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Sample records for pwr primary cooling

  1. Tribological study of hard coatings without cobalt intended to isolation components of PWR primary cooling system; Etude tribologique de revetements durs sans cobalt destines aux organes d`isolement du circuit primaire des REP

    Energy Technology Data Exchange (ETDEWEB)

    Cachon, L.

    1995-10-18

    The objective is to qualify coatings without cobalt to replace ``Stellites`` coatings in isolation valves of PWR primary cooling system, as Co is activated when passing in the reactor core and contaminated the cooling loop. Three families of coatings were tested: PVD thin films from 1 to 8 {mu}m monolayers of Cr/C{sub x} with x varying between 1.6 and 9.5 at% or multilayers of pure chromium and Cr/C{sub 1.6} at%, coatings with a thickness between 100 and 200 {mu}m of cermets NiCr{sub y} (y varying from 5 to 35 at%) matrix binding chromium or tungsten carbides, and thick coatings 2 mm thickness of cermets Nitronic 60 or Inconel 625 matrix binding 10, 20 or 30% titanium or niobium carbides. Stellite 6 (2 mm) is the reference coating for tribology. Coatings were qualified and selected by thermal shocks, corrosion and plane friction. The thin film and the thick families were disqualified by their destruction or by their high friction coefficient. Then coatings between 100 and 200 {mu}m were used in a valve mock-up working in PWR primary cooling system pressure and temperature conditions. Tests show that these coatings have better wear or tightness performances than stellite 6, except for a slightly higher friction coefficient. (A.B.).

  2. EPRI PWR primary water chemistry guidelines revision

    International Nuclear Information System (INIS)

    McElrath, Joel; Fruzzetti, Keith

    2014-01-01

    EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

  3. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author) [pt

  4. Natural-circulation-cooling characteristics during PWR accident simulations

    International Nuclear Information System (INIS)

    Adams, J.P.; McCreery, G.E.; Berta, V.T.

    1983-01-01

    A description of natural circulation cooling characteristics is presented. Data were obtained from several pressurized water reactor accident simulations in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). The reliability of natural circulation cooling, its cooling effectiveness, and the effect of changing system conditions are described. Quantitative comparison of flow rates and time constants with theory for both single- and two-phase fluid conditions were made. It is concluded that natural circulation cooling can be relied on in plant recovery procedures in the absence of forced convection whenever the steam generator heat sink is available

  5. Simplified model of a PWR primary circuit

    International Nuclear Information System (INIS)

    Souza, A.L.; Faya, A.J.G.

    1988-07-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analyzed by a nodal model. Average and hot channels are treated so that bulk response of the core and DNBR can be evaluated. A homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  6. A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

    International Nuclear Information System (INIS)

    Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

    2002-01-01

    A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

  7. Measurement of mist cooling of PWR during LOCA by LDA

    International Nuclear Information System (INIS)

    Lee, S.L.; Sheen, H.J.; Issapour, I.

    1985-01-01

    The prediction of temperature distribution and heat transfer within rod bundles during the refill and reflood phase of a LOCA (loss of coolant accident) is of critical importance for determining the location and size of blockages due to clad deformation in a pressurized water reactor (PWR). Mist cooling by small droplets generated from large droplets on hitting grid spacers has been suggested as one of the most important heat transfer mechanisms which are responsible for the development of this temperature transient. The questions to be asked are whether such small droplets indeed exist and, if so, how are they related to the cooling of the fuel rods. Hereby reported is the result of a direct experimental investigation on these questions by a special laser-Doppler anemometry (LDA) particle sizing technique together with temperature measurements of the rod claddings and flow in the subchannel

  8. Behavior of a nine-rod PWR bundle under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; Sparks, D.T.

    1979-01-01

    An experiment to characterize the behavior of a nine-rod pressurized water reactor (PWR) fuel bundle operating during power-cooling-mismatch (PCM) conditions has been conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The experiment, designated Test PCM-5, is part of a series of PCM experiments designed to evaluate light water reactor (LWR) fuel rod response under postulated accident conditions. Test PCM-5 was the first nine-rod bundle experiment in the PCM test series. The primary objectives and the results of the experiment are described

  9. Optimal design of passive containment cooling system for innovative PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon [Central Research Institute, Korea Hydro and Nuclear Power, Ltd., Daejeon (Korea, Republic of)

    2017-08-15

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  10. Optimal design of passive containment cooling system for innovative PWR

    Directory of Open Access Journals (Sweden)

    Huiun Ha

    2017-08-01

    Full Text Available Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS of an innovative pressurized water reactor (PWR. A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT geometry, PCCS heat exchanger (PCCX location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  11. Optimal design of passive containment cooling system for innovative PWR

    International Nuclear Information System (INIS)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon

    2017-01-01

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed

  12. Thermohydraulic calculations of PWR primary circuits

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1984-01-01

    Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

  13. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, V; Rosenberg, R [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  15. Data assimilation and PWR primary measurement

    International Nuclear Information System (INIS)

    Mercier, Thibaud

    2015-01-01

    A Pressurized Water Reactor (PWR) Reactor Coolant System (RCS) is a highly complex physical process: heterogeneous power, flow and temperature distributions are difficult to be accurately measured, since instrumentations are limited in number, thus leading to the relevant safety and protection margins. EDF R and D is seeking to assess the potential benefits of applying Data Assimilation to a PWR's RCS (Reactor Coolant System) measurements, in order to improve the estimators for parameters of a reactor's operating setpoint, i.e. improving accuracy and reducing uncertainties and biases of measured RCS parameters. In this thesis, we define a 0D semi-empirical model for RCS, satisfying the description level usually chosen by plant operators, and construct a Monte-Carlo Method (inspired from Ensemble Methods) in order to use this model with Data Assimilation tools. We apply this method on simulated data in order to assess the reduction of uncertainties on key parameters: results are beyond expectations, however strong hypothesis are required, implying a careful preprocessing of input data. (author)

  16. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  17. Effect of TOC [total organic carbon] on a PWR secondary cooling water system

    International Nuclear Information System (INIS)

    Gau, J.Y.; Oung, J.C.; Wang, T.Y.

    1989-01-01

    Increasing the amount of total organic carbon (TOC) during the wet layup of the steam generator was a problem in PWR nuclear power plant in Taiwan. The results of surveys of TOC in PWR secondary cooling water systems had shown that the impurity of hydrazine and the bacteria were the main reasons that increase TOC. These do not have a corrosion effect on Inconel 600 and carbon steel when the secondary cooling water containing the TOC is below 200 ppb. But the anaerobic bacteria from the steam generator in wet layup will increase corrosion rate of carbon steel and crevice corrosion of Inconel 600. (author)

  18. Substitution of cobalt alloying in PWR primary circuit gate valves

    International Nuclear Information System (INIS)

    Cachon, L.; Sudreau, F.; Brunel, L.

    1995-01-01

    The object of this study is qualify cobalt-free alternative alloys for valve applications. This paper focus on tribological characterization of numerous coatings is done by using the first one, of a classical type. Then tests are performed with the second one which simulates solicitations supported by gate valves in primary circuit of PWR. 35% Ni-Cr - 65% Cr 3 C 2 coating, deposited by detonation gun technology, gives us hope to find a substitute of Stelite 6. (author). 5 refs., 16 figs., 2 tabs

  19. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    International Nuclear Information System (INIS)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-01-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years

  20. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  1. SCC of Alloy 600 components in PWR primary loop

    International Nuclear Information System (INIS)

    Gomez-Briceno, Dolores; Lapena, Jesus; Castano, M. Luisa; Blazquez, Fernando

    2002-01-01

    Full text: Cracking due to PWSCC in PWR CRDM nozzles and other VHP nozzles fabricated from Alloy 600 is not a new issue. In 1991, a leak was discovered on one CRDM nozzle at Bugey 3 PWR plant in France. The cause of the cracking was identified as primary water stress corrosion cracking. From then, similar cracks have been found in other European and USA PWR plants. The cracks were predominantly axial in orientation and it was accepted that CRDM nozzles and weld cracking in PWR was not a immediate safety concern. However, this consideration has to be reassessed in light of the recent identification of circumferential cracking in CRDM nozzles at Oconee Nuclear Station Unit 2 and 3 along with axial cracking in the Alloy 182 J-groove welds at these two units and at Oconee Nuclear Station 1 and Arkansas Nuclear One Unit 1. Alloy 600 susceptibility in primary water has received an enormous research effort for many years since the Alloy 600 steam generators tube degradation started. A significant amount of information is available to characterise the susceptibility of Alloy 600. However, Alloy 600 susceptibility is strongly dependent on the heat thermomechanical history and both the crack initiation time and the crack growth rate data obtained from representative materials of the VHP nozzles seem to be necessary for the structural integrity assessment of cracking nozzles. An extensive experimental program has been performed at CIEMAT, to study the behaviour of Alloy 600 VHP nozzles in PWR primary conditions. Crack initiation and crack propagation tests have been performed using different types of products (forged bar, tube, plate and steam generator tubing). Long duration crack initiation tests have been carried out, at 330 deg. C and 360 deg. C in water and at 400 deg. C in steam, using ten Alloy 600 heats with yield strength ranging from 291 MPa to 489 MPa. The influence of several parameters (grain boundary carbide distribution, grain size and yield strength) on crack

  2. PWR primary system chemistry control during hot functional testing

    International Nuclear Information System (INIS)

    Reid, Richard D.; Little, Michael J.

    2014-01-01

    Hot Functional Testing (HFT) involves a number of pre-operational exercises performed to confirm the operability of plant systems at conditions expected during both normal and off-normal operation of a pressurized water reactor (PWR), including operability of safety systems. While the primary purposes of HFT are to demonstrate operability of plant systems and satisfy regulatory requirements, chemistry control during HFT is important to long-term integrity and performance of plant systems. Specifically, HFT is the first time plant equipment is exposed to high temperature water and the chemistry maintained during HFT can impact the passivation layers that form on wetted surfaces and long-term release of metals from these surfaces. Metals released from the inner surfaces of steam generator tubing and reactor coolant loop piping become activated in the core and can redeposit on ex-core surfaces. Because HFT is performed before fuel is loaded in the core, HFT provides an opportunity to produce a passive layer on primary surfaces that is free of activated corrosion products, resistant to metals release during subsequent plant operation, and also resistant to incorporation of activated corrosion products (once fuel is loaded in the core). Thus, maintaining desirable primary chemistry control during HFT is important for source term management, minimization of future shutdown activity releases, minimization of dose rates, and asset preservation. This paper presents an overview of passive film formation in the austenitic stainless steel and high nickel alloys that make up the majority of the primary circuit in advanced PWR designs. Based on this information, a summary is provided of the effects on passive film formation of key chemistry parameters that may be controlled during HFT. (author)

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

    International Nuclear Information System (INIS)

    Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

    2014-01-01

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

  4. Shutdown Chemistry Process Development for PWR Primary System

    Energy Technology Data Exchange (ETDEWEB)

    Sung, K.B. [Korea Electric Power Research Institute, Taejeon (Korea, Republic of)

    1997-12-31

    This study report presents the shutdown chemistry of PWR primary system to reduce and remove the radioactive corrosion products which were deposited on the nuclear fuel rods surface and the outside of core like steam generator channel head, RCS pipings etc. The major research results are the follows ; the deposition radioactive mechanism of corrosion products, the radiochemical composition, the condition of coolant chemistry to promote the dissolution of radioactive cobalt and nickel ferrite, the control method of dissolved hydrogen concentration in the coolant by the mechanical and chemical methods. The another part of study is to investigate the removal characteristics of corrosion product ions and particles by the demineralization system to suggest the method which the system could be operate effectively in shut-down purification period. (author). 19 refs., 25 figs., 48 tabs.

  5. Experimental studies of PWR primary piping under loca

    International Nuclear Information System (INIS)

    Caumette, Pierre; Garcia, J.L.

    1980-07-01

    The experimental program performed on AQUITAINE II facility is directed to study the mechanical behavior of primary PWR pipes and the forces exerted on the neighbouring structures as a consequence of a breach opening. It has been developed in the form of a quadripartite agreement between the Commissariat a l'Energie Atomique, Framatome, Electricite de France and Westinghouse. Some forty tests have been carried out with different pipe configurations (straight tube, elbow, S- or U-shaped tube) and different break types (single or double guillotine). The following aspects are investigated: - the dynamic behavior of the pipe and in particular the formation of a plastic hinge at the restraint; - the impact function of a pipe or an energy-absorbing bumper; - the lateral stability of both ends of a pipe, after a double-guillotine break [fr

  6. Simplified model of a PWR primary coolant circuit

    International Nuclear Information System (INIS)

    Souza, A.L. de; Faya, A.J.G.

    1988-01-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analysed by a nodal model. Average and hot channels are treated so that the bulk response of the core and DNBR can be evaluated. A Homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  7. Tendency of nuclear pumps for PWR primary system

    International Nuclear Information System (INIS)

    Shibata, Takeshi

    1976-01-01

    At present, large PWR power stations of more than 1,000 MW are successively constructed, and the pumps used there have become large. The progress and tendency of the technical development of main pumps in primary system are described. The increase of the capacity of power stations is accomplished by increasing the circulating coolant quantity per loop or the number of loops. Same standard primary coolant pumps are employed in the plants from 500 to 1,100 MW. The type of primary coolant pumps changed from canned type to shaft seal type, and the advantages of the shaft seal type are cheap production cost, high efficiency, and the easy utilization of inertia force. The bearings and shaft seals are thermally insulated from primary coolant. As for auxiliary pumps, reciprocating filling-up pumps and centrifugal high pressure injection pumps are used for 500 MW plants, but only centrifugal pumps are used for both purposes in 800 MW plants, and in 1,100 MW plants, the pumps of both types for separate purposes and centrifugal pumps for combined purposes are installed. Horizontal or vertical pumps of same type are used as containment vessel-spraying pumps and excess heat-eliminating pumps. The type of boric acid pumps changed from canned type to mechanical seal type. (Kako, I.)

  8. Assessment of Condenser Design in the Cooling System of PWR

    International Nuclear Information System (INIS)

    Sukmanto Dibyo

    2008-01-01

    This paper explains various cases related to the problem of condenser design principle. Condenser equipment is circuit component of cooling that is very important. Steam that goes down from the turbine, flowing to the condenser and the steam is condensate to be water. Step of design is beginning from identification of problem until mechanic design. Usually, the condenser is designed by conservative certain levels, as well as thermal aspect or mechanic. Beside that, the design analysis is also anticipated to the existing of degradation due to aging therefore safety aspect and requirement are accepted. The assessment that related to the condenser design has been discussed. Most important aspect is obtaining of performances. Beside that, the discussion is also described about design method and that complication should consider aspect of thermal design, physics, economic, aspect of corrosion and others factors. (author)

  9. Dose rate determining factors of PWR primary water

    International Nuclear Information System (INIS)

    Terachi, Takumi; Kuge, Toshiharu; Nakano, Nobuo

    2014-01-01

    The relationship between dose rate trends and water chemistry has been studied to clarify the determining factors on the dose rates. Therefore dose rate trends and water chemistry of 11 PWR plants of KEPCO (Kansai Electric Power Co., Inc.) were summarized. It is indicated that the chemical composition of the oxide film, behaviour of corrosion products and Co-58/Co-60 ratio in the primary system have effected dose rate trends based on plant operation experiences for over 40 years. According to plant operation experiences, the amount of Co-58 has been decreasing with the increasing duration of SG (Steam Generator) usage. It is indicated that the stable oxide film formation on the inner surface of SG tubing, is a major beneficial factor for radiation sources reduction. On the other hand, the reduction of the amount of Co-60 for the long term has been not clearly observed especially in particular high dose plants. The primary water parameters imply that considering release and purification balance on Co-59 is important to prevent accumulation of source term in primary water. In addition, the effect of zinc injection, which relates to the chemical composition of oxide film, was also assessed. As the results, the amount of radioactive Co has been clearly decreased. The decreasing trend seems to correlate to the half-life of Co-60, because it is considered that the injected zinc prevents the uptake of radioactive Co into the oxide film on the inner surface of the components and piping. In this paper, the influence of water chemistry and the replacement experiences of materials on the dose rates were discussed. (author)

  10. Development of forging technology for PWR primary piping

    International Nuclear Information System (INIS)

    Morin, F.; Badeau, J.P.; Lambs, R.

    1996-01-01

    The purpose of this presentation is to give information on the changes in the design and manufacture of Primary Piping for electronuclear boilers of the Pressurized Water Reactor type (PWR) which has resulted in the making of one-piece forged lines including stub pipes and arcs. The optimization of these items is aimed at improving the life of the new power stations as well as guaranteeing their safety, while reducing inspection and maintenance requirements in service. The demonstration of the manufacturing feasibility has just been completed. It has taken material form in the installation, on the CIVAUX 1 section, of the first one-piece cold leg in the world. It will shortly be followed by the installation on the CIVAUX 2 section of a complete loop of bent forged pipes. Therefore, this new know-how is going to be incorporated in the French Rules (RCC-M) and can be directly taken into consideration both in the next work to be done and in the design and definition of a future nuclear reactor

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. An Overview of the EPRI PWR Primary Chemistry Program

    International Nuclear Information System (INIS)

    Perkins, David; Fruzzetti, Keith; Haas, Carey; Wells, Dan

    2012-09-01

    and changed from stainless steel to the various advanced alloys utilized today. The ever present demand to shorten outage durations is compounding the challenges faced in shutdown chemistry. Utilities continue to optimize shutdown and cool down activities for refueling outages supporting outage durations < 30 days and in some cases < 20 days. The industry continues to review plant shutdown strategies including a rapid shutdown and cool down that allows utilities to enter cold shutdown conditions in = 6 hours from hot standby. The EPRI Materials Reliability Program (MRP), Fuel Reliability Program (FRP), Chemistry and Radiation Management (RM) programs have ongoing research related to optimized chemistry supporting industry efforts in dose reduction, mitigation of primary water stress corrosion cracking (PWSCC) in nickel-based alloys and improved fuel reliability. This paper describes the ongoing EPRI pressurized water reactor primary water chemistry research supporting these programs, as well as an update to previous papers presented in the conference series. (authors)

  13. PKL/K9, Refill and Reflood Experiment in a Simulated PWR Primary System (PKL)

    International Nuclear Information System (INIS)

    1981-01-01

    reflood process within a pressure vessel considering the influence of the simulated primary loops of a pressurized water reactor (PWR). As to the number of rods and the cross-sections of vessel and components the PKL test-facility is scaled down to 1:134 of a typical West Germany PWR. The elevations and locations have been designed in order to reproduce substantially the original extension. Compared to other cold leg injection tests in the same facility (K5. 3a, K5. 4a) test K9 differs in a uniform radial power profile in the core, an increased temperature of injected Emergency Core Cooling (ECC) water (feed water, from 35 to 53 deg. C), an injection mass flow rate highly reduced to 1:3, and a history of it corresponding to that of a typical US-PWR, on the average somewhat higher maximum initial temperatures in the bundle, and a smaller bundle heating power during the initial phase

  14. Measurement of grid spacer's enhanced droplet cooling under reflood condition in a PWR by LDA

    International Nuclear Information System (INIS)

    Lee, S.L.; Sheen, H.J.; Cho, S.K.; Issapour, I.; Hua, S.Q.

    1984-01-01

    Reported is an experiment designed for the measurements of grid spacer's enhanced droplet cooling under reflood condition at elevated temperatures in a steam environment. The flow channel consists of a simulated 1.60m-long pressurized water reactor (PWR) fuel rod bundle of 2 x 2 electrically heated rods. Embedded thermocouples are used to measure the rod cladding temperature at various axial levels and an unshielded Chromel-Alumel thermocouple sheathed by a small Inconel tube is traversed in the center of the subchannel to measure the temperatures of the water and steam coolant phases at various levels. The droplet dynamics across the grid spacer is directly obtained by a special laser-Doppler anemometry technique for the in situ simultaneous measurement of velocity and size of droplets through two observation windows on the test channel, one immediately before and one immediately after the grid spacer. Some results are presented and analyzed

  15. Reactivity and isotopic composition of spent PWR [pressurized-water-reactor] fuel as a function of initial enrichment, burnup, and cooling time

    International Nuclear Information System (INIS)

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub ∞/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub ∞/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub ∞/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs

  16. Decontamination of primary cooling system

    International Nuclear Information System (INIS)

    Morikawa, Yoshitake.

    1985-01-01

    Purpose: To effectively eliminate radioactivity accumulated in pipeways, equipments, etc in primary coolant circuits of BWR type power plants by utilizing ion displacement reactions. Method: The reactor pressure vessel is connected with a feedwater pipeway, steam pipeway and a recycling pipeway. The recycling pipeway is disposed with a recycling pump. A recycling by-pass line is branched from the recycling pipeway and disposed with a recycling system heat exchanger and chemical injection point. Water is filled in the primary coolant and heated 280 0 C. Then, while maintaining water at that temperature, non-radioactive cobalt ions are injected and circulated within the system, by which radioactivity accumulated in pipeways, equipments or the likes can effectively be removed. (Horiuchi, T.)

