled liquid. Special attention is paid to the dynamics of the vapor bubble detected by the high-speed video camera, and on the enhancement of the heat transfer due to the variation of interface area driven by the condensation.

  16. Gene pool conservation of teak in Myanmar

    International Nuclear Information System (INIS)

    Tin-Tun

    1995-01-01

    Myanmar with an area of 261, 228 Sq. miles is endowed with various types of forests which occupied nearly 50% of the country. Teak (Tectona grandis Linn. f.) is one of the most valuable timber species for its excellent wood quality and properties which are not observed with other timbers. Gene pool can be defined as a group of individual trees growing over a wide range of environmental conditions, and constituting different genetic complexes which can be transmitted to the offsprings. Topics such as: objectives of gene pool conservation, genetically improved seeds for large scale forest plantations, methodology of conservation, are discussed in the article. Myanmar teak dominates the world's teak market, and thus it is crucial to maintain the superiority in the conservation of gene complexes of teak. To some extent, the conservation of gene pools of teak and tree improvements are being undertaken by the Forest Research Institute of Myanmar. It is felt that the dissemination of the philosophy and concept of gene conservation to the personal involved in the forestry activities of the country are still inadequate

  17. Validation of the coupled neutron kinetic thermohydraulic code ATHLET/DYN3D with help of measured data of the OECD Turbine Trip Benchmarks. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.

    2003-12-01

    The project consisted in the validation of the coupled neutron kinetic/thermal hydraulic code system ATHLET/DYN3D for boiling water reactors by the participation at the OECD/NRC turbine trip benchmark. The benchmark defined by the OECD and the American NRC is based on an experiment with closure of the turbine stop valve which was carried out in 1977 in the nuclear power plant Peach Bottom 2 within the framework of a series of 3 experiments. In the experiment, the closure of the valve caused a pressure wave which propagated with attenuation into the reactor core. The condensation of steam in the reactor core caused by the increase of pressure lead to a positive reactivity insertion. The following rise of power was limited by the feedback and the insertion of the control rods. In the frame of the benchmark, the codes could be validated by comparisons with the measured results and the result of the other participants. The benchmark was divided into 3 phases or exercises. Phase I was used for checking the thermo-hydraulic model of the system using a given power release in the core. In phase II, three-dimensional core calculations were performed for given thermal-hydraulic boundary conditions. Coupled calculations were carried out for the selected experiment and four extreme scenarios in the phase III. In the frame of the project, FZR took part in phases II and III of the benchmark. The calculations for phase II were performed with DYN3D by using the assembly discontinuity factors (ADF) and 764 thermal-hydraulic channels (1 channel/assembly). The ATHLET input data set for the coolant system was obtained form the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). It was slightly modified for the phase III calculations carried out with the parallel coupling of ATHLET and DYN3D. For spatially averaged parameters, a good agreement with the results of measurement and the results of other codes was achieved. The influence of the different models was investigated with the

  18. POOL WATER TREATMENT AND COOLING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    King, V.

    2000-01-01

    The Pool Water Treatment and Cooling System is located in the Waste Handling Building (WHB), and is comprised of various process subsystems designed to support waste handling operations. This system maintains the pool water temperature within an acceptable range, maintains water quality standards that support remote underwater operations and prevent corrosion, detects leakage from the pool liner, provides the capability to remove debris from the pool, controls the pool water level, and helps limit radiological exposure to personnel. The pool structure and liner, pool lighting, and the fuel staging racks in the pool are not within the scope of the Pool Water Treatment and Cooling System. Pool water temperature control is accomplished by circulating the pool water through heat exchangers. Adequate circulation and mixing of the pool water is provided to prevent localized thermal hotspots in the pool. Treatment of the pool water is accomplished by a water treatment system that circulates the pool water through filters, and ion exchange units. These water treatment units remove radioactive and non-radioactive particulate and dissolved solids from the water, thereby providing the water clarity needed to conduct waste handling operations. The system also controls pool water chemistry to prevent advanced corrosion of the pool liner, pool components, and fuel assemblies. Removal of radioactivity from the pool water contributes to the project ALARA (as low as is reasonably achievable) goals. A leak detection system is provided to detect and alarm leaks through the pool liner. The pool level control system monitors the water level to ensure that the minimum water level required for adequate radiological shielding is maintained. Through interface with a demineralized water system, adequate makeup is provided to compensate for loss of water inventory through evaporation and waste handling operations. Interface with the Site Radiological Monitoring System provides continuous

  19. State space model extraction of thermohydraulic systems – Part II: A linear graph approach applied to a Brayton cycle-based power conversion unit