  17. Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

    International Nuclear Information System (INIS)

    Hong, S.Y.; Yeater, M.L.

    1985-01-01

    This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A ''analysis of flaw indication'' for the application to a PWR primary piping. Results of the analysis are discussed in detail. (orig.)

  18. Investigation of chloride-release of nuclear grade resin in PWR primary system coolant

    International Nuclear Information System (INIS)

    Cao Xiaoning; Li Yunde; Li Jinghong; Lin Fangliang

    1997-01-01

    A new preparation technique is developed for making the low-chloride nuclear-grade resin by commercial resin. The chloride remained in nuclear grade resin may release to PWR primary coolant. The amount of released chloride is depended on the concentration of boron, lithium, other anion impurities, and remained chloride concentration in resin

  19. Determination of burnup, cooling time and initial enrichment of PWR spent fuel by use of gamma-ray activity ratios

    International Nuclear Information System (INIS)

    Min, D.K.; Park, H.J.; Park, K.J.; Ro, S.G.; Park, H.S.

    1999-01-01

    The Korea Atomic Energy Institute has been developing the algorithms for sequential determination of cooling time, initial enrichment and burnup of the PWR spent fuel assembly by use of gamma ratio measurements, i.e. 134 Cs/ 137 Cs, 154 Eu/ 137 Cs and 106 Ru 137 Cs/( 134 Cs) 2 . Calculations were performed by applying the ORIGEN-S code. This method has advantages over combination techniques of neutron and gamma measurement, because of its simplicity and insensitivity to the measurement geometry. For verifying the algorithms an experiment for determining the cooling time, initial enrichment and burnup of the two PWR spent fuel rods was conducted by use of high-resolution gamma detector (HPGe) system only. This paper describes the method used and interim results of the experiment. This method can be applied for spent fuel characterization, burnup credit and safeguards of the spent fuel management facility

  20. Study on transient hydrogen behavior and effect on passive containment cooling system of the advanced PWR

    International Nuclear Information System (INIS)

    Wang Yan

    2014-01-01

    A certain amount of hydrogen will be generated due to zirconium-steam reaction or molten corium concrete interaction during severe accidents in the pressurized water reactor (PWR). The generated hydrogen releases into the containment, and the formed flammable mixture might cause deflagration or detonation to produce high thermal and pressure loads on the containment, which may threaten the integrity of the containment. The non-condensable hydrogen in containment may also reduce the steam condensation on the containment surface to affect the performance of the passive containment cooling system (PCCS). To study the transient hydrogen behavior in containment with the PCCS performance during the accidents is significant for the further study on the PCCS design and the hydrogen risk mitigation. In this paper, a new developed PCCS analysis code with self-reliance intellectual property rights, which had been validated by comparison on the transients in the containment during the design basis accidents with other developed PCCS analysis code, is brief introduced and used for the transient simulation in the containment under a postulated small break LOCA of cold-leg. The results show that the hydrogen will flow upwards with the coolant released from the break and spread in the containment by convection and diffusion, and it results in the increase of the pressure in the containment due to reducing the heat removal capacity of the PCCS. (author)

  1. NEPTUN/5052, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test

    International Nuclear Information System (INIS)

    Richner, M.; Analytis, G.Th.; Aksan, S.N.

    1993-01-01

    1 - Description of test facility: NEPTUN is designed to perform PWR LOCA simulation experiments, which provide the full length emergency cooling heat transfer tests for LOFT. Therefore the NEPTUN heater bundle with 33 electrical heater elements and 4 guide tubes simulates a section of the LOFT nuclear core. The main test loop also contains measuring systems for the carry-over rate and for the steam expelled, and a back-pressure control system. A water loop brings the water to the initial reflooding conditions. In addition, auxiliary systems maintain normal operating conditions. 2 - Description of test: Test 5052 is one of a series of 40 reflood tests performed in NEPTUN. Before the start of the test, the flooding water in its circuit is brought to the following conditions: pressure = 4.1 bar; velocity = 2.5 cm/sec; subcooling temperature = 78 C; single rod power = 2.45 kW; maximal initial cladding temperature = 867 C. 3 - Status: CSNI1013/01, 21-Jul-1993 Arrived at NEADB

  2. The deformation of Zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-08-01

    Simulated PWR fuel rods clad with Zircaloy-4 were tested under convective steam cooling conditions, by pressurising to 0.69-2.07MPa (100-300lb/in 2 ), then ramping at 10 0 C/s to various temperatures in the region 800-955 0 C and holding until either 600 s elapsed or rupture occurred. The length of cladding strained 33% or more was greatest (about 20 times the original diameter) when the initial internal pressure was 1.38+-0.17 PMa (200+-25lb/in 2 ), and the temperature 885 0 C. It is thought that this results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilising the deformation and/or partial superplastic deformation. To avoid adjacent rods in a fuel assembly touching at any temperature, the pressure would have to be less than about 1MPa (145 1b/in 2 ). If the pressure was 1.38MPa (200lb/in 2 ) then the rods would not swell sufficiently to touch if the temperature did not exceed about 840 0 C. (author)

  3. Study on B-10 consumption of PWR primary coolant during normal operation

    International Nuclear Information System (INIS)

    Liang, C.H.

    1994-01-01

    B-10 consumption under PWR primary coolant conditions has been analyzed. The result indicates its time-dependent change reacting with neutron in the normal operation. In this work, neutron energy assumed to be 4 eV; thermal neutron flux is in the range of 3 x 10 13 to 3 x 10 14 n/sec - cm 2 and the time of cycling of the primary coolant through the RCS is 8 sec. and its retention time in the core region is about 1 sec. Under this condition investigated, B-10 consumption is less than 5% at 3 x 10 13 n/sec - cm 2 thermal neutron flux, and closes to 27% at 3 x 10 14 n/sec - cm 2 by calculation at the 16th month of continuous operation. The effect of B-10 consumption on PWR primary water chemistry is also investigated. (author). 1 fig., 2 tabs., 4 refs

  4. Alternative water chemistry for the primary loop of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Henzel, N [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-02-01

    Advanced fuel element concepts (longer cycles, higher burnup, increased rod power) call for more reactivity binding capacity and, moreover, might produce higher void fractions, particularly in the hot channel. Thus, on the one hand, more alcalizing agent is needed to maintain a high coolant pH according to the approved ``modified boron-lithium mode of operation`` in the presence of more boric acid (chemical shim); on the other hand, increasing enrichment of coolant constituents due to local boiling (higher void fraction), which must not result in accelerated corrosion of fuel cladding and structural materials, imposes enhanced requirements on both, materials technology and water chemistry. At present, the use of boric acid enriched in B10 (the isotope effective in terms of reactivity control) appears to advantageously compromise in capturing more neutrons with less total boron while maintaining or even slightly reducing lithium concentrations at the same time. There is no feasible alternative for boric acid used as the chemical shim and for hydrogen gas as the reducing agent used to suppress oxygen formation by water radiolysis. Systematic screening as performed in phase 1 of a recent project proved potassium hydroxide to be the only potential candidate to favourably replace lithium 7 hydroxide as an alcalizing agent. Unfortunately, the results of pertinent comparative corrosion tests are not unambiguous, and available operational experience with potassium hydroxide in WWER plants is not readily applicable to western world-type PWR plants. Therefore, a switch-over from lithium to potassium can be envisaged only subsequent to a comprehensive qualification program which is planned to be the objective of phase 2 of the project. This program should also comprise zinc addition tests in order to confirm the alleged positive impact of this element on corrosion rates and activity buildup. (Abstract Truncated)

  5. Evaluation of CRUDTRAN code to predict transport of corrosion products and radioactivity in the PWR primary coolant system

    International Nuclear Information System (INIS)

    Lee, C.B.

    2002-01-01

    CRUDTRAN code is to predict transport of the corrosion products and their radio-activated nuclides such as cobalt-58 and cobalt-60 in the PWR primary coolant system. In CRUDTRAN code the PWR primary circuit is divided into three principal sections such as the core, the coolant and the steam generator. The main driving force for corrosion product transport in the PWR primary coolant comes from coolant temperature change throughout the system and a subsequent change in corrosion product solubility. As the coolant temperature changes around the PWR primary circuit, saturation status of the corrosion products in the coolant also changes such that under-saturation in steam generator and super-saturation in the core. CRUDTRAN code was evaluated by comparison with the results of the in-reactor loop tests simulating the PWR primary coolant system and PWR plant data. It showed that CRUDTRAN could predict variations of cobalt-58 and cobalt-60 radioactivity with time, plant cycle and coolant chemistry in the PWR plant. (author)

  6. Techniques for Primary-to-Secondary Leak Monitoring in PWR Plants

    International Nuclear Information System (INIS)

    Sohn, Wook; Chi, Jun Hwa; Kang, Duck Won; Tae, Jeong Woo

    2006-01-01

    Historically, corrosion and mechanical damage have made steam generator tubes in PWR plants see various types of degradation from both the primary and secondary sides of the tubes. Since the tube degradation can lead to through-wall failure, the plant personnel should make efforts to prevent the failure. One of such preventive efforts is to monitor primary-to-secondary leakage (PSL) that usually precedes the tube rupture. Thus the objective of PSL monitoring is to make operators to determine when to shutdown the plant in order to minimize the likelihood of propagation of leaks to tube rupture under normal and faulted conditions This paper addresses briefly the status of techniques for PSL monitoring used in PWR plants

  7. Layout of the primary circuit with its components for PWR and BWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1981-01-01

    The light water-moderated and cooled pressurized water reactors and boiling water reactors constitute the basis of economic utilization of nuclear energy all over the world. Pressurized water reactors up to capacities of 3,800 MWth are those most used for power generation. However, their potential capacities exceed 3,800 MWth, so that already in the near future PWR are conseivable which readily generate 1,500 to 2,000 MWe. The main problem for starting the next generation of PWRs are of safety measure and licensing questions. Interesting applications of the PWRs are nuclear district heating, generation of process steam of desalination plants, steam injection into the ground for oil production or chemical factories. A new generation of natural circulation boiling water reactors with a capacity of 200 to 400 MW will be used for development of small industrial areas or for countries without an integral grid system. The natural circulation boiling water reactor will be subject of a separate lecture. Due to the fact of the majority of the PWR all over the world this lecture will discuss mainly PWR design aspects. (orig./RW)

  8. Recent bibliography on analytical and sampling problems of a PWR primary coolant Pt. 1

    International Nuclear Information System (INIS)

    Illy, H.

    1981-12-01

    The first bibliography on analytical and sampling problems of a PWR primary coolant (KFKI Report-1980-48) was published in 1980 and it covered the literature published in the previous 8-10 years. The present supplement reviews the subsequent literature up till December 1981. It also includes some references overlooked in the first volume. The serial numbers are continued from the first bibliography. (author)

  9. Hard alloys testing-machine for values of PWR primary coolant circuits

    International Nuclear Information System (INIS)

    Campan, J.L.; Sauze, A.

    1980-01-01

    Testing of valve parts or material used in valve fabrication and particularly seizing conditions in friction of plane surfaces coated with hard alloys of the type stellite. The testing equipment called Marguerite is composed of a hot pressurized water loop in conditions similar to PWR primary coolant circuits (320 0 C, 150 bars) and a testing-machine with measuring instruments. Testing conditions and samples are described [fr

  10. Analytical and sampling problems in primary coolant circuits of PWR-type reactors

    International Nuclear Information System (INIS)

    Illy, H.

    1980-10-01

    Details of recent analytical methods on the analysis and sampling of a PWR primary coolant are given in the order as follows: sampling and preparation; analysis of the gases dissolved in the water; monitoring of radiating substances; checking of boric acid concentration which controls the reactivity. The bibliography of this work and directions for its use are published in a separate report: KFKI-80-48 (1980). (author)

  11. Primary water chemistry improvement for radiation exposure reduction at Japanese PWR Plants

    Energy Technology Data Exchange (ETDEWEB)

    Nishizawa, Eiichi [Omiya Technical Institute, Saitama-ken (Japan)

    1995-03-01

    Radiation exposure during the refueling outages at Japanese Pressurized Water Reactor (PWR) Plants has been gradually decreased through continuous efforts keeping the radiation dose rates at relatively low level. The improvement of primary water chemistry in respect to reduction of the radiation sources appears as one of the most important contributions to the achieved results and can be classified by the plant operation conditions as follows

  12. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  13. Recent bibliography on analytical and sampling problems of a PWR primary coolant Suppl. 4

    International Nuclear Information System (INIS)

    Illy, H.

    1986-09-01

    The 4th supplement of a bibliographical series comprising the analytical and sampling problems of the primary coolant of PWR type reactors covers the literature from 1985 up to July 1986 (220 items). References are listed according to the following topics: boric acid; chloride, chlorine; general; hydrogen isotopes; iodine; iodide; noble gases; oxygen; other elements; radiation monitoring; reactor safety; sampling; water chemistry. (V.N.)

  14. Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water

    International Nuclear Information System (INIS)

    Deleume, J.

    2007-11-01

    Improving the performance and reliability of the fuel assemblies of the pressurized water reactors requires having a perfect knowledge of the operating margins of both the components and the materials. The choice of alloy 718 as reference material for this study is justified by the industrial will to identify the first order parameters controlling the excellent resistance of this alloy to Stress Corrosion Cracking (SCC). For this purpose, a specific slow strain rate (SSR) crack initiation test using tensile specimen with a V-shaped hump in the middle of the gauge length was developed and modeled. The selectivity of such SSR tests in simulated PWR primary water at 350 C was clearly established by characterizing the SCC resistance of nine alloy 718 thin strip heats. Regardless of their origin and in spite of a similar thermo-mechanical history, they did not exhibit the same susceptibility to SCC crack initiation. All the characterized alloy 718 heats develop oxide scale of similar nature for various exposure times to PWR primary medium in the temperature range [320 C - 360 C]. δ phase precipitation has no impact on alloy 718 SCC initiation behavior when exposed to PWR primary water, contrary to interstitial contents and the triggering of plastic instabilities (PLC phenomenon). (author)

  15. The empirical intensity of PWR primary coolant pumps failure and repair

    International Nuclear Information System (INIS)

    Milivojevicj, S.; Riznicj, J.

    1988-01-01

    The wealth of operating experience concerning PWR type and nuclear reactors that has been regularly monitored and systematically processes since 1971, enabled an analysis of the PWR primary coolant pumps operation. Failure intensity α and repair intensity μ of the pump during its working life were calculated, as these values are necessary in order to determine the reliability and availability of the pump as the basis for analyzing its effect on the safety and efficiency of the nuclear power plant. The trend of failure intensity α follows the theoretically expected changes in α over time, and this is around 10 -5 in the majority of life-time. Repair intensity μ indicates a slow rise during life-time, i.e. its faster return to operation. (author).7 refs.; 5 figs

  16. DETERMINATION OF THE 129I IN PRIMARY COOLANT OF PWR

    Directory of Open Access Journals (Sweden)

    KE CHON CHOI

    2013-02-01

    In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of 129I was examined, as was the effect of 3H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous 3H presence was found with activity concentrations of 3H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 μg/mL.

  17. Behaviour of fission products in PWR primary coolant and defected fuel rods evaluation

    International Nuclear Information System (INIS)

    Bourgeois, P.; Stora, J.P.

    1979-01-01

    The activity surveillance of the PWR primary coolant by γ spectometry gives some informations on fuel failures. The activity of different nuclides e.g. Xenons, Kryptons, Iodines, can be correlated with the number of the defected fuel rods. Therefore the precharacterization with eventually a prelocalization of the related fuel assemblies direct the sipping-test and allows a saving of time during refueling. A model is proposed to calculate the number of the defected rods from the activity measurements of the primary coolant. A semi-empirical model of the release of the fission products has been built from the activity measurements of the primary coolant in a 900 MWe PWR. This model allows to calculate the number of the defected rods and also a typical parameter of the mean damage. Fission product release is described by three stages: release from uranium dioxide, transport across the gas gap and behaviour in the primary coolant. The model of release from the oxide considers a diffusion process in the grains with trapping. The release then occurs either directly to free surfaces or with a delay due to a transit into closed porosity of the oxide. The amount released is the same for iodine and rare gas. With the gas gap transit is associated a transport time and a probability of trapping for the iodines. In the primary coolant the purification and the radioactive decay are considered. (orig.)

  18. Application of CATE 2.0 code on evaluating activated corrosion products in a PWR cooling loop

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Jingyu; Li, Lu; Chen, Yixue [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering

    2017-03-15

    In PWR plants, most Occupational Radiation Exposure (ORE) for personnel results from Activated Corrosion Products (ACPs) in the cooling loop. In order to evaluate the ACPs in the cooling loop, a three-region transport model is built up based on the theory of driving force from the concentration difference in CATE 2.0 code. In order to analyze the nuclide composition of ACPs, the EAF-2007 nuclear database is embedded in CATE 2.0. The case of MIT PCCL test loop is simulated to test the availability of CATE 2.0 on PWR ACPs evaluation, and the activity of Co-58 and Co-60 after operation for 42 days calculated by CATE 2.0 is consistent with that from the code CRUDSIM adopted by MIT. Then, the nuclide composition of ACPs is analyzed in detail respectively for operation of 42 days and 12 months using CATE 2.0. The results show that the short-lived nuclides contribute a majority of the activity in the regions of in-flux wall and coolant, while the long-lived nuclides contribute most of the activity in the region of out-flux wall.

  19. Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2014-01-01

    Highlights: • A PWR spent fuel pool was modeled by using MAAP5. • Loss of pool cooling severe accident scenarios were studied. • Loss of pool cooling accidents with two mitigation measures were analyzed. - Abstract: The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents the analysis of loss of pool cooling accident scenarios and the discussion of mitigation measures for the SFP at a pressurized water reactor (PWR) NPP with the MAAP5 code. Analysis of uncompensated loss of water due to the loss of pool cooling with different initial pool water levels of 12.2 m (designated as a reference case) and 10.7 m have been performed based on a MAAP5 input model. Scenarios of the accident such as overheating of uncovered fuel assemblies, oxidation of claddings and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission products were predicted with the assumption that mitigation measures were unavailable. The results covered a broad spectrum of severe accident evaluations in the SFP. Furthermore, as important mitigation measures, the effects of recovering the SFP cooling system and makeup water in SFP on the accident progressions have also been investigated respectively based on the events of pool water boiling and spent fuels uncovery. Based upon the reference case, three cases with the recovery of SFP cooling system and three other cases with makeup water in SFP have been studied. The results showed that, severe accident might happen if SFP cooling system was not restored timely before the spent fuels started to become uncovered; spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate larger than the average evaporation rate defined as the division of pool water mass above the

  20. Response of unirradiated and irradiated PWR fuel rods tested under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Quapp, W.J.; Martinson, Z.R.; McCardell, R.K.; Mehner, A.S.

    1978-01-01

    This report summarizes the results from the single-rod power-cooling-mismatch (PCM) and irradiation effects (IE) tests conducted to date in the Power Burst Facility (PBF) at the U.S. DOE Idaho National Engineering Laboratory. This work was performed for the U.S. NRC under contact to the Department of Energy. These tests are part of the NRC Fuel Behavior Program, which is designed to provide data for the development and verification of analytical fuel behavior models that are used to predict fuel response to abnormal or postulated accident conditions in commercial LWRs. The mechanical, chemical and thermal response of both previously unirradiated and previously irradiated LWR-type fuel rods tested under power-cooling-mismatch condition is discussed. A brief description of the test designs is presented. The results of the PCM thermal-hydraulic studies are summarized. Primary emphasis is placed on the behavior of the fuel and cladding during and after stable film boiling. (orig.) [de

  1. Stress corrosion cracking of nickel base alloys in PWR primary water

    International Nuclear Information System (INIS)

    Guerre, C.; Chaumun, E.; Crepin, J.; De Curieres, I.; Duhamel, C.; Heripre, E.; Herms, E.; Laghoutaris, P.; Molins, R.; Sennour, M.; Vaillant, F.

    2013-01-01

    Stress corrosion cracking (SCC) of nickel base alloys and associated weld metals in primary water is one of the major concerns for pressurized water reactors (PWR). Since the 90's, highly cold-worked stainless steels (non-sensitized) were also found to be susceptible to SCC in PWR primary water ([1], [2], [3]). In the context of the life extension of pressurized water reactors, laboratory studies are performed in order to evaluate the SCC behaviour of components made of nickel base alloys and of stainless steels. Some examples of these laboratory studies performed at CEA will be given in the talk. This presentation deals with both initiation and propagation of stress corrosion cracks. The aims of these studies is, on one hand, to obtain more data regarding initiation time or crack growth rate and, one the other hand, to improve our knowledge of the SCC mechanisms. The aim of these approaches is to model SCC and to predict components life duration. Crack growth rate (CGR) tests on Alloy 82 with and without post weld heat treatment are performed in PWR primary water (Figure 1). The heat treatment seems to be highly beneficial by decreasing the CGR. This result could be explained by the effect of thermal treatment on the grain boundary nano-scopic precipitation in Alloy 82 [4]. The susceptibility to SCC of cold worked austenitic stainless steels is also studied. It is shown that for a given cold-working procedure, SCC susceptibility increases with increasing cold-work ([2], [5]). Despite the fact that the SCC behaviour of Alloy 600 has been widely studied for many years, recent laboratory experiments and analysis ([6], [7], [8]) showed that oxygen diffusion is not a rate-limiting step in the SCC mechanism and that chromium diffusion in the bulk close the crack tip could be a key parameter. (authors)

  2. Recent bibliography on analytical and sampling problems of a PWR primary coolant Suppl. 3

    International Nuclear Information System (INIS)

    Illy, H.

    1985-03-01

    The present supplement to the bibliography on analytical and sampling problems of PWR primary coolant covers the literature published in 1984 and includes some references overlooked in the previous volumes dealing with the publications of the last 10 years. References are devided into topics characterized by the following headlines: boric acid; chloride; chlorine; carbon dioxide; general; gas analysis; hydrogen isotopes; iodine; iodide; nitrogen; noble gases and radium; ammonia; ammonium; oxygen; other elements; radiation monitoring; reactor safety; sampling; water chemistry. Under a given subject bibliographical information is listed in alphabetical order of the authors. (V.N.)

  3. Study of the formation and transport of corrosion products in PWR primary circuit simulators

    International Nuclear Information System (INIS)

    Noe, M.; Frejaville, G.; Camp, J.J.

    1983-01-01

    The formation, migration and deposition of corrosion products in PWR primary circuits are studied in out-of-reactor loops. The aim of these studies is to limit the build-up of the radiation fields impinging on out-of-flux walls and to reduce the danger of rapid corrosion of fuel cans, taking into account the tougher conditions imposed on current trends in the operation of such industrial plants. Four simulator loops and their respective possibilities and research methods are described. (author)

  4. Zinc Addition Effects on General Corrosion of Austenitic Stainless Steels in PWR Primary Conditions

    International Nuclear Information System (INIS)

    Qiao Peipeng; Zhang Lefu; Liu Ruiqin; Jiang Suqing; Zhu Fawen

    2010-01-01

    Zinc addition effects on general corrosion of austenitic stainless steel 316 and 304 were investigated in simulated PWR primary coolant without zinc or with 50 ppb zinc addition at 315 degree C for 500 h. The results show that with the addition of zinc, the corrosion rate of austenitic stainless steel is effectively reduced, the surface oxide film is thinner, the morphology and chemical composition of surface oxide scales are evidently different from those without zinc. There are needle-like corrosion products on the surface of stainless steel 304. (authors)

  5. The impact of radiolytic yield on the calculated ECP in PWR primary coolant circuits

    International Nuclear Information System (INIS)

    Urquidi-Macdonald, Mirna; Pitt, Jonathan; Macdonald, Digby D.

    2007-01-01

    A code, PWR-ECP, comprising chemistry, radiolysis, and mixed potential models has been developed to calculate radiolytic species concentrations and the corrosion potential of structural components at closely spaced points around the primary coolant circuits of pressurized water reactors (PWRs). The pH(T) of the coolant is calculated at each point of the primary-loop using a chemistry model for the B(OH) 3 + LiOH system. Although the chemistry/radiolysis/mixed potential code has the ability to calculate the transient reactor response, only the reactor steady state condition (normal operation) is discussed in this paper. The radiolysis model is a modified version of the code previously developed by Macdonald and coworkers to model the radiochemistry and corrosion properties of boiling water reactor primary coolant circuits. In the present work, the PWR-ECP code is used to explore the sensitivity of the calculated electrochemical corrosion potential (ECP) to the set of radiolytic yield data adopted; in this case, one set had been developed from ambient temperature experiments and another set reported elevated temperatures data. The calculations show that the calculated ECP is sensitive to the adopted values for the radiolytic yields

  6. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  7. Thermodynamics and the transport of corrosion products in PWR primary circuits

    International Nuclear Information System (INIS)

    Turner, D.J.

    1992-01-01

    It is argued that practically useful models for the activation, transport and deposition of corrosion products in PWR primary circuits can only be produced on the basis of an improved understanding of the chemical processes which control them. In particular, if a model is to make reliable predictions it is essential that its thermodynamic basis be sound. This is not the case with most current models which employ the erroneous concept of a corrosion product 'solubility'. In addition to the misuse of this term, other complications are discussed. These include the need to take account of the consequences of Gibbs' phase rule and the fact that, for mixed spinels, neither the concept of a thermodynamic solubility nor of a solubility product is valid. There is no reason to believe that measured apparent solubilities of nickel ferrites or spinel mixtures containing cobalt can give any direct guidance on the direction of transport of Ni or Co in PWR primary circuits. This is more likely to be determined by the distribution of stable and unstable ferrites and chromites than by any temperature coefficient of apparent solubility. Most of the transport of Ni and Co into and out of the core probably occurs as a consequence of either chemical or mechanical transients. Most important is likely to be the oxidative destruction and subsequent re-precipitation of chromites which occurs as a consequence of the oxygenated conditions employed during plant shutdown. (author)

  8. Stress corrosion cracking of Alloy 600 in primary water of PWR: study of chromium diffusion

    International Nuclear Information System (INIS)

    Chetroiu, Bogdan-Adrian

    2015-01-01

    Alloy 600 (Ni-15%Cr-10%Fe) is known to be susceptible to Stress Corrosion Cracking (SCC) in primary water of Pressurized Water Reactors (PWR). Recent studies have shown that chromium diffusion is a controlling rate step in the comprehension of SCC mechanism. In order to improve the understanding and the modelling of SCC of Alloy 600 in PWR primary medium the aim of this study was to collect data on kinetics diffusion of chromium. Volume and grain boundary diffusion of chromium in pure nickel and Alloy 600 (mono and poly-crystals) has been measured in the temperature range 678 K to 1060 K by using Secondary Ions Mass Spectroscopy (SIMS) and Glow Discharge-Optical Spectrometry (GD-OES) techniques. A particular emphasis has been dedicated to the influence of plastic deformation on chromium diffusion in nickel single crystals (orientated <101>) for different metallurgical states. The experimental tests were carried out in order to compare the chromium diffusion coefficients in free lattice (not deformed), in pre-hardening specimens (4% and 20%) and in dynamic deformed tensile specimens at 773 K. It has been found that chromium diffusivity measured in dynamic plastic deformed creep specimens were six orders of magnitude greater than those obtained in not deformed or pre-hardening specimens. The enhancement of chromium diffusivity can be attributed to the presence of moving dislocations generated during plastic deformation. (author)

  9. Verification test for radiation reduction effect and material integrity on PWR primary system by zinc injection

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, H.; Nagata, T.; Yamada, M. [Nuclear Power Engineering Corp. (Japan); Kasahara, K.; Tsuruta, T.; Nishimura, T. [Mitsubishi Heavy Industries, Ltd. (Japan); Ishigure, K. [Saitama Inst. of Tech. (Japan)

    2002-07-01

    Zinc injection is known to be an effective method for the reduction of radiation source in the primary water system of a PWR. There is a need to verify the effect of Zn injection operation on radiation source reduction and materials integrity of PWR primary circuit. In order to confirm the effectiveness of Zn injection, verification test as a national program sponsored by Ministry of Economy, Trade and Industry (METI) was started in 1995 for 7-year program, and will be finished by the end of March in 2002. This program consists of irradiation test and material integrity test. Irradiation test as an In-Pile-Test managed by AEAT Plc(UK) was performed using the LVR-15 reactor of NRI Rez in Check Republic. Furthermore, Out-of-Pile-Test using film adding unit was also performed to obtain supplemental data for In-Pile-Test at Takasago Engineering Laboratory of NUPEC. Material Integrity test was planned to perform constant load test, constant strain test and corrosion test at the same time using large scale Loop and slow strain extension rate testing (SSRT) at Takasago Engineering Laboratory of NUPEC. In this paper, the results of the verification test for Zinc program at present are discussed. (authors)

  10. Application of surface science to the study of the corrosion of PWR primary circuit materials

    International Nuclear Information System (INIS)

    Harris, S.J.

    1989-04-01

    This thesis describes a study of the corrosion and oxidation of PWR primary circuit materials using surface sensitive spectroscopic techniques. An X-ray photoemission spectroscopy (XPS) study of a number of mixed oxides of known composition is described and the information obtained is related to XPS measurements made on the surface of iron and nickel based alloys oxidised under controlled conditions. A secondary ion mass spectroscopy (SIMA) study on these mixed transition metal oxides is also described. The gaseous oxidation of stainless steel 3041 and Inconel-690 is examined. Both alloys were oxidised at 600K in air with the composition of the oxide films formed studied by a range of surface spectroscopic methods. Further experimental work was performed on Inconel-690 to examine the effects of surface pretreatment and the effects of low oxygen partial pressures on the formation of oxide films at 600 K. The incorporation of the radionuclide, cobalt-60, into the oxide films formed on structural components of a PWR, result in the build up of radiation fields. A method of pretreating the surface of the alloy stainless steel 3041, in order to reduce the level of cobalt adsorbed into the oxide film formed under simulated primary coolant conditions is examined and contrasts with treatments which have been developed to release cobalt adsorbed in existing oxide layers under reactor conditions are discussed. (author)

  11. Laboratory results gained from cold worked type 316Ti under simulated PWR primary environment

    International Nuclear Information System (INIS)

    Devrient, B.; Kilian, R.; Koenig, G.; Widera, M.; Wermelinger, T.

    2015-01-01

    Beginning in 2005, intergranular stress corrosion cracking (IGSCC) of barrel bolts made from cold worked type 316Ti (German Material No. 1.4571 K) was observed in several S/KWU type PWRs. This mechanism was so far less understood for PWR primary conditions. Therefore an extended joint research program was launched by AREVA GmbH and VGB e.V. to clarify the specific conditions which contributed to the observed findings on barrel bolts. In the frame of this research program beneath the evaluation of the operational experience also laboratory tests on the general cracking behavior of cold worked type 316Ti material, which followed the same production line as for barrel bolt manufacturing in the eighties, with different cold work levels covering up to 30 % were performed to determine whether there is a specific susceptibility of cold worked austenitic stainless steel specimens to suffer IGSCC under simulated PWR primary conditions. All these slow strain rate tests on tapered specimens and component specimens came to the results that first, much higher cold work levels than used for the existing barrel bolts are needed for IGSCC initiation. Secondly, additional high active plastic deformation is needed to generate and propagate intergranular cracking. And thirdly, all specimens finally showed ductile fracture at the applied strain rates. (authors)

  12. 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

  13. The deformation of zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-01-01

    The deformation behaviour is reported of specimens of Zircaloy PWR fuel cladding when directly heated in flowing steam. The range of internal pressures studied was 0.69-2.07 MPa; this extended earlier studies using higher pressures. The specimens were ramped and then held at a steady test temperature until rupture or until 600 seconds had elapsed. Under these conditions it was found that extended deformation occurred with pressures down to 1 MPa at temperatures up to 900 deg C. At lower pressures and higher temperatures there was no large extended deformation; this is believed to result from the effects of oxidation

  14. Application of liquid chromatography techniques to the measurement of soluble transition metals in PWR primary coolant

    International Nuclear Information System (INIS)

    Amey, M.D.H.; Brown, G.R.

    1987-01-01

    Two chromatographic techniques have been developed, and evaluated for the on-line analysis of soluble transition metals, particularly cobalt, in PWR primary coolant. Automatic operation and control, together with data processing and storage has been achieved by interfacing a Dionex ion chromatograph to a microprocessor control system. An absolute detection limit of 0.1 ng cobalt has been obtained which, with on-line sample preconcentration (100 ml), has enabled measurements to be made down to part-per-trillion levels (0.001 ppb). Application of the techniques to PWR coolant analysis was demonstrated by a programme of work on the Half Megawatt Loop at Winfrith. During this work some aspects of the behaviour of soluble metal species have been studied in both de-oxygenated and hydrogenated conditions. The effects of changes in coolant chemistry, operating temperature, and sample line flowrates on circulating impurity levels are reported, together with the dramatic effects observed when part of the circuit pipework was replaced with new stainless steel tubing. (author)

  15. The impact of steam generator replacement on PWR primary system contamination

    International Nuclear Information System (INIS)

    Dacquait, F.; Marteau, H.; Guinard, L.; Ranchoux, G.; Taunier, S.; Wintergerst, M.; Bretelle, J.L.; Rocher, A.

    2010-01-01

    This paper analyses the impact of Steam Generator Replacement (SGR) on PWR primary circuit contamination. It presents a comparison of the activities deposited inside the primary system and released during refuelling outages after SGR with three different SG tube alloys (600, 690 and 800) and different SG tube manufacturing processes. A SGR has a great impact on the primary system contamination. After SGR, whatever the SG tube material is, the typical variations are the following: The 58 Co contamination increases for 1 to 3 cycles, and then decreases to very low levels in some cases, mainly depending on the manufacturing process of the replacement SG tubes; The 60 Co Co contamination tends to decrease on the primary coolant pipes and increases by a lower rate on the new SG tubes. This analysis highlights the importance on contamination levels after SGR of both the corrosion product deposits on the primary surfaces before SGR and the surface finish of the SG tubes related to their manufacturing process. (author)

  16. Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

    Science.gov (United States)

    Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

  17. On the failure probability of the primary piping of the PWR

    International Nuclear Information System (INIS)

    Schueller, G.I.; Hampl, N.C.

    1984-01-01

    A methodology for quantification of the structural reliability of the primary piping (PP) of a PWR under operational and accidental conditions is developed. Biblis B is utilized as reference plant. The PP structure is modeled utilizing finite element procedures. Based on the properties of the operational and internal accidental conditions, a static analysis suffices. However, a dynamic analysis considering non-linear effects of the soil-structure-interaction is to be used to determine load effects due to earthquake induced loading. Considering realistically the presence of initial cracks in welds and considering annual frequencies of occurrence of the various loading conditions, a crack propagation calculation utilizing the Forman model is carried out. Simultaneously leak and break probabilities using the 'Two Criteria'-Aproach are computed. A Monte Carlo simulation procedure is used as method of solution. (Author) [pt

  18. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1992-03-01

    The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

  19. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1993-01-01

    The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy-800, Inconel-600, Inconel-690 and 321 SS which are used for steam generator in PWR. The effects of material metallurgy, shot-peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC are investigated

  20. Influence of deformation on SCC susceptibility of austenitic stainless steel in PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Kaneshima, Yoshiari; Totsuka, Nobuo; Nakajima, Nobuo [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    Slow strain rate tests (SSRT) were carried out to evaluate the SCC susceptibility of four types of austenitic stainless steels (SUS304, SUS316, SUS304L and SUS316L) in PWR primary water. The influence of deformation on SCC susceptibility of SUS316 was studied. All types of stainless steel were susceptible to SCC, and the SCC susceptibility varied depending on the steel type. The comparison of the SSRT results and tensile test in air based on the reduction of area measurement showed that the SCC susceptibility increased with increasing the degree of deformation. For explaining the influence of deformation on SCC susceptibility, it is necessary to evaluate both intergranular and transgranular fractures. (author)

  1. Housing maintenance of primary cooling pump I of Kartini reactor

    International Nuclear Information System (INIS)

    Agung Nugroho; Wahyu Imam W

    2013-01-01

    Housing maintenance of Primary Cooling Pump have been done with purpose to enhance capability of fluid block and stopping leakage. The procedures of modification are follow: Replace mechanical seal type Nock 560 – 1 ½'' by mechanical seal type Nock 560 - 38, modification gland plate/housing with to lathe the diameter of gland plate/housing from diameter of 54 mm to 60 mm, in size setting-up mechanical seal, alignment and then function test. The result of the modification are: mechanical seal has been installed, housing has been modified, and the leakage of primary cooling water has been repaired and operated properly. Conclusion of the maintenance are the primary cooling water pump is working well, because the primary cooling water is not leakage any more. (author)

  2. Major activated corrosion products cobalt, silver and antimony in the primary coolant of PWR power plants

    International Nuclear Information System (INIS)

    Xu Mingxia

    2012-01-01

    The production of the major activated corrosion products such as cobalt, silver and antimony in the primary coolant of PWR power plants and the impacts on the increase of the dose rates caused by these corrosion products during the shutdown are described in the paper. Investigating the corrosion product behavior during the operation and shutdown periods aims at detecting the appearance of these radiological pollutants in the early time and searching relevant solutions that may enable eventually to decrease the dose rate. The solutions may include: Replacing critical material in the primary system's equipment and components, which contact with primary coolant circuit to possibly limit the source term, Elaborating strictly the specific chemical and shutdown procedure to optimize the purification capacity and to minimize the over-contaminations; Improving purification techniques according to the real operation circumstance, and limiting the impacts of these pollutants. It is obvious in the real practices that implementing appropriate solution will be benefit to decrease or limit the pollutants species like cobalt, silver and antimony. (author)

  3. Processing of coke oven gas. Primary gas cooling

    Energy Technology Data Exchange (ETDEWEB)

    Ullrich, H [Otto (C.) und Co. G.m.b.H., Bochum (Germany, F.R.)

    1976-11-01

    The primary cooler is an indispensable part of all byproduct processing plants. Its purpose is to cool the raw gas from the coke oven battery and to remove the accompanying water vapor. The greater part of the cooling capacity is utilized for the condensation of water vapor and only a small capacity is needed for the gas cooling. Impurities in the gas, like naphthalene, tar and solid particles, necessitate a special design in view of the inclination to dirt accumulation. Standard types of direct and indirect primary gas coolers are described, with a discussion of their advantages and disadvantages.

  4. Zircaloy PWR fuel cladding deformation tests under mainly convective cooling conditions

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.

    1980-01-01

    In a loss-of-coolant accident the temperature of the cladding of the fuel rods may rise to levels (650-810 0 C) where the ductility of Zircaloy is high (approximately 80%). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of PWR Zircaloy cladding tubing 450 mm long under internal pressure had shown that strains of over 50% could be produced over considerable lengths (greater than twenty tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel element. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss believed to be more representative of a fuel element following a loss of coolant. Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 845 0 C in flowing steam at atmospheric pressure. Internal test pressures were in the range 2.9-11.0 MPa (400-1600 1b/in 2 ). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the steam flow. These results are compared with those from multi-rod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behaviour of fuel elements in a loss-of-coolant accident are outlined. (author)

  5. Low cycle fatigue of Alloy 690 and welds in a simulated PWR primary water environment

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jongdae; Cho, Pyungyeon; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Cho, Pyungyeon [Khalifa Univ., Abu Dhabi (United Arab Emirates); Kim, Tae Soon; Lee, Yong Sung [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2013-05-15

    In this study, environmental fatigue tests for these materials were performed and the new prediction model of fatigue life of Alloy 690 and weld in primary water condition was proposed. To evaluate the fatigue life of Alloy 690 and 52M in a PWR environment, low cycle fatigue tests were performed and revised fatigue life prediction models and environmental factor were proposed. With the revised Fen model for Alloy 690 and 52M, the reliability of the fatigue life prediction has been improved. The reduction of low cycle fatigue life of metallic materials in the primary coolant water environments has been the subject of debate between the utility and regulator since 1980s. It became the significant licensing problem since the issue of RG-1.207 by U. S. NRC. The statistical model for the environmental factor, Fen, specified in RG-1.207 was based on the extensive test results accumulated by the ANL and Japanese national program. Of the materials, the limited fatigue life data of Ni-Cr-Fe alloys were used to develop the Fen for the alloys. Furthermore, test data for Alloy 690 and its weld are limited. Considering that Alloy 690 will be extensively used in the new nuclear power plants, additional effort to validate or improve current Fen model is required.

  6. VGB primary and secondary side water chemistry guidelines for PWR plants

    International Nuclear Information System (INIS)

    Neder, H.; Wolter, D.; Staudt, U.

    2007-01-01

    The recent revision of the VGB Water Chemistry Guidelines was issued in 2005 and published in the second half of 2006. These guidelines are based on the primary and secondary side operating chemistry experience with all Siemens designed pressurized water reactors gained since the beginning of the 1980s. These guidelines cover For the primary side chemistry Modified lithium boron chemistry, Zinc chemistry for dose rate reduction, Enriched boric acid (EBA) chemistry for high duty core design For the secondary side chemistry High all-volatile treatment (AVT) chemistry (high pH operation) Oxygen injection in the secondary side Especially for the secondary side chemistry, compared with the water chemistry guidelines of other organizations worldwide, these Guidelines are less stringent, providing more operational flexibility to the plant operation, and can be applied for all new designs of steam generators with egg-crates or broached hole tube supports and with I 690TT or I 800 tubing materials. This paper gives an overview of the 2006 revision of the VGB Water Chemistry Guidelines for PWR plants and describes the fundamental goals of water chemistry operation strategies. In addition, the reasons for the selected control parameters and action levels, to achieve an adequate plant performance, are presented based on the operating experience. (orig.)