    International Nuclear Information System (INIS)

    Uren, Kenneth Richard; Schoor, George van

    2013-01-01

    This second paper in a two part series presents the application of a developed state space model extraction methodology applied to a Brayton cycle-based PCU (power conversion unit) of a PBMR (pebble bed modular reactor). The goal is to investigate if the state space extraction methodology can cope with larger and more complex thermohydraulic systems. In Part I the state space model extraction methodology for the purpose of control was described in detail and a state space representation was extracted for a U-tube system to illustrate the concept. In this paper a 25th order nonlinear state space representation in terms of the different energy domains is extracted. This state space representation is solved and the responses of a number of important states are compared with results obtained from a PBMR PCU Flownex ® model. Flownex ® is a validated thermo fluid simulation software package. The results show that the state space model closely resembles the dynamics of the PBMR PCU. This kind of model may be used for nonlinear MIMO (multi-input, multi-output) type of control strategies. However, there is still a need for linear state space models since many control system design and analysis techniques require a linear state space model. This issue is also addressed in this paper by showing how a linear state space model can be derived from the extracted nonlinear state space model. The linearised state space model is also validated by comparing the state space model to an existing linear Simulink ® model of the PBMR PCU system. - Highlights: • State space model extraction of a pebble bed modular reactor PCU (power conversion unit). • A 25th order nonlinear time varying state space model is obtained. • Linearisation of a nonlinear state space model for use in power output control. • Non-minimum phase characteristic that is challenging in terms of control. • Models derived are useful for MIMO control strategies

  20. Current status of JAERI Tokai hot cell facilities

    International Nuclear Information System (INIS)

    Itami, Hiroharu; Morozumi, Minoru; Yamahara, Takeshi

    1992-01-01

    JAERI has 4 hot cell facilities in order to examine high radioactive materials. Three of them, the Research Hot Laboratory, the Reactor Fuel Examination Facility and the Waste Safety Testing Facility are located in the JAERI Tokai site, and the rest is the JMTR Hot Laboratory in the Oarai site. The Research Hot Laboratory (RHL) was constructed for post-irradiation examination (PIE), especially nuclear related basic research experiment, such as metallurgical, chemical and mechanical examination on fuels and materials irradiated in research and test reactors. This facility has 10 large dimension concrete and 38 lead cells. At present the RHL is used for various kinds of examinations of high radioactive samples such as fuels of research and test reactors, power reactors and high temperature testing reactor (HTTR), and structural materials. The Reactor Fuel Examination Facility (RFEF) was designed and constructed for carrying out PIE of irradiated full-size fuel assemblies of light water reactors (LWRs). This facility has a storage pool, 8 concrete and 5 lead cells. They are currently used for safety evaluation on high burnup and advanced lWR fuels as part of the national program. The Waste Safety Testing Facility (WASTEF) was designed and constructed for safety research on long-term storage and disposal of high level radioactive wastes, generated by fuel reprocessing. The WASTEF has 5 concrete cells and 1 lead cell. Examinations on the behavior of various long-lived fission products in a glass form and in a canister and, releasing behavior of them out of a canister are carrying out under the condition at storage. (author)

  1. Solar Technician Program Blows Hot

    Science.gov (United States)

    Ziegler, Peg Moran

    1977-01-01

    A training program for solar heating technicians was initiated at Sonoma State College's School of Environmental Studies for CETA applicants. Among the projects designed and built were a solar alternative energy center, a solar hot water system, and a solar greenhouse. (MF)

  2. The design of hot laboratories

    International Nuclear Information System (INIS)

    1976-01-01

    The need for specialized laboratories to handle radioactive substances of high activity has increased greatly due to the expansion of the nuclear power industry and the widespread use of radioisotopes in scientific research and technology. Such laboratories, which are called hot laboratories, are specially designed and equipped to handle radioactive materials of high activity, including plutonium and transplutonium elements. The handling of plutonium and transplutonium elements presents special radiation-protection and safety problems because of their high specific activity and high radiotoxicity. Therefore, the planning, design, construction and operation of hot laboratories must meet the stringent safety, containment, ventilation, shielding, criticality control and fire-protection requirements. The IAEA has published two manuals in its Safety Series, one on the safety aspects of design and equipment of hot laboratories (SS No.30) and the other on the safe handling of plutonium (SS No.39). The purpose of the symposium in Otaniemi was to collect information on recent developments in the safety features of hot laboratories and to review the present state of knowledge. A number of new developments have taken place as the result of growing sophistication in the philosophy of radiation protection as given in the ICRP recommendations (Report No.22) and in the Agency's basic safety standards (No.9). The topics discussed were safety features of planning and design, air cleaning, transfer and transport systems, criticality control, fire protection, radiological protection, waste management, administrative arrangements and operating experience

  3. Interfaces in hot gauge theory

    CERN Document Server

    Bronoff, S.