  7. Modelling of the local chemistry in stagnant areas in the PWR primary circuit

    International Nuclear Information System (INIS)

    Reid, Rick; Fruzzetti, Keith; Ahluwalia, Al; Summe, Alex; Dame, Cecile; Schmitt, Kyle

    2014-01-01

    MRP-236 demonstrated a correlation between stagnant or low flow conditions and stress corrosion cracking (SCC) of stainless steel components in the PWR primary system. Of the approximately 140 SCC events documented (affecting 15 different components), 83% involved stagnant or low flow conditions that were likely to be associated with chemical environments different from the well mixed bulk coolant. The chemistry in such locations is typically not monitored, and sampling is difficult or impossible. Actions to improve chemistry in regions of low or no coolant flow, such as flushing, cycling of components and imposition of more stringent make up water chemistry controls affect both operational costs and outage schedules. Similarly, design changes to improve flow in affected areas are costly or impracticable. Improving the understanding of the factors controlling chemistry in such areas and development of the capability to predict typical and worst case conditions will allow an informed assessment of procedural actions and/or design changes to improve local chemistry and thereby reduce SCC susceptibility. A project was undertaken to develop a model to predict local chemistry conditions in stagnant locations. The model comprises the iterative application of the EPRI MULTEQ solution chemistry equilibrium code and standard thermodynamic relationships to predict local chemistry conditions considered likely to have been present at the surfaces of components when SCC was initiated. The starting chemistry conditions are based on PWR primary system chemistry from different plant maneuvers (e.g., startup and shutdown conditions). The model was applied to three example components where SCC has occurred in the field. The selected components were: control rod drive mechanism canopy seals; valve drain lines; and reactor vessel o-ring leak-off lines. This paper provides a summary of the model and predicted local chemistry conditions that develop for the three example component as a

  8. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors

    International Nuclear Information System (INIS)

    Robbe, M.F.

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  9. Recent bibliography on analytical and sampling problems of a PWR primary coolant

    International Nuclear Information System (INIS)

    Illy, H.

    1980-07-01

    An extensive bibliography on the problems of analysis and sampling of the primary cooling water of PWRs is presented. The aim was to collect the analytical methods for dissolved gases. The sampling and preparation are also taken into account. last 8-10 years is included. The bibliography is arranged into alphabetical order by topics. The most important topics are as follows: boric acid, gas analysis, hydrogen isotopes, iodine, noble gases, radiation monitoring, sampling and preparation, water chemistry. (R.J.)

  10. Possibility of Localized Corrosion of PWR primary side materials in oxidative decontamination condition

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Yoon; Jung, Jun Young; Won, Huijun; Kim, Seon Byeong; Choi, Wangkyu; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Primary circuit of a PWR (radionuclides uptake in the inner oxide layer and oxide/metal interface occurred inevitably. Therefore, it is necessary to remove the inner oxide layer as well as the outer oxide layer to achieve excellent decontamination effects. It is known that the outer oxide layers are typically composed of Fe{sub 3}O{sub 4} and NiFe{sub 2}O{sub 4}. On the other hand, the inner oxide layers are composed of Cr{sub 2}O{sub 3}, (Ni{sub 1-x}Ni{sub x})(Cr{sub 1-y}Fe{sub y}){sub 2}O{sub 4}, and FeCr{sub 2}O{sub 4}. Because of chromium in the trivalent oxidation state which is difficult to dissolve, the oxide layer has an excellent protectiveness and is hard to decontaminate. For the dissolution of chromium-rich oxide, there have been developed an alkaline permanganate (AP) or nitric permanganate (NP). A disadvantage of the AP process is the generation of a large volume of secondary waste. On the other hand, NP process is highly incompatible to the corrosion of the structure materials. In this study as a part of developing decontamination process, we investigated the corrosion behavior of the structure materials such as Inconel-600 and type 304 stainless steel in NP and AP oxidative decontamination conditions for the safe use of an oxidative phase in PWR system decontamination. The corrosion behavior was analyzed through the potential-pH equilibrium for the system of Cr-H{sub 2}O / Mn-H{sub 2}O at 90 .deg. C and potentiodynamic polarization in a typical AP and NP solution were evaluated. The AP or NP treated specimen surface was observed using an optical microscope and scanning electron microscopy (SEM) for an evaluation of the localized corrosion. The possibility of localized corrosion of PWR primary side materials under oxidative decontamination condition was evaluated using a potentiodynamic polarization technique, observation of localized corrosion morphology, and consideration of potential-pH diagrams at 90 .deg. C. From the results of these tests, we

  11. Natural circulation cooling in a PWR geometry under accident-induced conditions

    International Nuclear Information System (INIS)

    Shimeck, D.J.; Johnsen, G.W.

    1983-01-01

    The characteristics and limits of natural circulation heat rejection over a wide range of conditions were experimentally investigated in a small-scale model of a pressurized water reactor. Conditions that were varied included primary and secondary coolant inventory, decay heat power, and primary noncondensable gas content. The results have defined three distinct modes of natural circulation, their limits and transition points, and the characteristic signatures accompanying natural circulation behavior. Particular emphasis is focused on the limits of natural circulation under severely degraded primary and secondary conditions

  12. An experimental investigation of the effect of clad ballooning on the effectiveness of PWR emergency cooling

    International Nuclear Information System (INIS)

    Cooper, C.A.; Pearson, K.G.; Jowitt, D.

    1985-01-01

    A series of single phase cooling, forced reflood, gravity reflood and level swell experiments has been performed on a full length, electrically heated 7x7 rod fuel cluster (in the THETIS Rig) containing a blockage simulating very severe clad ballooning. The single phase cooling experiments provided data on the level of heat transfer within the cluster and the enhancement produced by turbulence created by the spacer grids. The forced reflood experiments have led to a better understanding of the rewetting processes during bottom reflooding, the very important influence of spacer grids in two-phase flow and the complicated heat transfer processes within the blockage. The insight obtained has been used to develop a mechanistic model. The gravity reflood experiments investigated steam binding and inlet flow oscillation effects. The variation in inlet flow produced by the variations in system parameters was the dominant influence. The inlet flow oscillations which occurred in gravity reflood appeared not to influence overall rewetting or heat removal performance. The level swell experiments investigated the relationship between void fraction and superficial steam velocity at pressures up to 40 bar and compared the data with various correlations. The correlation of Gardner was found to be most satisfactory. 107 refs., 252 figs.

  13. Measurement and Elemental Analysis of Muria Sea Water for Primary Water of PWR

    International Nuclear Information System (INIS)

    Dwi Biyantoro; Kris Tri Basuki

    2007-01-01

    Treatment process of water of free mineral and study area of processing of sea water to meet the need of water cooling of PWR. Desalination is a separation process used to reduce the dissolved salt content of saline water. All desalination processes involve three liquid streams: the saline feedwater (brackish water or seawater), low salinity product water, and very saline concentrate. Seawater Muria Jepara is feed type 1 with content salt is low and relative suspension and weight metals is low. There are two types of membrane process used for desalination: reverse osmosis (RO) and electrodialysis (ED). The starting of process is difficult and expensive. However, starting since 1950, desalination process appear to be economically practical for ordinary use. There are 3 key elements significantly affect to the technical performance and long term characteristic type, i.e. 1) energy, 2) corrosivity of seawater, and 3) desalination process technology. These elements closely inter related and important for design improvements and technical performance optimization. From the analysis of sea water of gulf Muria Jepara was found as follows : Fe = 0.176 ± 0.012 ppm, Pb = 0.612 ± 0.017 ppm, Cd = 52.567 ± 0.750 ppm, Cu = 0.044 ± 0.005 ppm, Zn = 0.061 ± 0.003 ppm, Mn = 0.057 ± 0.003 ppm, Ca between = 365.256 - 368.654 ppm, Na between = 9572.000 - 9775.000 ppm, Mg between 759.000 - 779.00 ppm and Ni = 0.524 ± 0.005 ppm. Cost production of process using reverse osmosis as around 0.9 - 1 US$/m3, while using electrodialysis is around 1.2 US$/m3, and by using evaporation process or distillation process is around 1.4 - 1.6 US$/m3. (author)

  14. The validation of corrosion inhibitors for ancillary cooling systems at Sizewell B PWR

    International Nuclear Information System (INIS)

    Allan, S.J.; Garnsey, R.; Lawson, W.F.

    1994-01-01

    Sizewell B has several low temperature ancillary cooling circuits for removing low grade heat. One of the principal circuits is the Component Cooling Water System (CCWS) primarily constructed from carbon steel pipework but also containing stainless steel, titanium, copper and cupronickel components. Linked to the CCWS is another carbon steel system namely the Reserve Ultimate Heat Sink (RUHS) a system unique to Sizewell B providing an independent heat removal capability for the CCWS as a safety back-up. Both systems utilize demineralized water with the addition of an inhibitor to prevent corrosion. The vendor recommended inhibitor namely potassium chromate is highly toxic to marine life and initial discussions with the regulatory authorities indicated that it would be extremely difficult to obtain a discharge consent to cover leakage or potential system purges and dumps associated with maintenance operations. Thus an alternative was sought. Sodium nitrite (anhydrous NaN0 2 )/borax (Na 2 B 4 0 7 1OH 2 0) inhibitor solutions have been used for many years in the power industry for preservation of carbon steel plant and several proprietary formulations are based on this mixture. There has been extensive test work in the UK on this inhibitor for secondary side boiler storage on gas reactors and therefore it was considered a prime candidate for use at Sizewell B. Starting in 1985 a programme of work sponsored by Nuclear Electric plc has been completed by NNC Ltd to compare nitrite/borax with other suitable inhibitors and validate its use for Sizewell B. (authors). 5 figs., 1 tab

  15. SCC growth behaviors of austenitic stainless steels in simulated PWR primary water

    Science.gov (United States)

    Terachi, T.; Yamada, T.; Miyamoto, T.; Arioka, K.

    2012-07-01

    The rates of SCC growth were measured under simulated PWR primary water conditions (500 ppm B + 2 ppm Li + 30 cm3/kg-H2O-STP DH2) using cold worked 316SS and 304SS. The direct current potential drop method was applied to measure the crack growth rates for 53 specimens. Dependence of the major engineering factors, such as yield strength, temperature and stress intensity was systematically examined. The rates of crack growth were proportional to the 2.9 power of yield strength, and directly proportional to the apparent yield strength. The estimated apparent activation energy was 84 kJ/mol. No significant differences in the SCC growth rates and behaviors were identified between 316SS and 304SS. Based on the measured results, an empirical equation for crack growth rate was proposed for engineering applications. Although there were deviations, 92.8% of the measured crack growth rates did not exceed twice the value calculated by the empirical equation.

  16. Precursor evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Ziqing; Kruska, Karen; Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2017-03-27

    Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for the 21% and 31%CW CLT specimens loaded at their yield stress after ~9,220 h, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400h exposure at constant stress intensity, which resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Interestingly, post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and will discuss their effects on crack initiation in CW alloy 690.

  17. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  18. Influence of Localized Plasticity on IASCC Sensitivity of Austenitic Stainless Steels under PWR Primary Water

    Science.gov (United States)

    Cissé, Sarata; Tanguy, Benoit; Laffont, Lydia; Lafont, Marie-Christine; Guerre, Catherine; Andrieu, Eric

    The sensibility of precipitation-strengthened A286 austenitic stainless steel to Stress Corrosion Cracking (SCC) is studied by means of Slow Strain Rate Tests (SSRT). First, alloy cold working by Low Cycle Fatigue (LCF) is investigated. Fatigue tests under plastic strain control are performed at different strain levels (Δ ɛp/2=0.2%, 0.5% and 0.8%) in order to establish correlation between stress softening and deformation microstructure resulting from LCF tests. Deformed microstructures have been identified through TEM investigations. Three states of cyclic behaviour for precipitation-strengthened A286 have been identified: hardening, cyclic softening and finally saturation of softening. It is shown that the A286 alloy cyclic softening is due to microstructural features such as defects — free deformation bands resulting from dislocations motion along family plans , that swept defects or γ' precipitates and lead to deformation localization. In order to quantify effects of plastic localized deformation on intergranular stress corrosion cracking (IGSCC) of the A286 alloy in PWR primary water, slow strain rate tests are conducted. For each cycling conditions, two specimens at a similar stress level are tested: the first containing free precipitate deformation bands, the other not significant of a localized deformation state. SSRT tests are still in progress.

  19. Replacement of Co-base alloy for radiation exposure reduction in the primary system of PWR

    International Nuclear Information System (INIS)

    Han, Jeong Ho; Nyo, Kye Ho; Lee, Deok Hyun; Lim, Deok Jae; Ahn, Jin Keun; Kim, Sun Jin

    1996-01-01

    Of numerous Co-free alloys developed to replace Co-base stellite used in valve hardfacing material, two iron-base alloys of Armacor M and Tristelle 5183 and one nickel-base alloy of Nucalloy 488 were selected as candidate Co-free alloys, and Stellite 6 was also selected as a standard hardfacing material. These four alloys were welded on 316SS substrate using TIG welding method. The first corrosion test loop of KAERI simulating the water chemistry and operation condition of the primary system of PWR was designed and fabricated. Corrosion behaviors of the above four kinds of alloys were evaluated using this test loop under the condition of 300 deg C, 1500 psi. Microstructures of weldment of these alloys were observed to identify both matrix and secondary phase in each weldment. Hardnesses of weld deposit layer including HAZ and substrate were measured using micro-Vickers hardness tester. The status on the technology of Co-base alloy replacement in valve components was reviewed with respect to the classification of valves to be replaced, the development of Co-free alloys, the application of Co-free alloys and its experiences in foreign NPPs, and the Co reduction program in domestic NPPs and industries. 18 tabs., 20 figs., 22 refs. (Author)

  20. Structural analysis of surface film on alloy 600 formed under environment of PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Terachi, Takumi; Totsuka, Nobuo; Yamada, Takuyo; Nakagawa, Tomokazu [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan); Deguchi, Hiroshi [Kansai Electric Power Co., Inc., Osaka (Japan); Horiuchi, Masaki; Oshitani, Masato [Kanden Kako Co., Ltd., Osaka (Japan)

    2002-09-01

    It has been shown by one of the present authors and so forth that PWSCC of alloy 600 relates to dissolved hydrogen concentration (DH) in water and oxide film structure. However, the mechanism of PWSCC has not been clear yet. Therefore, in order to investigate relationship between them, structural analysis of the oxide film formed under the environment of PWR primary water was carried out by using X-ray diffraction, the scanning electron microscope and the transmission electron microscope. Especially, to perform accurate analysis, the synchrotron orbital radiation with SPring-8 was tried to use for thin film X-ray diffraction measurement. From the results, observed are as follows: 1. the oxide film is mainly composed of NiO, under the condition without hydrogen. 2. In the environment of DH 2.75ppm, the oxide film forms thin spinel structures. 3. On the other hand, needlelike oxides are formed at DH 1ppm. For this reason, around 1ppm of DH there would be the boundary that stable NiO and spinel oxide generate, and it agrees with the peak range of the PWSCC susceptibility on hydrogen. From this, it is suggested that the boundary of NiO/spinel oxide affects the SCC susceptibility. (author)

  1. Influence of dissolved hydrogen on oxide film and PWSCC of Alloy 600 in PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tomokazu; Totsuka, Nobuo; Nakajima, Nobuo [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    In order to investigate the influence of dissolved hydrogen (DH) on the corrosion behavior and PWSCC of Alloy 600 in primary water of PWR under actual operating temperature range, we carried out electrochemical polarization measurement, repassivation test, analysis of the oxide film on the alloy by AES, XPS and PWSCC test. In all cases, the content of DH was changed from 0 to 45 cc/kgH{sub 2}O. The anodic polarization curve reveals that the peak current density increases with increasing DH. The result of the repassivation test shows that the repassivation rate decreases with increasing DH, and the changes of the above two become larger between 11 and 22 cc/kgH{sub 2}O of DH. According to the results of oxide film analysis, it is seen that the oxide films formed below 11 cc/kgH{sub 2}O of DH are relatively thick and rich in Ni, but those formed at higher DH contents are relatively thin and rich in Cr and Fe. The susceptibility of the alloy to PWSCC has a peak at 11 cc/kgH{sub 2}O of DH, which reveals that the property of the oxide film may play important role in PWSCC of alloy. (author)

  2. Analysis of corrosion product transport in PWR primary system under non-convective condition

    International Nuclear Information System (INIS)

    Han, Byoung Sub

    1992-02-01

    Product TRANsport), which can predict the corrosion product and radioactivity transport within the primary coolant system, and also can be utilized for the computer simulation with actual plant data of currently operating Korean nuclear power plants to predict the transport of the radionuclides. In this study, the following problems will be updated, improved and compared with the already existing codes: 1) development and analysis of recent mechanistic modelling of corrosion product deposition, 2) application and modification due to the temperature kinetic effect, 3) separation of the effect of Fe, Co, Ni and Mn solubility rather than Fe solubility alone, and 4) consideration of Ni activation and recoil process. By applying the above updated and improved mechanisms, the corrosion product behavior in PWR of currently operating Korean unclear power plants has been simulated. In addition, the evaluation of particulate transport, independent solubility data of major radionuclides and acute nodalization were included and extended. Then, with the developed computer code, we have evaluated and analyzed the activity and corrosion product build-up controlled by many parameters such as pH, composition of metal, and auxiliary system performance

  3. BWR water chemistry guidelines and PWR primary water chemistry guidelines in Japan – Purpose and technical background

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hirotaka, E-mail: kawamuh@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (Japan); Hirano, Hideo [Central Research Institute of Electric Power Industry (Japan); Katsumura, Yousuke [University of Tokyo (Japan); Uchida, Shunsuke [Tohoku University (Japan); Mizuno, Takayuki [Mie University (Japan); Kitajima, Hideaki; Tsuzuki, Yasuo [Japan Nuclear Safety Institute (Japan); Terachi, Takumi [Institute of Nuclear Safety System, Inc. (Japan); Nagase, Makoto; Usui, Naoshi [Hitachi-GE Nuclear Energy, Ltd. (Japan); Takagi, Junichi; Urata, Hidehiro [Toshiba Corporation (Japan); Shoda, Yasuhiko; Nishimura, Takao [Mitsubishi Heavy Industry, Ltd. (Japan)

    2016-12-01

    Highlights: • Framework of BWR/PWR water chemistry Guidelines in Japan are presented. • Guideline necessity, definitions, philosophy and technical background are mentioned. • Some guideline settings for control parameters and recommendations are explaines. • Chemistry strategy is also mentioned. - Abstract: After 40 years of light water reactor (LWR) operations in Japan, the sustainable development of water chemistry technologies has aimed to ensure the highest coolant system component integrity and fuel reliability performance for maintaining LWRs in the world; additionally, it aimed to achieve an excellent dose rate reduction. Although reasonable control and diagnostic parameters are utilized by each boiling water reactor (BWR) and pressurized water reactor (PWR) owner, it is recognized that specific values are not shared among everyone involved. To ensure the reliability of BWR and PWR operation and maintenance, relevant members of the Atomic Energy Society of Japan (AESJ) decided to establish guidelines for water chemistry. The Japanese BWR and PWR water chemistry guidelines provide strategies to improve material and fuel reliability performance as well as to reduce dosing rates. The guidelines also provide reasonable “control values”, “diagnostic values” and “action levels” for multiple parameters, and they stipulate responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the Japanese PWR primary water chemistry guidelines. Good practices for operational conditions are also discussed with reference to long-term experience. This paper presents the purpose, technical background and framework of the preliminary water chemistry guidelines for Japanese BWRs and PWRs. It is expected that the guidelines will be helpful as an introduction to achieve safety and reliability during operations.

  4. The hold-time effects on the low cycle fatigue behaviors of 316 SS in PWR primary environment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Junho; Hong, Jong-Dae; Seo, Myung-Gyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The effects of the environments on fatigue life of the structural materials used in nuclear power plants (NPPs) were known to be significant according to the extensive test results. Accordingly, the fatigue analysis procedures and the design fatigue curves were proposed in the ASME Code. However, the implication that the existing ASME design fatigue curves did not sufficiently reflect the effect of the operation conditions of nuclear power plants emerged as an issue to be resolved. One of possible reasons to explain the discrepancy is that the laboratory test conditions do not represent the actual plant transients. Therefore, it is necessary to clarify the effects of light water environments on fatigue life while considering more plant-relevant transient conditions such as hold-time. For this reason, this study will focus on the fatigue life of type 316 stainless steel (SS) in the pressurized water reactor (PWR) environments while incorporating the hold-time during the low cycle fatigue (LCF) test in simulated PWR environments. The objective of this study is to characterize the effects of hold-time on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 SS in 310 .deg. C air and simulated PWR environments. To simulate the heat-up and cool-down transient, sub-peak strain holding during the down-hill of strain amplitude was chosen. Currently, LCF tests with 60 seconds holding are in progress. The 0.4, 0.04%/s strain rate condition test results are presented in this study, which shows somewhat longer fatigue life.

  5. The hold-time effects on the low cycle fatigue behaviors of 316 SS in PWR primary environment

    International Nuclear Information System (INIS)

    Lee, Junho; Hong, Jong-Dae; Seo, Myung-Gyu; Jang, Changheui

    2015-01-01

    The effects of the environments on fatigue life of the structural materials used in nuclear power plants (NPPs) were known to be significant according to the extensive test results. Accordingly, the fatigue analysis procedures and the design fatigue curves were proposed in the ASME Code. However, the implication that the existing ASME design fatigue curves did not sufficiently reflect the effect of the operation conditions of nuclear power plants emerged as an issue to be resolved. One of possible reasons to explain the discrepancy is that the laboratory test conditions do not represent the actual plant transients. Therefore, it is necessary to clarify the effects of light water environments on fatigue life while considering more plant-relevant transient conditions such as hold-time. For this reason, this study will focus on the fatigue life of type 316 stainless steel (SS) in the pressurized water reactor (PWR) environments while incorporating the hold-time during the low cycle fatigue (LCF) test in simulated PWR environments. The objective of this study is to characterize the effects of hold-time on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 SS in 310 .deg. C air and simulated PWR environments. To simulate the heat-up and cool-down transient, sub-peak strain holding during the down-hill of strain amplitude was chosen. Currently, LCF tests with 60 seconds holding are in progress. The 0.4, 0.04%/s strain rate condition test results are presented in this study, which shows somewhat longer fatigue life

  6. Primary cooling system for BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Eishi; Takahashi, Masanori; Aoki, Yasuko

    1993-01-01

    The present invention effectively uses information from a plurality of sensors in order to suppress corrosion circumstance of a nuclear reactor. That is, a predetermined general water quality factor at a predetermined position is determined as a standard index. A concentration of a water quality improver is controlled such that the index is within an aimed range. For this purpose, the entire sensor groups disposed in a primary coolant system of a nuclear reactor are divided into a plural systems of sensor groups each disposed on every different positions. Then, a predetermined sensor group (standard sensor group) is connected to a computing device and a data base so that it is always monitored for calculating and estimating the standard index. Only oxidative ingredient in water at the measuring point is noted, and a concentration distribution which agrees with an actually measured value of oxidative ingredients is extracted from data base and used as a correct concentration distribution. With such procedures, reactor water quality can be estimated accurately while compensating erroneous factors of individual sensors. Even when a new sensor is used, it is not necessary to greatly change control logic. (I.S.)

  7. Preoperational test report, primary ventilation condenser cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Clifton, F.T.

    1997-10-29

    This represents the preoperational test report for the Primary Ventilation Condenser Cooling System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system uses a closed chilled water piping loop to provide offgas effluent cooling for tanks AY101, AY102, AZ1O1, AZ102; the offgas is cooled from a nominal 100 F to 40 F. Resulting condensation removes tritiated vapor from the exhaust stack stream. The piping system includes a package outdoor air-cooled water chiller with parallel redundant circulating pumps; the condenser coil is located inside a shielded ventilation equipment cell. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

  8. Preoperational test report, primary ventilation condenser cooling system

    International Nuclear Information System (INIS)

    Clifton, F.T.

    1997-01-01

    This represents the preoperational test report for the Primary Ventilation Condenser Cooling System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system uses a closed chilled water piping loop to provide offgas effluent cooling for tanks AY101, AY102, AZ1O1, AZ102; the offgas is cooled from a nominal 100 F to 40 F. Resulting condensation removes tritiated vapor from the exhaust stack stream. The piping system includes a package outdoor air-cooled water chiller with parallel redundant circulating pumps; the condenser coil is located inside a shielded ventilation equipment cell. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System

  9. Implementation of the structural integrity analysis for PWR primary components and piping

    International Nuclear Information System (INIS)

    Pellissier-Tanon, A.

    1982-01-01

    The trends on the definition, the assessment and the application of fracture strength evaluation methodology, which have arisen through experience in the design, construction and operation of French 900-MW plants are reviewed. The main features of the methodology proposed in a draft of Appendix ZG of the RCC-M code of practice for the design verification of fracture strength of primary components are presented. The research programs are surveyed and discussed from four viewpoints, first implementation of the LEFM analysis, secondly implementation of the fatigue crack propagation analysis, thirdly analysis of vessel integrity during emergency core cooling, and fourthly methodology for tear fracture analysis. (author)

  10. Assessment of cracked pipes in primary piping systems of PWR nuclear reactors

    International Nuclear Information System (INIS)

    Jong, Rudolf Peter de

    2004-01-01

    Pipes related to the Primary System of Pressurized Water Reactors (PWR) are manufactured from high toughness austenitic and low alloy ferritic steels, which are resistant to the unstable growth of defects. A crack in a piping system should cause a leakage in a considerable rate allowing its identification, before its growth could cause a catastrophic rupture of the piping. This is the LBB (Leak Before Break) concept. An essential step in applying the LBB concept consists in the analysis of the stability of a postulated through wall crack in a specific piping system. The methods for the assessment of flawed components fabricated from ductile materials require the use of Elasto-Plastic Fracture Mechanics (EPFM). Considering that the use of numerical methods to apply the concepts of EPFM may be expensive and time consuming, the existence of the so called simplified methods for the assessment of flaws in piping are still considered of great relevance. In this work, some of the simplified methods, normalized procedures and criteria for the assessment of the ductile behavior of flawed components available in literature are described and evaluated. Aspects related to the selection of the material properties necessary for the application of these methods are also discussed. In a next .step, the methods are applied to determine the instability load in some piping configurations under bending and containing circumferential through wall cracks. Geometry and material variations are considered. The instability loads, obtained for these piping as the result of the application of the selected methods, are analyzed and compared among them and with some experimental results obtained from literature. The predictions done with the methods demonstrated that they provide consistent results, with good level of accuracy with regard to the determination of maximum loads. These methods are also applied to a specific Study Case. The obtained results are then analyzed in order to give

  11. Magnetite solubility studies under simulated PWR primary-side conditions, using lithiated, hydrogenated water

    International Nuclear Information System (INIS)

    Hewett, John; Morrison, Jonathan; Cooper, Christopher; Ponton, Clive; Connolly, Brian; Dickinson, Shirley; Henshaw, Jim

    2014-01-01

    As software for modelling dissolution, precipitation, and transport of metallic species and subsequent CRUD deposition within nuclear plant becomes more advanced, there is an increasing need for accurate and reliable thermodynamic data. The solubility behaviour of magnetite is an example of such data, and is central to any treatment of CRUD solubility due to the prevalence of magnetite and nickel ferrites in CRUD. Several workers have shown the most consistent solubility data comes from once-through flowing systems. However, despite a strong consensus between the results in acidic to mildly alkaline solutions, there is disagreement between the results at approximately pH 25C 9 and higher. A programme of experimental work is on-going at the University of Birmingham, focusing on solubility of metal oxides (e.g., magnetite) in conditions relevant to PWR primary coolant. One objective of this programme is to calculate thermodynamic constants from the data obtained. Magnetite solubility from 200 to 300°C, in lithiated, hydrogenated water of pH 25C 9–11 is being studied using a once-through rig constructed of 316L stainless steel. The feedwater is pumped at 100 bar pressure through a heated bed of magnetite granules, and the output solution is collected and analysed for iron and several other metals by ICP-MS. This paper presents results from preliminary tests without magnetite granules, in which the corroding surface of the rig itself was used as the sole source of soluble iron and of dissolved hydrogen. Levels of iron were generally within an order of magnitude of literature solubility values. Comparison of results at different flow rates and temperatures, in conjunction with conclusions drawn from the published literature, suggests that this is likely due to the presence of particulate matter in a greatly under-saturated solution, compensating for the low surface area of oxide in contact with the solution. (author)

  12. Fatigue crack growth threshold of austenitic stainless steels in simulated PWR primary water

    International Nuclear Information System (INIS)

    Tsutsumi, Kazuya; Yamamoto, Kenji; Nitta, Yoshikazu

    2007-01-01

    Many studies have revealed that fatigue crack growth (FCG) rate of austenitic stainless steels is accelerated in light water reactor environment compared to that in air at room temperature. Major driving factors in the acceleration of FCG rate are stress ratio, temperature and stress rise time. Based on this knowledge, FCG curves have been developed considering these factors as parameters. However, there are few data of FCG threshold ΔK th in light water reactor environment. Hence it is necessary to clarify FCG rate under near-threshold condition for more accurate evaluation of fatigue crack growth behavior under cyclic stress with relatively low ΔK. In the present study, therefore, ΔK th was determined for austenitic stainless steels in simulated PWR primary water, and FCG behavior under near-threshold condition was revealed by collecting fatigue crack propagation data. The results are summarized as follows: No propagation of fatigue crack was found in high temperature water, and there was a definite ΔK th . Average ΔK eff,th was 4.3 MPa·m 0.5 at 325degC, 3.3 MPa·m 0.5 at 100degC, and there was no considerable reduction compared to currently known ΔK eff,th in air. Thus, it was revealed tha ambient conditions had minimal effect, on ΔK eff,th , ΔK th increases with increasing temperature and decreasing frequency. As a result of fracture surface observation, oxide-induced-crack-closure was considered to be a cause of the dependency described above. In addition, it was suggested that changes in material properties also had influence on ΔK th, since ΔK eff,th itself increased at elevated temperature. (author)

  13. Environmental effect on cracking of an 304L austenitic stainless steels in PWR primary environment under cyclic loading

    International Nuclear Information System (INIS)

    Huin, N.

    2013-01-01

    The present study was undertaken in order to get further insights on cracking mechanisms in a 304L stainless steel. More precisely, a first objective of this study was to evaluate the effect of various cold working conditions on the cyclic stress-strain behavior and the fatigue life in air and in PWR primary environment. In air a prior hardening was found to reduce the fatigue life in the LCF regime but not in primary environment. In both environments, the fatigue limit of the hardened materials was increased after cold working.The second objective addresses the effect of the air and the PWR primary environments on the cracking mechanisms (initiation and propagation) in the annealed material in the LCF regime. More precisely, the kinetics of crack initiation and micro crack propagation were evaluated with a multi scale microscopic approach in air and in primary environment. In PWR primary environment, during the first cycles, preferential oxidation occurs along emerging dissociated dislocation and each cycle generates a new C-rich/Fe-rich oxide layer. Then, during cycling, the microstructure evolves from stacking fault into micro twinning and preferential oxidation occurs by continuous shearing and dissolution of the passive film. Beyond a certain crack depth (≤3 μm), the crack starts to propagate with a direction close to a 90 degrees angle from the surface. The crack continues its propagation by successive generation of shear bands and fatigue striations at each cycle up to failure. The role of corrosion hydrogen on these processes is finally discussed. (author)

  14. Appropriate zinc addition management into PWR primary coolant after the plant long-term maintenance

    International Nuclear Information System (INIS)

    Hirose, Atsushi; Matsui, Ryo; Imamura, Haruki; Takahashi, Akira; Shimizu, Yuichi; Kogawa, Noritaka; Nagamine, Kunitaka

    2014-01-01

    Zinc addition into the PWR primary coolant is known as an effective method to reduce the radioactivity build up. The reduction effect has been confirmed by actual plant experience of the Genkai Nuclear Power Plant Unit 1 to 4 and the Sendai Nuclear Power Plant Unit 1 to 2 which are operated by Kyushu Electric Power Co. in Japan. Zinc addition is suspended at shut-down, and is resumed after heat up or arrival at full power. In usual maintenance, the period when zinc addition is not applied is short; thus it is considered that suspension of zinc addition does not have practical influence on the corrosion and the radioactivity buildup in the oxide layer of surface for the primary equipment and piping. On the other hand, in case the maintenance period is much longer, the new oxide which does not contain zinc has grown, and then the structure of the oxide layer may be changed. Therefore, it is considered that zinc addition suspension in long-term period has possibilities to deteriorate the dose reduction effect. In order to verify the effect of long-term suspension of zinc addition upon oxide layer, the lab experiment was carried out using TT690 alloy which is the constitution material of the steam generator tubes under the conditions of long-term and the subsequent resuming operations. After the experiment, the specimens were analyzed by IMA and chemical analysis. These measurement results suggest the difference of the oxide layer is little or none between long-term suspension of zinc addition and short-term suspension of zinc suspension. Hence it is considered that influence of long-term maintenance on the oxide layer is small. Furthermore, in this study, in order to evaluate the influence of the suspension of zinc addition in the operation period, specimens of oxide film formed with zinc were carried out the corrosion test in the simulated RCS condition without zinc. These measurement results indicate the effect of reduction of the activity build up will become less

  15. Development status of nuclear power in China and fundamental research progress on PWR primary water chemistry in China

    International Nuclear Information System (INIS)

    Wu, Xinqiang; Liu, Xiahe; Han, En-Hou; Ke, Wei; Xu, Yuming

    2015-01-01

    China's non-fossil fuels are expected to reach 20% in primary energy ratio by 2030. It is urgent for China to speed up the development of nuclear power to increase energy supply, reduce gas emissions and optimize resource allocation. Chinese government slowed down the approval of new nuclear power plant (NPP) projects after Fukushima accident in 2011. At the end of 2012, the State Council approved the nuclear safety program and adjusted long-term nuclear power development plan (2011-2020), the new NPP's projects have been restarted. In June 2015, there are 23 operating units in mainland in China with total installed capacity of about 21.386 GWe; another 26 units are under construction with total installed capacity of 28.5 GWe. The main type of reactors in operation and under construction in China is pressurized water reactor (PWR), including the first AP1000 NPPs in the world (units 1 in Sanmen) and China self-developed Hualong one NPPs (units 5 and 6 in Fuqing). Currently, China's nuclear power development is facing historic opportunities and also a series of challenges. One of the most important is the safety and economy of nuclear power. The optimization of primary water chemistry is one of the most effective ways to minimize radiation field, mitigate material degradation and maintain fuel performance in PWR NPPs, which is also a preferred path to achieve both safety and economy for operating NPPs. In recent years, an increased attention has been paid to fundamental research and engineering application of PWR primary water chemistry in China. The present talk mainly consists of four parts: (1) development status of China's nuclear power industry; (2) safety of nuclear power and operating water chemistry; (3) fundamental research progress on Zn-injected water chemistry in China; (4) summary and future. (author)

  16. Behaviour of radiation fields in the Spanish PWR by the changes in coolant chemistry and primary system materials

    International Nuclear Information System (INIS)

    Llovet, R.; Fernandez Lillo, E.

    1995-01-01

    The Spanish PWR Owners Group established a program to evaluate the behavior of ex-core radiation fields and discriminate the effects of changes in coolant chemistry and primary system materials. Data from Vandellos, Asco, Almaraz and Trillo NPPs were analyzed Vandellos 2 was chosen as the lead plant and its data were thoroughly studied. The dose-rates evolution could be explained at each plant as a consequence of this sucessful program.Actions derived from the developed knowledge on this field have produced the stabilization or even reduction of radiation fields at these plants

  17. Probabilistic study of primary pump trip in a P.W.R. reactor: use of response surface methodology

    International Nuclear Information System (INIS)

    Bars, C.; Duchemin, B.; Maigret, N.; Peltier, J.; Rostan, O.; Villeneuve, M.J. de; Lanore, J.M.

    1981-09-01

    This paper describes a probabilistic study about the consequences of the trip or blockage of one of the three PWR reactor primary pumps. The distribution of the input parameters is taken into account and the resulting distribution of the consequence (number of failed fuel rods) is assessed. The necessity to do this study with the response surface methodology and the precautions to take are outlined. The results show that the probability to have failed fuel rods is about 10 -4 for pump trip and 0.16 for blockage with, in this case, a mean of 196 failed rods, that is 0.5 % of total number of rods

  18. Control of the flanges of the thermal barriers fitting the 900 MWe PWR primary pumps

    International Nuclear Information System (INIS)

    Cleurennec, M.; Thebault, Y.; Abittan, E.; Pages, C.; Lhote, P.A.; Randrianarivo, L.

    1998-01-01

    During maintenance visit on 93 D type primary pumps of French 900 MWe nuclear units, cracking has been evidenced on the thermal barrier, first on the flange, on the face of connection of the cooling, water coils, and then on the weld between the housing and the flange. Laboratory examinations have exhibited that this cracking is due to a fatigue phenomenon which is initiated on locations where high residual stresses are present. One pump, in service in a plant, has received an instrumentation in order to determine stress cycling. Measurements of temperature on the surface of the metal have shown the presence of thermal cycling due to the thermohydraulic conditions inside the thermal barrier. A non destructive testing method using ultrasounds has been developed in order to asses the magnitude cracking. Corrective and preventive actions have been implemented for repairing and improving thermal barrier when cracking is detected. (authors)

  19. Verification on reliability of heat exchanger for primary cooling system

    International Nuclear Information System (INIS)

    Koike, Sumio; Gorai, Shigeru; Onoue, Ryuji; Ohtsuka, Kaoru

    2010-07-01

    Prior to the JMTR refurbishment, verification on reliability of the heat exchangers for primary cooling system was carried out to investigate an integrity of continuously use component. From a result of the significant corrosion, decrease of tube thickness, crack were not observed on the heat exchangers, and integrity of heat exchangers were confirmed. In the long terms usage of the heat exchangers, the maintenance based on periodical inspection and a long-term maintenance plan is scheduled. (author)

  20. Design Of Pump Monitoring Of Primary Cooling System

    International Nuclear Information System (INIS)

    Indrakoesoema, Koes; Sujarwono

    2000-01-01

    Monitoring of 3 primary cooling pumps done visually by operator on the spot. The operator must be check oil in a sight glass, oil leakage during pump operation and water leakage. If reaktor power increase about more than 3 MW, the radiation exposure also increase in the primary cell and that's way the operator can not check the pumps. To continuing monitor all pump without delay, one system has been added I.e Closed Circuit Television (CCTV). This system using 3 video camera to monitor 3 pumps and connected to one receiver video monitor by coaxial cable located in Main Control Room. The sequence monitoring can be done by sequential switcher

  1. Primary circuit leak detection an application on PWR vessel head penetrations

    International Nuclear Information System (INIS)

    Loisy, F.; Germain, J.L.; Chauvel, L.

    1996-01-01

    In 1991, cracks were discovered and localized in the lower part of certain vessel head adapters in EDF PWR units. While awaiting the replacement of the vessel heads in question, EDF developed systems to enable continuous monitoring of vessel head penetration, by means of early detection of leaks. One of these systems in based on detection of water vapour in a confined space above the vessel head. The efficiency of the measurement chain is particularly dependent on dilution of the leakage in the confined space prior TO entry in the sampling circuit. The detection threshold for this method is on the order of 1.2 liters/hour for a dilution rate of 1500 rate of 1500 m 3 /h and a dew point of 22 deg C. This system has now been in operation on three 1300-MW PWR units for three years, and has proved to function satisfactorily. (authors)

  2. Crack growth testing of cold worked stainless steel in a simulated PWR primary water environment to assess susceptibility to stress corrosion cracking

    International Nuclear Information System (INIS)

    Tice, D.R.; Stairmand, J.W.; Fairbrother, H.J.; Stock, A.

    2007-01-01

    Although austenitic stainless steels do not show a high degree of susceptibility to stress corrosion cracking (SCC) in PWR primary environments, there is limited evidence from laboratory testing that crack propagation may occur under some conditions for materials in a cold-worked condition. A test program is therefore underway to examine the factors influencing SCC propagation in good quality PWR primary coolant. Type 304 stainless steel was subjected to cold working by either rolling (at ambient or elevated temperature) or fatigue cycling, to produce a range of yield strengths. Compact tension specimens were fabricated from these materials and tested in simulated high temperature (250-300 o C) PWR primary coolant. It was observed that the degree of crack propagation was influenced by the degree of cold work, the crack growth orientation relative to the rolling direction and the method of working. (author)

  3. Effect of hardening on the crack growth rate of austenitic stainless steels in primary PWR conditions

    International Nuclear Information System (INIS)

    Castano, M.L.; Garcia, M.S.; Diego, G. de; Gomez-Briceno, D.; Francia, L.