    1996-01-01

    The string tension at low T and the free energy of domain walls at high T can be computed from one and the same observable. We show by explicit calculation that domain walls in hot Z(2) gauge theory have good thermodynamical behaviour. This is due to roughening of the wall, which expresses the restoration of translational symmetry.

  4. Was the big bang hot

    International Nuclear Information System (INIS)

    Wright, E.L.

    1983-01-01

    The author considers experiments to confirm the substantial deviations from a Planck curve in the Woody and Richards spectrum of the microwave background, and search for conducting needles in our galaxy. Spectral deviations and needle-shaped grains are expected for a cold Big Bang, but are not required by a hot Big Bang. (Auth.)

  5. A new hot pressing technique

    International Nuclear Information System (INIS)

    Carcey, J.

    1975-01-01

    An original hot pressing method which may be applied to ceramics, metals, and refractory powders is described. The products obtained are fine grained polycristalline materials, with homogeneous structure, very high density, unstrained and of very large dimensions (several square meters). This process equally applies to composite materials including powders, fibers, etc.. [fr

  6. Hot atom chemistry of carbon

    International Nuclear Information System (INIS)

    Wolf, A.P.

    1975-01-01

    The chemistry of energetic carbon atoms is discussed. The experimental approach to studies that have been carried out is described and the mechanistic framework of hot carbon atom reactions is considered in some detail. Finally, the direction that future work might take is examined, including the relationship of experimental to theoretical work. (author)

  7. Detection of Hot Halo Gets Theory Out of Hot Water

    Science.gov (United States)

    2006-02-01

    Scientists using NASA's Chandra X-ray Observatory have detected an extensive halo of hot gas around a quiescent spiral galaxy. This discovery is evidence that galaxies like our Milky Way are still accumulating matter from the gradual inflow of intergalactic gas. "What we are likely witnessing here is the ongoing galaxy formation process," said Kristian Pedersen of the University of Copenhagen, Denmark, and lead author of a report on the discovery. Chandra observations show that the hot halo extends more than 60,000 light years on either side of the disk of the galaxy known as NGC 5746. The detection of such a large halo alleviates a long-standing problem for the theory of galaxy formation. Spiral galaxies are thought to form from enormous clouds of intergalactic gas that collapse to form giant, spinning disks of stars and gas. Chandra X-ray Image of NGC 5746 Chandra X-ray Image of NGC 5746 One prediction of this theory is that large spiral galaxies should be immersed in halos of hot gas left over from the galaxy formation process. Hot gas has been detected around spiral galaxies in which vigorous star formation is ejecting matter from the galaxy, but until now hot halos due to infall of intergalactic matter have not been detected. "Our observations solve the mystery of the missing hot halos around spiral galaxies," said Pedersen. "The halos exist, but are so faint that an extremely sensitive telescope such as Chandra is needed to detect them." DSS Optical Image of NGC 5746 DSS Optical Image of NGC 5746 NGC 5746 is a massive spiral galaxy about a 100 million light years from Earth. Its disk of stars and gas is viewed almost edge-on. The galaxy shows no signs of unusual star formation, or energetic activity from its nuclear region, making it unlikely that the hot halo is produced by gas flowing out of the galaxy. "We targeted NGC 5746 because we thought its distance and orientation would give us the best chance to detect a hot halo caused by the infall of

  8. Infections Unlikely to be Spread Through Swimming Pools

    Science.gov (United States)

    ... Water Home Infections Unlikely to be Spread Through Swimming Pools Language: English (US) Español (Spanish) Recommend on ... included below. Infections Unlikely to be Spread by Swimming Pools Head Lice Head lice are unlikely to ...

  9. Thermal analyses of solar swimming pool heating in Pakistan

    International Nuclear Information System (INIS)

    Ahmad, I.

    2011-01-01

    Hotels and swimming clubs in Pakistan pay huge gas bills for heating Swimming pools in winter. Winter days in most parts of Pakistan remain sunny and unglazed low cost solar collectors may be used to extend the swimming season. Installing the pool in a wind-protected area, which receives unobstructed solar radiation, may further reduce the size of the solar collectors required to heat the swimming pools. The pools should be covered with plastic sheet to eliminate evaporative heat losses and to prevent dust and tree leaves falling in the pool. The results of the thermal analysis show that in some parts of the country, a solar exposed pool can maintain comfortable temperature simply by using a plastic sheet on the pool surface. On the other hand, there are cities where solar collector array equal to twice the surface area of the pool is required to keep desired temperature in winter. (author)

  10. Analysis of sodium pool fire in SFEF for assessing the limiting pool fire

    International Nuclear Information System (INIS)

    Mangarjuna Rao, P.; Ramesh, S.S.; Nashine, B.K.; Kasinathan, N.; Chellapandi, P.