    2002-01-01

    Intergranular cracking of non-sensitized materials, found in light water reactor (LWR) components exposed to neutron radiation, has been attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC). Cracking of baffle former bolts, fabricated of AISI-316L and AISI-347, have been reported in some Europeans and US PWR plants. Examinations of removed bolts indicate the intergranular cracking characteristics can be associated with IASCC phenomena. Neutron radiation produce critical modifications of the microstructure and microchemical of stainless steels such hardening due to irradiation and Radiation Induce Segregation (RIS) at grain boundaries, among others. Chromium depletion at grain boundary due to RIS seems to justify the intergranular cracking of irradiated materials, both in plant and in lab tests, at high electrochemical corrosion potential (BWR-NWC environments), but it is not enough to explain cracking at low corrosion potential (BWR-HWC and PWR environments). In these latter conditions, hardening is considered a possible additional mechanism to explain the behavior of irradiated material. Radiation Hardening can be simulated in non irradiated material by mechanical deformation. Although some differences exists in the types of defects produced by radiation and mechanical deformation, it is accepted that the study of the stress corrosion behavior of unirradiated austenitic steels with different hardening levels would contribute to the understanding of IASCC mechanism. In order to evaluate the influence of hardening on the stress corrosion susceptibility of austenitic steels, crack growth rate tests with 316L and 347 stainless steels with nominal yield strengths from 500 to 900 MPa, produced by cold work are being carried out at 340 deg C in PWR conditions. Preliminary results indicate that crack propagation was obtained in the 316Lss and 347ss cold worked, even with a yield strength of 550 MPa. (authors)

  4. The application of transition metal ion chromatography to the determination of elemental and radiochemical species in PWR primary coolant

    International Nuclear Information System (INIS)

    Bridle, D.A.; Brown, G.R.; Johnson, P.A.V.

    1992-01-01

    The accurate determination of both elemental and radiochemical transition metal corrosion products, particularly cobalt and nickel, in PWR coolants is necessary if the transport mechanisms and their role in the development of out-of-core radiation fields are to be fully understood. AEA Technology, Winfrith, has collaborated for several years with a number of PWR utilities in Europe, developing advanced sampling and analytical techniques for the determination of both soluble and insoluble corrosion products in primary coolant. The design and installation of continuously flowing isokinetic capillary modifications to the existing sampling systems has been shown to be an effective method of providing a low, but representative, sample flow from high pressure systems for on-line determination of corrosion product species. Transition metal ion chromatography coupled with gamma-spectrometry has been used to determine both insoluble and soluble elemental and radiochemical species in reactor coolant, with particular attention being given to the determination of soluble elemental cobalt at levels as low as 1 ng per kg. Soluble species were determined directly following their concentration from up to 1 litre of coolant. Insoluble species collected on 0.45 micron filter membranes, following filtration of up to 1500 litres of coolant, were solubilised by fusion with potassium hydrogen sulphate before the application of ion chromatography. In each case the eluant from the chromatographic column was collected and the radionuclides determined by gamma-spectrometry

  5. Relationship between stress corrosion cracking and low frequency fatigue-corrosion of alloy 600 in PWR primary water

    International Nuclear Information System (INIS)

    Bosch, C.

    1998-01-01

    Stress corrosion cracking of PWR vessel head adapters is a main problem for nuclear industry. With the aim to better understand the influence of the mechanical parameters on the cracking phenomena (by stress corrosion (SCC) or fatigue corrosion (FC)) of alloy 600 exposed to primary PWR coolant, a parametrical study has been carried out. Crack propagation tests on CT test specimens have been implemented under static loads (stress corrosion tests) or low frequency cyclic loads (fatigue corrosion tests). Results (frequency influence, type of cycles, ratio charge on velocities and propagation modes of cracks) have allowed to characterize the transition domain between the crack phenomena of SCC and FC. With the obtained results, it has been possible too to differentiate the effects due to environmental factors and the effects due to mechanical factors. At last, a quantitative fractographic study and the observations of the microstructure at the tip of crack have led to a better understanding of the transitions of the crack propagation mode between the SCC and the FC. (O.M.)

  6. Study of the influence of temperature and time on the electroplating nickel layer in Inconel 718 strips used in spacer grid of Pressurized Water Cooled nuclear reactors (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Rezende, Renato; Abati, Amanda; Verne, Júlio; Panossian, Zehbour, E-mail: amanda.abati@marinha.mil.br, E-mail: jvernegropp@gmail.com, E-mail: renato.rezende@marinha.mil.br, E-mail: zep@ipt.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil). Laboratório de Desenvolvimento e Instrumentação de Combustível Nuclear; Instituto de Pesquisas Tecnológicas (IPT), São Paulo, SP (Brazil)

    2017-07-01

    The Inconel 718 (UNS N07718: Ni-{sup 19}Cr-{sup 18}Fe-{sup 5}Nb-3 Mo) is a precipitation hardenable nickel alloy that has good corrosion resistance and high mechanical strength. These strips are used for assembling the spacer grid of fuel element of pressurized water cooled nuclear reactors (PWR). The spacer grid is a structural component of fundamental importance in fuel elements of PWR reactors, maintaining the position and necessary spacing of the fuel rods within the arrangement of the fuel element. The spacer grid is formed by joining the points of intersection of the strips, by a joint process called brazing. For this process, these strips are stamped and plated with a thin layer of nickel by means of electroplating in order to protect against oxidation and allow a better flowability and wettability of the addition metal in the strips during brazing. Oxidation at the surface of the base material harms wettability and inhibits spreading of the liquid addition metal on the substrate surface during the brazing process. The use of coatings such as nickel plating is used to ensure such conditions. The results showed that there is a process of diffusion de some chemical elements such as chromium, iron, titanium and aluminum from the substrate to the nickel layer and nickel from the layer to the substrate. These chemical elements are responsible for the oxidation at the surface of the strip. (author)

  7. Evaluation of feed and bleed cooling mode in case of total loss of feedwater on 900 MWe PWR

    International Nuclear Information System (INIS)

    Champ, M.; Cornille, Y.

    1989-07-01

    The physical studies carried out with the CATHARE code to assess the feed and bleed procedure developed in order to cope with the total loss of feed water on a 900 MWe PWR are presented. These studies allowed the definition of the maximum delays of intervention which would prevent the core from uncovering. Different cases of equipment availability are considered. The data generated will be used in the 900 MWe Probabilistic Safety Assessment which is under way at the Institut de Protection et de Surete Nucleaire

  8. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Steady state and transients computations

    International Nuclear Information System (INIS)

    Martin, A.; Alvarez, D.; Cases, F.

    1996-03-01

    The paper explains the chronological account and the first results obtained in the R and D program on the mixing in the 900 MW PWR vessels. After the presentation of the plant type simulated, we define the numerical tool, the (Finite Element Modelling) FEM N3S code. Two results are presented with a comparison with the experiment results issued of the BORA BORA mock up. The first case is dealing with the isothermal steady state mixing in the vessel with the three loops mass flow rate balanced. This case identified as a validation of our numerical tool shows a good agreement. The second case is dealing with the transient mixing of a clear plug in the vessel when one primary pump starts-up. We compare the numerical and experiment results giving the mean boron concentration at the core inlet for several clear water plugs. The results show again a good agreement. (authors). 12 refs., 10 figs., 1 tab

  9. Effect of Ni and Cr on IGSCC growth rate of Ni-Cr-Fe alloys in PWR primary water

    International Nuclear Information System (INIS)

    Arioka, K.; Yamada, T.; Aoki, M.; Miyamoto, T.

    2015-01-01

    The purpose of this research is to examine the dependence of SCC (Stress Corrosion Crack) growth on nickel and chromium in PWR primary water; the objective is to obtain the basic knowledge to understand SCC behavior of steam generator tubing materials. The second objective is to understand whether accelerated testing at higher temperatures is appropriate for predicting SCC initiation and growth at lower temperatures. For these objectives, SCC growth was measured in PWR primary water at 290, 320, 330, 340, and 360 C. degrees under static load conditions. Tests were performed using 0.5 T compact tension type specimen using 20%CW X%Ni-16%Cr-Fe alloys in the range of nickel concentration between 16 to 60% and laboratory melted nuclear grade 20% cold worked Alloy 800 (USN N08800, CW800NG). Four important patterns were observed. First, significant effect of nickel on IGSCC resistance was observed at 340 and 360 C. degrees. The rate of IGSCC growth decreases with increasing nickel concentration in the range of nickel concentration between 10% to 25% nickel; and then, the rate of IGSCC increases with increasing nickel concentration in the range of Ni content between 50% and 76%. This trend is quite similar to the results reported by Coriou and Staehle tested in deaerated pure water at 350 C. degrees. However, no significant dependence of Ni content on IGSCC in PWR water at 320 and 290 C. degrees was observed. The change in SCC growth dependence on nickel concentration suggested that the main rate limiting processes on IGSCC growth seems to change between 320 and 340 C. degrees. Secondly, significant beneficial effects of chromium in alloys were observed at 320 C. degrees. However, no beneficial effect of chromium addition in alloys was observed at 360 C. degrees. Thirdly, peak temperatures in growth rate of IGSCC were observed in almost all test materials except for 20%CW Alloy 600. Finally, intergranular attack was observed in some alloys at lower temperature, and the

  10. SCC of cold-worked austenitic stainless steels exposed to PWR primary water conditions: susceptibility to initiation

    International Nuclear Information System (INIS)

    Herms, E.; Raquet, O.; Sejourne, L.; Vaillant, F.

    2009-01-01

    Heavily cold-worked austenitic stainless steels (AISI 304L and 316L types) could be significantly susceptible to Stress Corrosion Cracking (SCC) when exposed to PWR nominal primary water conditions even in absence of any pollutants. Susceptibility to SCC was shown to be related with some conditions such as initial hardness, procedure of cold-work or dynamic straining. A dedicated program devoted to better understand the initiation stage on CW austenitic stainless steels in PWR water is presented. Initiation is studied thanks to SCC test conditions leading to an intergranular cracking propagation mode on a CW austenitic stainless steel which is the mode generally reported after field experience. SCC tests are carried out in typical primary water conditions (composition 1000 ppm B and 2 ppm Li) and for temperature in the range 290 - 340 C. Material selected is 316L cold-worked essentially by rolling (reduction in thickness of 40%). Initiation tests are carried out under various stress levels with the aim to investigate the evolution of the initiation period versus the value of applied stress. SCC tests are performed on cylindrical notched specimens in order to increase the applied stress and allow accelerated testing without modify the exposure conditions to strictly nominal hydrogenated PWR water. Respective influences of cyclic/dynamic conditions on SCC initiation are presented and discussed. Dedicated interrupted tests help to investigate the behaviour of the crack initiation process. These SCC tests have shown that crack initiation could be obtained after a very short time under dynamic loading conditions on heavily pre-strained austenitic stainless steels. Actual results show that the most limiting stage of the cracking process on CW 316L seems to be the transition from slow transgranular propagation of surface initiated cracks to intergranular fast propagation through the thickness of the sample. The duration of this stage during crack initiation tests is

  11. Characterization of interfacial reactions and oxide films on 316L stainless steel in various simulated PWR primary water environments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Junjie; Xiao, Qian [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@t.shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Shanghai Key Laboratory of Advanced Ferrometallurgy, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Ru, Xiangkun; Peng, Hao; Xiong, Qi; Li, Hongjuan [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China)

    2017-06-15

    The effect of water chemistry on the electrochemical and oxidizing behaviors of 316L SS was investigated in hydrogenated, deaerated and oxygenated PWR primary water at 310 °C. Water chemistry significantly influenced the electrochemical impedance spectroscopy parameters. The highest charge-transfer resistance and oxide-film resistance occurred in oxygenated water. The highest electric double-layer capacitance and constant phase element of the oxide film were in hydrogenated water. The oxide films formed in deaerated and hydrogenated environments were similar in composition but different in morphology. An oxide film with spinel outer particles and a compact and Cr-rich inner layer was formed in both hydrogenated and deaerated water. Larger and more loosely distributed outer oxide particles were formed in deaerated water. In oxygenated water, an oxide film with hematite outer particles and a porous and Ni-rich inner layer was formed. The reaction kinetics parameters obtained by electrochemical impedance spectroscopy measurements and oxidation film properties relating to the steady or quasi-steady state conditions in the time-period of measurements could provide fundamental information for understanding stress corrosion cracking processes and controlling parameters. - Highlights: •Long-term EIS measurements of 316L SS in simulated PWR primary water. •Highest charge-transfer resistance and oxide film resistance in oxygenated water. •Highest electric double-layer capacitance and oxide film CPE in hydrogenated water. •Similar compositions, different shapes of oxides in deaerated/hydrogenated water. •Inner layer Cr-rich in hydrogenated/deaerated water, Ni-rich in oxygenated water.

  12. Oligo-cyclic damage and behaviour of a 304 L austenitic stainless steel according to environment (vacuum, air, PWR primary water) at 300 C

    International Nuclear Information System (INIS)

    De Baglion, L.

    2011-01-01

    Nowadays, for nuclear power plants licensing or operating life extensions, various safety authorities require the consideration of the primary water environment effect on the fatigue life of Pressurized Water Reactor (PWR) components. Thus, this work focused on the study of low cycle fatigue damage kinetics and mechanisms, of a type 304L austenitic stainless steel. Several parameters effects such as temperature, strain rate or strain amplitude were investigated in air as in PWR water. Thanks to targeted in-vacuum tests, the intrinsic influence of these parameters and environments on the fatigue behaviour of the material was studied. It appears that compared with vacuum, air is already an active environment which is responsible for a strong decrease in fatigue lifetime of this steel, especially at 300 C and low strain amplitude. The PWR water coolant environment is more active than air and leads to increased damage kinetics, without any modifications of the initiation sites or propagation modes. Moreover, the decreased fatigue life in PWR water is essentially attributed to an enhancement of both initiation and micropropagation of 'short cracks'. Finally, the deleterious influence of low strain rates on the 304L austenitic stainless steel fatigue lifetime was observed in PWR water environment, in air and also in vacuum without any environmental effects. This intrinsic strain rate effect is attributed to the occurrence of the Dynamic Strain Aging phenomenon which is responsible for a change in deformation modes and for an enhancement of cracks initiation. (author)

  13. Quantitative measurement of trace amounts of dissolved oxygen in the primary and secondary systems of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Castaneda, H.B.; Neale, T.A.

    1989-01-01

    Establishing and maintaining the correct water chemistry conditions in the primary and secondary systems of pressurized water reactor (PWR) nuclear power plants is essential in order to maximize the operating life and guarantee the uninterrupted availability of the major components of each PWR unit. The exact specifications for maintaining the correct water chemistry are well established. One of the most important parameters that must be closely monitored in a modern power generation plant is the level of dissolved oxygen (DO) present in the system. Because of the high temperatures and pressures involved, even minute traces of DO---on the order of a few parts per billion (ppb)---can be detrimental to the heat transfer surfaces in steam generators, heaters, etc. The authors argue that the method of determining trace levels of DO presented here is a modification of the original method that has greatly increased the detection level obtainable with Rhodazine-D. Measurements down to less than 1 ppb (μg/Liter), with a resolution of 0.5 ppb (μ/Liter), are now easily obtainable. No calibration procedures are required and no maintenance of critical components is needed. This quantitative method is based on the instantaneous stoichiometric reaction of Rhodazine-D with oxygen. After less than one minute the oxidation reaction is complete and the fully developed color is compared with a set of stable liquid color standards. The color standards are formulated using the oxidized form of Rhodazine-D, thus providing an exact color match for the reacted sample-reagent. Supporting data are presented that confirm the relative accuracy and sensitivity of the new method, as well as results of a comparative evaluation of the method versus in-line dissolved oxygen analyzers

  14. The effect of small specimen volume on the deformation of Zircaloy-4 PWR cladding under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Oakden, M.M.; Reynolds, A.E.

    1983-01-01

    Creep rupture tests were performed in flowing steam on single rod specimens of 17x17 type PWR Zircaloy-4 cladding 460 mm long. They were tested at temperatures between 640 deg.C and 985 deg.C with internal pressures in the range 1.00-9.65 MPa (gauge) (145-1400 lb/in 2 ). The internal free volume was limited to 2.9 ml. Axially extended 'carrot' shaped deformations were produced, the range of temperatures and pressures over which these occurred was found to be similar to that observed in previous tests conducted with a much larger free volume. The main result of limiting the internal volume was that straining of the specimens was accompanied by a more rapid drop in the internal pressure than occurred previously, and which reduced the extent of the deformation compared with that seen in the earlier work. However, diametral strains in excess of 33% were observed which would result in mechanical interaction of neighbouring bulges if this occurred in a multi-rod array. (author)

  15. Modelling of Transport of Radioactive Substances in the Primary Circuit of Water Cooled Reactors

    International Nuclear Information System (INIS)

    2012-03-01

    coordinated research project (CRP) was proposed to determine the accuracy of existing computer codes and to identify how they could be improved through application of this body of work. Specifically, the CRP was expected to: - Build a database for selected pressurized water reactor (PWR) plants that would contain the design information suitable for their description within a computer code, as well as give the operating history of the plant, which would include the water chemistry data over several refuelling cycles; - Show the contamination of selected out-of-core surfaces such as circulating loops and steam generator channel heads versus operating history and compare the prediction of surface contamination versus time from modern radioactivity transport codes with actual plant data in a blind benchmarking exercise; - Determine how current codes, as well as new ones, could be improved and encourage the development of accurate new codes in Member States using the recommendations from the present work. This report uses as its basis the results of this CRP on 'Modelling of Transport of Radioactive Substances in the Primary Circuit of Water Cooled Reactors', which was conducted over the period 1996-2001 for PWR type reactors. The report also describes the significant progress demonstrated in this field in the period that followed.

  16. DRUFAN-01/MOD2, Transient Thermohydraulics of PWR Primary System LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Burwell, M J; Lerchl, G; Steinhoff, F; Wolfert, K [Gesellschaft fuer Reaktorsicherheit (GRS) mbH, Forschungsgelaende, 8046 Garching (Germany)

    1982-12-13

    1 - Description of problem or function: DRUFAN is an advanced best estimate code for simulation of the transient thermal hydraulic behaviour during PWR-blowdown with large break size. 2 - Method of solution: The code is based on the lumped parameter approach and allows flexible control volume configurations. The physical model takes into account thermodynamic nonequilibrium. Using finite difference techniques a 1-dimensional representation of the discharge flow path including geometrical influences is possible. The physical model is based on separated field equations for liquid and vapour mass and overall field equations for energy and momentum. The mass transfer rates between phases during evaporation and condensation are based on correlations for the controlled growth and shrinkage of vapour bubbles or liquid droplets, respectively. A heat conductor model based on the energy transport equation is available for simulation of structures, electrical heater rods and fuel rods. For the heat transfer between solid structures and the fluid a comprehensive package of flow regime dependent heat transfer and critical heat flux correlations can be used. Simulation of components (valve, pressurizer, accumulator, pump, steam generator) is possible with functions or models. Power generation in solid structures may be simulated by an input time function, an electrical heater model or a neutron kinetics models. As a result of the lumped parameter approach a set of ordinary differential equations is obtained from the field equations. These equations, together with those resulting from the simulation of critical discharge flow near the outlet by a finite difference method, are solved by an explicit/implicit integration method with automatic time step, order and error control. The ordinary differential equations representing heat conductors are solved by an essentially implicit integration method. 3 - Restrictions on the complexity of the problem: - Vapour or liquid phase are

  17. Fatigue analysis of HANARO primary cooling system piping

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo

    1998-05-01

    A main form of piping failure which occurring leak before break (LBB) is fatigue failure. The fatigue analysis of HANARO primary cooling system (PCS) piping was performed. The PCS piping had been designed in accordance with ASME Class 3 for service conditions. However fatigue analysis is not required in Class 3. In this study the quantitative fatigue analysis was carried out according to ASME Class 1. The highest stress points which have the largest possibility of ASME class 1. The highest stress points which have the largest possibility of the fatigue were determined from the piping stress analysis for each subsection piping. The fatigue analysis was performed for 3 highest stress points, i.e., branch connection, anchor point and butt welding joint. After calculating the peak stress intensity range the fatigue usage factors were evaluated considering operating cycles and S-N curve. The cumulative usage factors for 3 highest stress points were much less than 1. The results show that the possibility of fatigue failure for PCS piping subjected to thermal expansion and seismic loads is very small. The structural integrity of the HANARO PCS piping for fatigue failure was proved to apply the LBB. (author). 11 tabs., 6 figs

  18. Volumetric and chemical control auxiliary circuit for a PWR primary circuit

    International Nuclear Information System (INIS)

    Costes, D.

    1990-01-01

    The volumetric and chemical control circuit has an expansion tank with at least one water-steam chamber connected to the primary circuit by a sampling pipe and a reinjection pipe. The sampling pipe feeds jet pumps controlled by valves. An action on these valves and pumps regulates the volume of the water in the primary circuit. A safety pipe controlled by a flap automatically injects water from the chamber into the primary circuit in case of ruptures. The auxiliary circuit has also systems for purifying the water and controlling the boric acid and hydrogen content [fr

  19. Modification and application of water film model in COCOSYS for PWR's passive containment cooling

    International Nuclear Information System (INIS)

    Huang, Xi; Cheng, Xu

    2014-01-01

    Highlights: • Water film model in COCOSYS has been modified by considering film breakup. • Shear stress on film surface created by countercurrent flow has been considered. • Formation and development of rivulets have been taken into account. • Modified model has been applied for passive containment cooling system. • The modified water film model has optimized the simulation results. - Abstract: In this paper the physical model describing water film behaviors in German containment code system COCOSYS has been modified by taking into consideration the film breakup and subsequent phenomena as well as the effect of film interfacial shear stress created by countercurrent air flow. The modified model has extended its capability to predict particular water film behaviors including breakup at a critical film thickness based on minimum total energy criterion, the formation of rivulets according to total energy equilibrium as well as subsequent performance of rivulets according to several assumptions and observations from experiments. Furthermore, the modification considers also the change of velocity distribution on the cross section of film/rivulets due to shear stress. Based on the geometry of AP1000 and Generic Containment, simulations predicting containment pressure variation during accidents with operation of passive containment cooling system have been carried out. With the new model, considerably larger peak pressures are observed by comparing with those predicted with original water film model within a certain range of water film flow rate. Sensitivity analyses also point out that contact angle between water rivulets and steel substrate plays a significant role in the film cooling

  20. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant

    International Nuclear Information System (INIS)

    Monteiro, Iara Arraes

    1999-02-01

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  1. Reliability of an HTR-module primary circuit pressure boundary. Influences, sensitivity, and comparison with a PWR

    International Nuclear Information System (INIS)

    Staat, M.

    1995-01-01

    The reliability of the HTR-module for electricity and steam generation was analysed for normal operation, as well as accident conditions. The probabilistic fracture mechanics assessment was performed with a modification of the ZERBERUS code on the basis of widely used data. The calculated failure probabilities may thus be compared with similar investigations. The HTR-module primary circuit pressure boundary as a unit showed leak-before-break behaviour in a probabilistic sense, although a break was more probable than a leak for some of its parts.However, the findings may depend greatly on the stochastic data. Therefore a stochastic reference problem is defined and the results are compared with the Japanese round robin on a PWR section. Possible changes of failure probabilities and of the leak-before-break behaviour are discussed for different criteria for the events leading to a leak, and for modifications of the stochastic reference problem such as the inclusion of NDE. The results may be used to identify those stochastic variables which have the greatest influence on the computed failure probabilities, and to perhaps justify further work which would provide more detailed information on these probabilities. Furthermore, there is an obvious need for reduction of the non-statistical reasons for great variations of failure probabilities. (orig.)

  2. Quantitative assessment of intergranular damage due to PWR primary water exposure in structural Ni-based alloys

    International Nuclear Information System (INIS)

    Ter-Ovanessian, Benoît; Deleume, Julien; Cloué, Jean-Marc; Andrieu, Eric

    2013-01-01

    Highlights: ► IG damage occurred on Ni-base alloys during exposure at high temperature water. ► Two characterization methods yield a tomographic analysis of this IG damage. ► Connected or isolated intergranular oxygen/oxide penetrations are quantified. ► Such quantitative description provides information on IGSCC susceptibility. - Abstract: Two nickel-based alloys, alloy 718 and alloy 600, known to have different resistances to IGSCC, were exposed to a simulated PWR primary water environment at 360 °C for 1000 h. The intergranular oxidation damage was analyzed in detail using an original approach involving two characterization methods (Incremental Mechanical Polishing/Microcopy procedure and SIMS imaging) which yielded a tomographic analysis of the damage. Intergranular oxygen/oxide penetrations occurred either as connected or isolated penetrations deep under the external oxide/substrate interface as far as 10 μm for alloy 600 and only 4 μm for alloy 718. Therefore, assessing this damage precisely is essential to interpret IGSCC susceptibility.

  3. Effects of PbO on the oxide films of incoloy 800HT in simulated primary circuit of PWR

    International Nuclear Information System (INIS)

    Tan, Yu; Yang, Junhan; Wang, Wanwan; Shi, Rongxue; Liang, Kexin; Zhang, Shenghan

    2016-01-01

    Effects of trace PbO on oxide films of Incoloy 800HT were investigated in simulated primary circuit water chemistry of PWR, also with proper Co addition. The trace PbO addition in high temperature water blocked the protective spinel oxides formation of the oxide films of Incoloy 800HT. XPS results indicated that the lead, added as PbO into the high temperature water, shows not only +2 valance but also +4 and 0 valances in the oxide film of 800HT co-operated with Fe, Cr and Ni to form oxides films. Potentiodynamic polarization results indicated that as PbO concentration increased, the current densities of the less protective oxide films of Incoloy 800HT decreased in a buffer solution tested at room temperature. The capacitance results indicated that the donor densities of oxidation film of Incoloy 800HT decreased as trace PbO addition into the high temperature water. - Highlights: • Trace PbO addition into the high temperature water block the formation of spinel oxides on Incoloy 800HT. • The donor density of oxide film decreases with trace PbO addition. • The current density of potentiodynamic polarization decreases of oxide film with trace PbO addition.

  4. Effect of surface state on the oxidation behavior of welded 308L in simulated nominal primary water of PWR

    Science.gov (United States)

    Ming, Hongliang; Zhang, Zhiming; Wang, Jiazhen; Zhu, Ruolin; Ding, Jie; Wang, Jianqiu; Han, En-Hou; Ke, Wei

    2015-05-01

    The oxidation behavior of 308L weld metal (WM) with different surface state in the simulated nominal primary water of pressurized water reactor (PWR) was studied by scanning electron microscopy (SEM) equipped with energy dispersive X-ray spectroscopy (EDS), X-ray diffraction (XRD) analyzer and X-ray photoelectron spectroscopy (XPS). After 480 h immersion, a duplex oxide film composed of a Fe-rich outer layer (Fe3O4, Fe2O3 and a small amount of NiFe2O4, Ni(OH)2, Cr(OH)3 and (Ni, Fe)Cr2O4) and a Cr-rich inner layer (FeCr2O4 and NiCr2O4) can be formed on the 308L WM samples with different surface state. The surface state has no influence on the phase composition of the oxide films but obviously affects the thickness of the oxide films and the morphology of the oxides (number & size). With increasing the density of dislocations and subgrain boundaries in the cold-worked superficial layer, the thickness of the oxide film, the number and size of the oxides decrease.

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

    International Nuclear Information System (INIS)

    Cognet, C.; Gandrille, P.

    1999-01-01

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

  6. Estimation on the Pressure Loss of the Conceptual Primary Cooling System and Design of the Primary Cooling Pump for a Research Reactor

    International Nuclear Information System (INIS)

    Seo, Kyoung Woo; Oh, Jae Min; Park, Jong Hark; Chae, Hee Taek; Seo, Jae Kwang; Park, Cheon Tae; Yoon, Ju Hyeon; Lee, Doo Jeong

    2009-01-01

    A new conceptual primary cooling system (PCS) for a research reactor has been designed for an adequate cooling to the reactor core which has various powers ranging from 30MW through 80MW. The developed primary cooling system consisted of decay tanks, pumps, heat exchangers, vacuum breakers, some isolation and check valves, connection piping, and instruments. Because the system flow rate should be determined by the thermal hydraulic design analysis for the core, the heads to design the primary cooling pumps (PCPs) in a PCS will be estimated by the variable system flow rates. The heads of the part of a research reactor vessel was evaluated by the previous study. The various pressure losses of the PCS can be calculated by the dimensional analysis of the pipe flow and the head loss coefficient of the components. The purpose of this research is to estimate the various pressure losses and to design the PCPs

  7. T/sub hot/ reduction: a program for lowering primary temperatures on a PWR

    International Nuclear Information System (INIS)

    Augustine, D.B.; DiTommaso, S.M.; Manz, E.M.; Reister, P.

    1987-01-01

    This paper focuses on the key technical issues addressed in a program to support operation of the Byron Unit 1 pressurized water reactor at primary side temperatures significantly lowered with respect at primary side temperatures significantly lowered with respect to the original design temperatures. These operating temperatures were lowered in order to reduce the potential for initiation of primary water stress corrosion cracking in the steam generator tubing. The efforts of this program were aimed at maintaining operation of the unit at the maximum possible power level at the reduced temperatures. In addition, the program is designed to allow for cycle-to-cycle flexibility within a range of operating temperatures from the original design temperatures to temperatures lowered by ∼ 11 0 C (20 0 F)

  8. Study of colloidal particles behaviour in the PWR primary circuit conditions

    International Nuclear Information System (INIS)

    Barale, M.

    2006-12-01

    EDF wants to understand, model and limit primary circuit contamination of Pressurized Water Reactors by colloidal particles resulting from corrosion. The electrostatic behaviour of representative oxide particles (cobalt ferrite, nickel ferrite and magnetite) has been studied in primary circuit conditions with the influence of boric acid and lithium hydroxide. The isoelectric point (IEP) and the point of zero charge (PZC) of particles, measured between 5 C and 320 C, exhibit a minimum towards 200 C. The thermodynamic constants of the protonation equilibrium of surface sites were calculated. When boric acid is added, zeta potential and IEP decrease because of borate ions sorption. On the contrary, there is not effect of lithium ions. The modelling of these results under conditions representative of primary circuit shows that these oxides exhibit a negative surface charge, explaining their sorption and adhesion behaviour. (author)

  9. Integrated equipment for increasing and maintaining coolant pressure in primary circuit of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Sykora, D.

    1986-01-01

    An open heat pump circuit is claimed connected to the primary circuit. The pump circuit consists of a steam pressurizer with a built-in steam distributor, a compressor, an expander, a reducing valve, an auxiliary pump, and of water and steam pipes. The operation is described and a block diagram is shown of integrated equipment for increasing and maintaining pressure in the nuclear power plant primary circuit. The appropriate entropy diagram is also shown. The advantage of the open pump circuit consists in reducing the electric power input and electric power consumption for the steam pressurizers, removing entropy loss in heat transfer with high temperature gradient, in the possibility of inserting, between the expander and the auxiliary pump, a primary circuit coolant treatment station, in simplified design and manufacture of the high-pressure steam pressurizer vessel, reducing the weight of the steam pressurizer by changing its shape from cylindrical to spherical, increasing the rate of pressure growth in the primary circuit. (E.S.)

  10. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    Energy Technology Data Exchange (ETDEWEB)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2016-12-15

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

  11. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    International Nuclear Information System (INIS)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung

    2016-01-01

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask

  12. Modeling PWR systems for monitoring primary-to-secondary leakage using tritium tracer

    International Nuclear Information System (INIS)

    Peiffer, D.G.

    1992-01-01

    This paper discusses several techniques available for monitoring primary to secondary leakage, focusing on the advantages and disadvantages of each. A mathematical model of Millstone 2 describes the behavior of tritium activity in the secondary plant water when a leak exists. Real data from Millstone 2 illustrate the accuracy and reliability of the model and use of the model to measure the mass of water in the secondary system

  13. PWR type reactor equipped with a primary circuit loop water level gauge

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro.

    1990-01-01

    The time of lowering a water level to less than the position of high temperature side pipeway nozzle has been rather delayed because of the swelling of mixed water level due to heat generation of the reactor core. Further, there has been a certain restriction for the installation, maintenance and adjustment of a water level gauge since it is at a position under high radiation exposure. Then, a differential pressure type water level gauge with temperature compensation is disposed at a portion below a water level gauge of a pressurizer and between the steam generator exit plenum and the lower end of the loop seal. Further, a similar water level system is disposed to all of the loops of the primary circulation circuits. In a case that the amount of water contained in a reactor container should decreased upon occurrence of loss of coolant accidents caused by small rupture and stoppage of primary circuit pumps, lowering of the water level preceding to the lowering of the water level in the reactor core is detected to ensure the amount of water. Since they are disposed to all of the loops and ensure the excess margin, reliability for the detection of the amount of contained water can be improved by averaging time for the data of the water level and averaging the entire systems, even when there are vibrations in the fluid or pressure in the primary circuit. (N.H.)

  14. Contamination of a PWR primary circuit by fuel pins with failed cladding

    International Nuclear Information System (INIS)

    Janvier, J.C.; Chagrot, M.

    1979-01-01

    The safety authorities in the principal nuclear countries appear to be attaching increasing importance to keeping reactor primary circuits as contamination-free as possible. Therefore, the consequences of cladding failures and especially of those resulting from fabrication defects have to be evaluated, for when these failures become systematic in nature they constitute an important source of contamination in pressurized-water reactors. The Grenoble Nuclear Research Centre is implementing a programme on the study of such failures with a view to analysing the behaviour of failed fuel elements. A distinction is made between two types of cladding failure, depending on whether the primary water enters the fuel pin as soon as the circuits are pressurized (fabrication defect) or whether the failure is caused during operation. The emission of gaseous fission products and halogens has been analysed in different operating modes (steady-state or transient), and in spite of the complexity of the phenomena involved, some results have been obtained which already enable one to evaluate fission product contamination of the primary circuit. (author)

  15. Experimental fast reactor JOYO MK-III functional test. Primary auxiliary cooling system test

    International Nuclear Information System (INIS)

    Karube, Koji; Akagi, Shinji; Terano, Toshihiro; Onuki, Osamu; Ito, Hideaki; Aoki, Hiroshi; Odo, Toshihiro

    2004-03-01

    This paper describes the results of primary auxiliary cooling system, which were done as a part of JOYO MK-III function test. The aim of the tests was to confirm the operational performance of primary auxiliary EMP and the protection system including siphon breaker of primary auxiliary cooling system. The items of the tests were: (Test No.): (Test item). 1) SKS-117: EMP start up test. 2) SKS-118-1: EMP start up test when pony motor running. 3) SKS-121: Function test of siphon breaker. The results of the tests satisfied the required performance, and demonstrated successful operation of primary auxiliary cooling system. (author)

  16. Effect of strain-path on stress corrosion cracking of AISI 304L stainless steel in PWR primary environment at 360 deg. C

    International Nuclear Information System (INIS)

    Couvant, T.; Vaillant, F.; Boursier, JM.; Delafosse, D.

    2004-01-01

    Austenitic stainless steels (ASS) are widespread in primary and auxiliary circuits of PWR. Moreover, some components suffer stress corrosion cracking (SCC) under neutron irradiation. This degradation could be the result of the increase of hardness or the modification of chemical composition at the grain boundary by irradiation. In order to avoid complex and costly corrosion facilities, the effects of irradiation on the material are commonly simulated by applying a cold work on non-irradiated material prior to stress corrosion cracking tests. Slow strain rate tests were conducted on an austenitic stainless steel (SS) AISI 304L in PWR environment (360 deg. C). Particular attention was directed towards pre-straining effects on crack growth rate (CGR) and crack growth path (CGP). Results have demonstrated that the susceptibility of 304L to SCC in high-temperature hydrogenated water was enhanced by pre-straining. It seemed that IGSCC was enhanced by complex strain paths. (authors)

  17. Identification of Standing Pressure Waves Sources in Primary Loops of NPP with WWER and PWR

    Directory of Open Access Journals (Sweden)

    K.N. Proskuriakov

    2016-05-01

    Full Text Available Results of measurement and calculation of Eigen frequencies of coolant pressure oscillations in primary loops of NPP are presented. The simple calculation model based on equivalence of electric circuit with elastic wave propagation in liquids and gases, which gives a sensible interpretation of standing pressure waves sources is developed. It is shown, that pressurizer manifest itself as managed Helmholtz resonator generating a number of SPW (with Eigen frequencies of steam volume, water volume and their combination with coolant volume of respiratory line.

  18. High temperature filtration of radioactivable corrosion products in the primary circuit of PWR type reactors

    International Nuclear Information System (INIS)

    Dolle, L.

    1976-01-01

    A effective limitation to the deposition of radioactive corrosion products in the core of a reactor at power operation, is to be obtained by filtering the water of the primary circuit at a flow rate upper than 1% of the coolant flow rate. However, in view of accounting for more important release of corrosion products during the reactor start-up and also for some possible variations in the efficiency of the system, it is better that the flow rate to be treated by the cleaning circuit is stated at 5%. Filtration must be effected at the temperature of the primary circuit and preferably on each loop. To this end, the feasibility of electromagnetic filtration or filtration through a deep bed of granulated graphite has been studied. The on-loop tests effected on each filter gave efficiencies and yields respectively upper than 90% and 99% for magnetite and ferrite particles in suspension in water at 250 deg C. Such results confirm the interest lying in high temperature filtration and lead to envisage its application to reactors [fr

  19. Study of the corrosion products in the primary system of PWR plants as the source of radiation fields build-up

    International Nuclear Information System (INIS)

    Brabant, R. van; Regge, P. de.

    1982-01-01

    In the first part the behaviour of the corrosion products in the primary system of PWR plants is depicted on the basis of a literature review of the field. Water chemistry, corrosion processes and activation of corrosion products are the main topics. In the second part the results of the characterization of corrosion particles in the primary coolant circuit of the Doel 1 and 2 reactors are described, during steady state operation and transient phases. In the third part the possibilities for radiation control at nuclear power plants are outlined. The filtration possibilities for the reactor coolant are explored in detail. (author)

  20. Effect of Residual Strain on PWSCC Initiation of Alloy 690 in Primary Water of PWR

    International Nuclear Information System (INIS)

    Lee, Eun-Hee; Kim, Sung-Woo; Eom, Ki-Hyeon; Hwang, Seong-Sik

    2016-01-01

    The object of this study is to attempt to correlate the susceptibility to crack initiation of Alloy 690 with the levels of cold work in simulated primary water. Experiments were conducted in high-temperature water to accelerate the cracking behavior and to determine the severity of crack initiation as a function of level of cold work for Alloy 690. Cracking susceptibility was determined by measuring the crack length per unit area and crack density. SCC initiation susceptibility of Alloy 690 increased with increase of the residual strain induced by the prior cold-rolling process. The 20% cold-rolled specimen exhibited mostly IG cracks, while the 30% and 40% cold-rolled specimens revealed significant amounts of IG and TG cracks. Most cracks were observed near the smallest cross-section area of the tapered specimen, where the maximum stress was applied, indicating that the SCC initiation susceptibility was strongly dependent on the applied stress. In order to find the correlation between SCC initiation susceptibility and residual strain induced by cold work, more analyses need to be performed in terms of the crack length per unit area and the crack density, as a strong indicative for SCC initiation susceptibility

  1. Importance of ECP in the prediction of radiation fields in PWR and VVER primary circuits

    International Nuclear Information System (INIS)

    Urquidi-Macdonald, M.; Jacesko, S.L.; Macdonald, Digby D.; Salter-Williams, M.

    2002-01-01

    A model has been developed for predicting mass and activity transport in the primary coolant circuits of PWRs and VVERs with the objective of demonstrating and quantifying the importance of the electrochemical corrosion potential (ECP) in determining the impact of both processes on reactor operation. The model initially employs a radiolysis/mixed potential code to calculate the ECP at four locations (core, hot leg, steam generator, cold leg) and the ECP is then used to estimate the local magnetite solubility. The solubility is then averaged around the loop to yield the ''background'' solubility. Comparison of the background solubility with the local solubility determines whether precipitation or dissolution will occur at any given point in the circuit under any given set of conditions. It is further assumed that the concentration of 59 Co in the coolant is given by the isotopic fraction of this species compared with iron averaged over all materials and weighted by the respective wetted areas. Activation of 59 Co to 60 Co is assumed to occur in the coolant phase by fast, epithermal, and thermal neutron capture. The calculated activity is then used to train an artificial neural network (ANN) to establish relationships between activity at any given location and the operating properties of the reactor, including coolant pH, ECP, temperature, power level, etc. The model predicts that during shut down, magnetite (and hence 59 Co) migrates to the core, where it is irradiated and activated, particularly during subsequent start-up. During start-up, the magnetite (and hence 60 Co) migrates from the core to out-of-core surfaces where it establishes the radiation fields. (authors)

  2. PWR primary system chemistry: Experience with elevated pH at Millstone Point Unit 3

    International Nuclear Information System (INIS)

    Bergmann, C.A.; McAtee, K.R.; Roesmer, J.

    1990-08-01

    The maintenance of an elevated pH of the hot reactor coolant in several plants for a fraction of their respective fuel cycles confirmed earlier laboratory findings that an elevated pH (>6.9 at temperature) resulted in reduced component radiation fields. In view of the potential radiological benefits of elevated coolant pH operation, Northeast Utilities (NU) in support of an EPRI- Westinghouse program, agreed to operate the Millstone 3 plants as a demonstration of elevated coolant pH during operation with extended fuel cycles. This report represents the data and describes the effects of operation of the Millstone 3 cycle 2 core at an elevated hot coolant pH of 7.0--7.5 on Zircaloy cladding corrosion, and on the buildup of core crud and excore radiation fields. The following preliminary conclusion were made based on an evaluation of the data obtained after one cycle of Millstone 3 elevated coolant pH operation. Although some of the measured zirconium dioxide thicknesses are greater than expected, the data does not indicate a significant lithium effect on the corrosion rate of the fuel cladding because of the relatively high variability of thickness measurements on rods of similar duty. There was no observable effect of the higher pH operation on any of the other fuel assembly components. Changes in core crude deposit thickness, chemical and radiochemical composition and apparent residence time were noted. These changes are consistent with the expected effects of a more dissolving coolant on activity and crud transport in a primary circuit. The overall radiation exposure rate trends from cycle 1 to cycle 2 indicated a favorable effect of the higher pH operation. 17 figs., 11 tabs

  3. PWR plant transient analyses using TRAC-PF1

    International Nuclear Information System (INIS)

    Ireland, J.R.; Boyack, B.E.