    2011-01-01

    Accidental sodium leaks and resultant sodium fires in Liquid Metal Fast Breeder Reactor (LMFBR) systems can create a threat to the safe operation of the plant. To avoid this defence-in depth approach is implemented from the design stage of reactor itself. Rapid detection of sodium leak and fast dumping of the sodium into the storage tank of a defective circuit, leak collection trays, adequate lining of load bearing structural concrete and extinguishment of the sodium fire are the important defensive measures in the design, construction and operation of a LMFBR for protection against sodium leaks and their resultant fires. Evaluation of sodium leak events and their consequences by conducting large scale engineering experiments is very essential for effective implementation of the above protection measures for sodium fire safety. For this purpose a Sodium Fire Experimental Facility (SFEF) is constructed at SED, IGCAR. SFEF is having an experimental hall of size 9 m x 6 m x 10 m with 540 m 3 volume and its design pressure is 50 kPa. It is a concrete structure and provided with SS 304 liner, which is fixed to the inside surfaces of walls, ceiling and floor. A leak tight door of size (1.8 m x 2.0 m) is provided to the experimental hall and the facility is provided with a sodium equipment hall and a control room. Experimental evaluation of sodium pool fire consequences is an important activity in the LMFBR sodium fire safety related studies. An experimental program has been planned for different types of sodium fire studies in SFEF. A prior to that numerical analysis have been carried out for enclosed sodium pool fires using SOFIRE-II sodium pool fire code for SFEF experimental hall configuration to evaluate the limiting pool fire. This paper brings out results of the analysis carried out for this purpose. Limiting pool fire of SFEF depends on the exposed surface area of the pool, amount of sodium in the pool, oxygen concentration and initial sodium temperature. Limiting

  11. (Important hygienic aspects for swimming pools (author's transl))

    Energy Technology Data Exchange (ETDEWEB)

    Somosi, G

    1981-01-01

    The major epidemics which occurred in Hungary and originated from water in swimming pools are reported. The difficulties encountered in producing epidemiological evidence and in monitoring infections originating from water in swimming pools are mentioned. The possibilities of controlling the water quality in swimming pools and of preventing infections are discussed. Reference is made to the existing bacteriological limit values in Hungary to be observed in the recirculation of water in swimming pools.

  12. NACUBO's Guide to Unitizing Investment Pools. Second Edition

    Science.gov (United States)

    Wheeler, Mary S.

    2011-01-01

    The National Association of College and University Business Officers' (NACUBO's) "Guide to Unitizing Investment Pools" addresses the principles and concepts for administering a consolidated investment pool. Unitization is the mechanism by which investment funds are pooled to maximize investment efficiencies and provide information for donors,…

  13. A skeleton for distributed work pools in Eden

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Berthold, Jost; Loogen, Rita

    2010-01-01

    We present a flexible skeleton for implementing distributed work pools in our parallel functional language Eden. The skeleton manages a pool of tasks (work pool) in a distributed manner using a demand-driven work stealing approach for load balancing. All coordination is done locally within...

  14. Opportunities and challenges when pooling milk samples using ELISA

    DEFF Research Database (Denmark)

    Græsbøll, Kaare; Andresen, Lars Ole; Hisham Beshara Halasa, Tariq

    2017-01-01

    -positive samples by pooling. To illustrate this, the sensitivity of antibody ELISA on pooled samples of bovine milk for Salmonella Dublin, Mycobacterium avium spp. paratuberculosis, and bovine virus diarrhea was tested. For these milk assays, the analytical sensitivity decreased rapidly with increasing pool sizes...

  15. 17 CFR 4.22 - Reporting to pool participants.

    Science.gov (United States)

    2010-04-01

    ..., the financial statements are not required to include consolidated information for all series. (7) For... event that the International Financial Reporting Standards require consolidated financial statements for... reporting pool's consolidated financial statements. (ii) The commodity pool operator of a pool that meets...

  16. 48 CFR 873.114 - Best value pool.

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Best value pool. 873.114... SUPPLEMENTARY REGULATIONS SIMPLIFIED ACQUISITION PROCEDURES FOR HEALTH-CARE RESOURCES 873.114 Best value pool... solicitation. These vendors constitute the best value pool. This determination is within the sole discretion of...

  17. Hot Flashes amd Night Sweats (PDQ)

    Science.gov (United States)

    ... Professionals Questions to Ask about Your Treatment Research Hot Flashes and Night Sweats (PDQ®)–Patient Version Overview ... quality of life in many patients with cancer. Hot flashes and night sweats may be side effects ...

  18. Hot Laboratories and Remote Handling

    International Nuclear Information System (INIS)

    Bart, G.; Blanc, J.Y.; Duwe, R.

    2003-01-01

    The European Working Group on ' Hot Laboratories and Remote Handling' is firmly established as the major contact forum for the nuclear R and D facilities at the European scale. The yearly plenary meetings intend to: - Exchange experience on analytical methods, their implementation in hot cells, the methodologies used and their application in nuclear research; - Share experience on common infrastructure exploitation matters such as remote handling techniques, safety features, QA-certification, waste handling; - Promote normalization and co-operation, e.g., by looking at mutual complementarities; - Prospect present and future demands from the nuclear industry and to draw strategic conclusions regarding further needs. The 41. plenary meeting was held in CEA Saclay from September 22 to 24, 2003 in the premises and with the technical support of the INSTN (National Institute for Nuclear Science and Technology). The Nuclear Energy Division of CEA sponsored it. The Saclay meeting was divided in three topical oral sessions covering: - Post irradiation examination: new analysis methods and methodologies, small specimen technology, programmes and results; - Hot laboratory infrastructure: decommissioning, refurbishment, waste, safety, nuclear transports; - Prospective research on materials for future applications: innovative fuels (Generation IV, HTR, transmutation, ADS), spallation source materials, and candidate materials for fusion reactor. A poster session was opened to transport companies and laboratory suppliers. The meeting addressed in three sessions the following items: Session 1 - Post Irradiation Examinations. Out of 12 papers (including 1 poster) 7 dealt with surface and solid state micro analysis, another one with an equally complex wet chemical instrumental analytical technique, while the other four papers (including the poster) presented new concepts for digital x-ray image analysis; Session 2 - Hot laboratory infrastructure (including waste theme) which was

  19. Modeling of condensation, stratification, and mixing phenomena in a pool of water

    Energy Technology Data Exchange (ETDEWEB)

    Li, H.; Kudinov, P.; Villanueva, W. (Royal Institute of Technology (KTH). Div. of Nuclear Power Safety, Stockholm (Sweden))

    2010-12-15

    This work pertains to the research program on Containment Thermal-Hydraulics at KTH. The objective is to evaluate and improve performance of methods, which are used to analyze thermal-hydraulics of steam suppression pools in a BWR plant under different abnormal transient and accident conditions. As a passive safety system, the function of steam pressure suppression pools is paramount to the containment performance. In the present work, the focus is on apparently-benign but intricate and potentially risk-significant scenarios in which thermal stratification could significantly impede the pool's pressure suppression capacity. For the case of small flow rates of steam influx, the steam condenses rapidly in the pool and the hot condensate rises in a narrow plume above the steam injection plane and spreads into a thin layer at the pool's free surface. When the steam flow rate increases significantly, momentum introduced by the steam injection and/or periodic expansion and shrink of large steam bubbles due to direct contact condensation can cause breakdown of the stratified layers and lead to mixing of the pool water. Accurate prediction of the pool thermal-hydraulics in such scenarios presents a computational challenge. Lumped-parameter models have no capability to predict temperature distribution of water pool during thermal stratification development. While high-order-accurate CFD (RANS, LES) methods are not practical due to excessive computing power needed to calculate 3D high-Rayleighnumber natural circulation flow in long transients. In the present work, a middleground approach is used, namely CFD-like model of the general purpose thermalhydraulic code GOTHIC. Each cell of 3D GOTHIC grid uses lumped parameter volume type closures for modeling of various heat and mass transfer processes at subgrid scale. We use GOTHIC to simulate POOLEX/PPOOLEX experiment, in order to (a) quantify errors due to GOTHIC's physical models and numerical schemes, and (b

  20. OUT Success Stories: Solar Hot Water Technology

    International Nuclear Information System (INIS)

    Clyne, R.

    2000-01-01

    Solar hot water technology was made great strides in the past two decades. Every home, commercial building, and industrial facility requires hot water. DOE has helped to develop reliable and durable solar hot water systems. For industrial applications, the growth potential lies in large-scale systems, using flat-plate and trough-type collectors. Flat-plate collectors are commonly used in residential hot water systems and can be integrated into the architectural design of the building