    1984-01-01

    This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

  4. Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Iguchi, Tadashi; Murao, Yoshio

    1985-01-01

    In the reactor safety assessment during reflood phase of a PWR-LOCA, it is assumed implicitly that the core thermal hydraulic behavior is evaluated by the one-dimensional model with an average power rod. In order to assess the applicability of the one-dimensional treatment, integral tests were performed with various core radial power profiles using the Cylindrical Core Test Facility (CCTF) whose core includes about 2,000 heater rods. The CCTF results confirm that the core radial power profile has weak effect on the thermal hydraulic behavior in the primary system except core. It is also confirmed that the core differential pressure in the axial direction is predicted by the one-dimensional core model with an average power rod even in the case with a steep radial power profile in the core. Even though the core heat transfer coefficient is dependent on the core radial power profile, it is found that the error of the peak clad surface temperature calculation is less than 15 K using the one-dimensional model in the CCTF tests. The CCTF results support the one-dimensional treatment assumed in the reactor safety assessment. (author)

  5. Oxidation of Alloy 82 in nominal PWR primary water at 340 deg. C and in hydrogenated steam at 400 deg. C

    International Nuclear Information System (INIS)

    Chaumun, Elizabeth; Guerre Catherine; Duhamel, Cecilie; Sennour, Mohamed; Curieres, Ian-de

    2012-09-01

    Nickel-base weld metals are susceptible to stress corrosion cracking (SCC) in Pressurized Water Reactor (PWR) primary water. As tests in laboratory need to last, in some cases, at least several thousand hours to get stress corrosion crack initiation or propagation in simulated primary water, pure hydrogenated steam at 400 deg. C was used to perform accelerated tests. To confirm that these conditions are still representative of primary water conditions, results of oxidation tests of coupons in hydrogenated steam at 400 deg. C and in primary water at 340 deg. C have been compared. Surface oxide layers have been characterized in order to discuss the influence of the temperature and of the media (water or steam). (authors)

  6. The effect of primary copper slag cooling rate on the copper valorization in the flotation process

    Directory of Open Access Journals (Sweden)

    Aleksandar Mihajlović

    2015-06-01

    Full Text Available Technological procedure of slow cooling slag from primary copper production is applied in the purpose of copper recovery in the level of 98.5% to blister. This technological procedure is divided into two phases, first slow cooling of slag on the air for 24 hours, and then accelerated cooling with water for 48 hours. Within the research following methods were used: calculation of nonstationary slag cooling, verification of the calculation using computer simulation of slag cooling in the software package COMSOL Multiphysics and experimental verification of simulation results. After testing of the experimentally gained samples of slowly cooled slag it was found that this technological procedure gives the best results in promoting growth or coagulation of dispersed particles of copper sulfide and copper in the slag, thereby increasing the utilization of the flotation process with a decrease of copper losses through very fine particles.

  7. Vaporization Rate Analysis of Primary Cooling Water from Reactor PUSPATI TRIGA (RTP) Tank

    International Nuclear Information System (INIS)

    Tonny Anak Lanyau; Mohd Fazli Zakaria; Yahya Ismail

    2011-01-01

    Primary cooling system consists of pumps, heat exchangers, probes, a nitrogen-16 diffuser and associated valves is connected to the reactor TRIGA PUSPATI (RTP) tank by aluminium pipes. Both the primary cooling system and the reactor tank is filled with demineralized light water (H 2 O), which serves as a coolant, moderator as well as shielding. During reactor operation, vaporization in the reactor tank will reduce the primary water and contribute to the formation of vapor in the reactor hall. The vaporization may influence the function of the water subsequently may affect the safety of the reactor operation. It is essential to know the vaporization rate of the primary water to ensure its functionality. This paper will present the vaporization rate of the primary cooling water from the reactor tank and the influence of temperature of the water in the reactor tank to the vaporization rate. (author)

  8. Changes in water chemistry and primary productivity of a reactor cooling reservoir (Par Pond)

    International Nuclear Information System (INIS)

    Tilly, L.J.

    1975-01-01

    Water chemistry and primary productivity of a reactor cooling reservoir have been studied for 8 years. Initially the primary productivity increased sixfold, and the dissolved solids doubled. The dissolved-solids increase appears to have been caused by additions of makeup water from the Savannah River and by evaporative concentration during the cooling process. As the dissolved-solids concentrations and the conductivity of makeup water leveled off, the primary productivity stabilized. Major cation and anion concentrations generally followed total dissolved solids through the increase and plateau; however, silica concentrations declined steadily during the initial period of increased plankton productivity. Standing crops of net seston and centrifuge seston did not increase during this initial period. The collective data show the effects of thermal input to a cooling reservoir, illustrate the need for limnological studies before reactor siting, and suggest the possibility of using makeup-water additions to power reactor cooling basins as a reservoir management tool

  9. The effect of zinc injection into PWR primary coolant on the reduction of radiation buildup and corrosion control. The solubilities of zinc, nickel and cobalt spinel oxides

    International Nuclear Information System (INIS)

    Miyajima, Kaori; Hirano, Hideo

    1999-01-01

    The use of zinc injection into PWR primary coolant to reduce radiation buildup has been widely studied, and te reduction effect has been experimentally confirmed. However, some items, such as the optimal concentration of zinc required to reduce radiation buildup, the corrosion control effect of zinc injection, and the influence of zinc injection on the integrity of fuel cladding, have not been clarified yet. In particular, the corrosion suppression effect of zinc remains unconfirmed. Therefore, it is necessary to measure and calculate the solubilities of zinc and nickel spinel oxides, which are formed on the surface of Ni-based alloys in PWR primary systems. In this study, in order to assess the effectiveness of zinc injection in the reduction of radiation buildup and the corrosion control of Ni-based alloy, the potential-pH diagrams for Zn-Cr-H 2 O, Ni-Cr-H 2 O, and Co-Cr-H 2 O systems at 300degC were constructed and the solubilities of Zn-Cr, Ni-Cr, and Co-Cr spinel oxides were calculated. It is concluded that under pH conditions for which NiCr 2 O 4 is stable, zinc injection is effective in corrosion control as well as in reducing radiation buildup. (author)

  10. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao

    2016-01-01

    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  11. The Selection Method of RCM in the Primary Cooling System of RSG GA. Siwabessy Related to Functions as the Primary Cooling Reactor RSG GA. Siwabessy

    International Nuclear Information System (INIS)

    Mohammad Tahril Azis; Salman Suprawhardana, M.; Teguh Pudji Purwanto

    2010-01-01

    Reactor RSG GA. Siwabessy to ensure the temperature inside the reactor core and reflectors within the limits of allowable operations during reactor operation. The primary cooling system components must refer to the thermal power reactors and to minimize failure probability of components to operate the reactor in safe and secure. The RCM Method Development (Reliability Centered Maintenance) with a web-based Free Open Source Software (FOSS)/GPL (General Public License), will assist the maintenance support information system that can work in the intranet / internet. Free Open Source Software (FOSS) is software that can provide assurance to the user to perform the development, sharing and make changes if necessary, especially users feel confident that the software actually legal and free (free software). The RCM method recommends maintenance types of 52 task selection to be applied in the primary cooling system with details time directed (td) 35% (18 tasks), condition directed (cd) 63% (33 tasks) and 1% failure finding (1 task). (author)

  12. Modelling the radiolysis of RSG-GAS primary cooling water

    Science.gov (United States)

    Butarbutar, S. L.; Kusumastuti, R.; Subekti, M.; Sunaryo, G. R.

    2018-02-01

    Water chemistry control for light water coolant reactor required a reliable understanding of radiolysis effect in mitigating corrosion and degradation of reactor structure material. It is known that oxidator products can promote the corrosion, cracking and hydrogen pickup both in the core and in the associated piping components of the reactor. The objective of this work is to provide the radiolysis model of RSG GAS cooling water and further more to predict the oxidator concentration which can lead to corrosion of reactor material. Direct observations or measurements of the chemistry in and around the high-flux core region of a nuclear reactor are difficult due to the extreme conditions of high temperature, pressure, and mixed radiation fields. For this reason, chemical models and computer simulations of the radiolysis of water under these conditions are an important route of investigation. FACSIMILE were used to calculate the concentration of O2 formed at relatively long-time by the pure water γ and neutron irradiation (pH=7) at temperature between 25 and 50 °C. This simulation method is based on a complex chemical reaction kinetic. In this present work, 300 MeV-proton were used to mimic γ-rays radiolysis and 2 MeV fast neutrons. Concentration of O2 were calculated at 10-6 - 106 s time scale.

  13. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

  14. Water chemistry in PWR

    International Nuclear Information System (INIS)

    Abe, Kenji

    1987-01-01

    This article outlines major features and basic concept of the secondary system of PWR's and water properties control measures adopted in recent PWR plants. The secondary system of a PWR consists of a condenser cooling pipe (aluminum-brass, titanium, or stainless steel), low-pressure make-up water heating pipe (aluminum-brass or stainless steel), high-ressure make-up water heating pipe (cupro-nickel or stainless steel), steam generator heat-transfer pipe (Inconel 600 or 690), and bleed/drain pipe (carbon steel, low alloy steel or stainless steel). Other major pipes and equipment are made of carbon steel or stainless steel. Major troubles likely to be caused by water in the secondary system include reduction in wall thickness of the heat-transfer pipe, stress corrosion cracking in the heat-transfer pipe, and denting. All of these are caused by local corrosion due to concentration of purities contained in water. For controlling the water properties in the secondary system, it is necessary to prevent impurities from entering the system, to remove impurities and corrosion products from the system, and to prevent corrosion of apparatus making up the system. Measures widely adopted for controlling the formation of IGA include the addition of boric acid for decreasing the concentration of free alkali and high hydrazine operation for providing a highly reducing atmospere. (Nogami, K.)

  15. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  16. Decontamination before dismantling a fast breeder reactor primary cooling system

    International Nuclear Information System (INIS)

    Costes, J.R.; Antoine, P.; Gauchon, J.P.

    1997-01-01

    The large-scale decontamination of FBR sodium loops is a novel task, as only a limited number of laboratory-scale results are available to date. The principal objective of this work is to develop a suitable decontamination procedure for application to the primary loops of the RAPSODIE fast breeder reactor as part of decommissioning to Stage 2. After disconnecting the piping from the main vessel, the pipes were treated by circulating chemical solutions and the vessels by spraying. The dose rate in the areas to be dismantled was divided by ten. A decontamination factor of about 300 was obtained, and should allow austenitic steel parts to be melted in special furnaces for unrestricted release. (author)

  17. The Effects of Hot Bending on the Low Cycle Fatigue Behaviors of 347 SS in PWR Primary Environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho-Sub; Hong, Jong-Dae; Lee, Junho; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    Fatigue damage could be significant for some locations, especially the welds and bends where stress concentration is typically high. As a possible solution, a large radius hot-bending method has been suggested to eliminate some weld joints and all tight bends. However, for the hot-bending process which involves a high temperature thermal cycle, there is a concern about changes in mechanical properties including low cycle fatigue behaviors. In APR1400, Type 347 SS have been used as surge line pipes. Therefore, to verify the applicability of hot-bending on 347 SS surge line pipes, an environmental fatigue test program was initiated. In this paper, the preliminary results of the on-going test program are introduced. Also, the low cycle fatigue behaviors of 347 SS are compared with those of other grade of stainless steels. The effects of hot bending on the low cycle fatigue behavior of 347 SS were quantitatively evaluated. The fatigue life was compared with the estimated values per NUREG 6909 rev. 1. There are no distinct differences between NUREG 6909 and LCF tests. According to fractography and cross section analysis in progress, basically, the reduction of LCF life of 347 SS in PWR water was caused by operation of HIC mechanism. The cyclic stress responses shows that there is no secondary hardening in 330 .deg.C air and PWR water.

  18. A scaling study of the natural circulation flow of the ex-vessel core catcher cooling system of a 1400MW PWR for designing a scale-down test facility

    International Nuclear Information System (INIS)

    Rhee, Bo. W.; Ha, K. S.; Park, R. J.; Song, J. H.

    2012-01-01

    A scaling study on the steady state natural circulation flow along the flow path of the ex-vessel core catcher cooling system of 1400MWe PWR is described. The scaling criteria for reproducing the same thermalhydraulic characteristics of the natural circulation flow as the prototype core catcher cooling system in the scale-down test facility is derived and the resulting natural circulation flow characteristics of the prototype and scale-down facility analyzed and compared. The purpose of this study is to apply the similarity law to the prototype EU-APR1400 core catcher cooling system and the model test facility of this prototype system and derive a relationship between the heating channel characteristics and the down-comer piping characteristics so as to determine the down-comer pipe size and the orifice size of the model test facility. As the geometry and the heating wall heat flux of the heating channel of the model test facility will be the same as those of the prototype core catcher cooling system except the width of the heating channel is reduced, the axial distribution of the coolant quality (or void fraction) is expected to resemble each other between the prototype and model facility. Thus using this fact, the down-comer piping design characteristics of the model facility can be determined from the relationship derived from the similarity law

  19. Core cooling systems

    International Nuclear Information System (INIS)

    Hoeppner, G.

    1980-01-01

    The reactor cooling system transports the heat liberated in the reactor core to the component - heat exchanger, steam generator or turbine - where the energy is removed. This basic task can be performed with a variety of coolants circulating in appropriately designed cooling systems. The choice of any one system is governed by principles of economics and natural policies, the design is determined by the laws of nuclear physics, thermal-hydraulics and by the requirement of reliability and public safety. PWR- and BWR- reactors today generate the bulk of nuclear energy. Their primary cooling systems are discussed under the following aspects: 1. General design, nuclear physics constraints, energy transfer, hydraulics, thermodynamics. 2. Design and performance under conditions of steady state and mild transients; control systems. 3. Design and performance under conditions of severe transients and loss of coolant accidents; safety systems. (orig./RW)

  20. Transport of radioactive corrosion products in primary system of sodium-cooled fast breeder reactor 'MONJU'

    International Nuclear Information System (INIS)

    Matuo, Youichirou; Hasegawa, Masanori; Maegawa, Yoshiharu; Miyahara, Shinya

    2011-01-01

    Radioactive corrosion products (CP) are primary cause of personal radiation exposure during maintenance work at FBR plants with no breached fuel. The PSYCHE code has been developed based on the Solution-Precipitation model for analysis of CP transfer behavior. We predicted and analyzed the CP solution and precipitation behavior of MONJU to evaluate the applicability of the PSYCHE code to MONJU, using the parameters verified in the calculations for JOYO. From the calculation result pertaining to the MONJU system, distribution of 54 Mn deposited in the primary cooling system over 20 years of operation is predicted to be approximately 7 times larger than that of 60 Co. In particular, predictions show a notable tendency for 54 Mn precipitation to be distributed in the primary pump and cold-leg. The calculated distribution of 54 Mn and 60 Co in the primary cooling system of MONJU agreed with tendencies of measured distribution of JOYO. (author)

  1. Vibration analysis of cooling system of upgraded PARR-1: (primary pumps)

    International Nuclear Information System (INIS)

    Ayazuddin, S.K.; Baig, R.; Pervez, S.

    1992-12-01

    During the conversion and up gradation of PARR-1, major changes were made in the cooling system of the reactor with the addition of new heat exchanger assemblies and cooling tower. It was therefore, planned to perform vibration analysis on the cooling system to check proper installation and investigate any abnormality in the operation. As a first step, vibration measurements was made on the primary pumps PW-P1 and PW-P2. Power spectral density (PSD) or frequency spectrum of the signal produced from an accelerometer placed on the pump motor assembly was analysed to identify faults which are commonly found in rotating and reciprocating machinery such as unbalance, shaft misalignment and bearing instability. The root mean square (RMS) of the signal was compared with the vibration criterion chart to determine the operating condition of the pump motor assembly. The procedure used for the analysis and faults detected in the primary pump-motor system are discussed. 9 figs. (author)

  2. Effects of cold working ratio and stress intensity factor on intergranular stress corrosion cracking susceptibility of non-sensitized austenitic stainless steels in simulated BWR and PWR primary water

    International Nuclear Information System (INIS)

    Yaguchi, Seiji; Yonezawa, Toshio

    2012-01-01

    To evaluate the effects of cold working ratio, stress intensity factor and water chemistry on an IGSCC susceptibility of non-sensitized austenitic stainless steel, constant displacement DCB specimens were applied to SCC tests in simulated BWR and PWR primary water for the three types of austenitic stainless steels, Types 316L, 347 and 321. IGSCC was observed on the test specimens in simulated BWR and PWR primary water. The observed IGSCC was categorized into the following two types. The one is that the IGSCC observed on the same plane of the pre-fatigue crack plane, and the other is that the IGSCC observed on a plane perpendicular to the pre-fatigue crack plane. The later IGSCC fractured plane is parallel to the rolling plane of a cold rolled material. Two types of IGSCC fractured planes were changed according to the combination of the testing conditions (cold working ratio, stress intensity factor and simulated water). It seems to suggest that the most susceptible plane due to fabrication process of materials might play a significant role of IGSCC for non-sensitized cold worked austenitic stainless steels, especially, in simulated PWR primary water. Based upon evaluating on the reference crack growth rate (R-CGR) of the test specimens, the R-CGR seems to be mainly affected by cold working ratio. In case of simulated PWR primary water, it seems that the effect of metallurgical aspects dominates IGSCC susceptibility. (author)

  3. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  4. Stress corrosion cracking of austenitic stainless steels in PWR primary water: an update of metallurgical investigations performed on French withdrawn components

    International Nuclear Information System (INIS)

    Boursier, J.M.; Gallet, S.; Rouillon, Y.; Bordes, P.

    2002-01-01

    Austenitic stainless steels (AISI 304, 304L, 316 and 316L) are largely used in Nuclear Power Plants because of their good resistance to corrosion and their satisfactory mechanical properties. Nevertheless, on various French PWR Nuclear Power Plants, several cases of corrosion have been encountered in auxiliary circuit portions where deleterious species and oxygen can be present. This paper focuses on the metallurgical investigations performed on pulled out components such as Canopy welds or 'dead legs' (auxiliary circuit portions connected to the main primary loops) in terms of cracking locations and degradation parameters. In addition, some comparisons between Nuclear Power Plant feedback and fundamental research and development studies are discussed, particularly in the scope of temperature, microstructure, stresses (applied and residual) and medium responsible for the degradation. (authors)

  5. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    International Nuclear Information System (INIS)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav

    2017-01-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  6. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Numerical simulation of the accurate RCP start-up flow rate

    International Nuclear Information System (INIS)

    Martin, A.; Alvarez, D.; Cases, F.; Stelletta, S.

    1997-06-01

    This report explains the last results about the mixing in the 900 MW PWR vessels. The accurate fluid flow transient, induced by the RCP starting-up, is represented. In a first time, we present the Thermalhydraulic Finite Element Code N3S used for the 3D numerical computations. After that, results obtained for one reactor operation case are given. This case is dealing with the transient mixing of a clear plug in the vessel when one primary pump starts-up. A comparison made between two injection modes; a steady state fluid flow conditions or the accurate RCP transient fluid flow conditions. The results giving the local minimum of concentration and the time response of the mean concentration at the core inlet are compared. The results show the real importance of the unsteadiness characteristics of the fluid flow transport of the clear water plug. (author)

  7. The use of Zeolite into the controlling of Lithium concentration in the PWR primary water coolant (I) : the influences of Ca, Mg and Boric Acid concentration into the exchanges capacity of Ammonium Zeolite

    International Nuclear Information System (INIS)

    Sumijanto; Siti-Amini

    1996-01-01

    In this first part of research, the influences of calsium, magnesium and boric acid concentrations to the zeolite uptake of lithium in the PWR primary water coolant have been studied. The ammonium form of zeolite was found by modification of the natural zeolite which was originated from Bayah. The results showed that the boric acid concentration in the normal condition of PWR operation absolutely did not affects the lithium uptake. The Li uptake efficiency was influenced by the presence of Ca and Mg ions in order to the presence of cations competition which was dominated by Ca ion

  8. Retrofitting the instrumentation and control system of primary cooling circuit from TRIGA INR 14 MW reactor

    International Nuclear Information System (INIS)

    Preda, M.; Ciocanescu, M.; Ana, E. M.; Cristea, D.

    2008-01-01

    Activities of retrofitting the instrumentation and control system from TRIGA INR primary cooling circuit consists in replacement of actual system for: - parameter measurement; - safety; - reactor external scramming; - protection, command and supply for electrical elements of the system. This retrofitting project is designed to ensure the necessary features of reactor external safety and for technological parameter measurement. The new safety system of main cooling circuit is completely separated from its operating system and is arranged in a panel assembly in reactor control room. The operating system has the following features: - data acquisition; - parameter value and state of command elements displaying; - command elements on hierarchical levels; - operator information through visual and acoustic alarm. (authors)

  9. ROX PWR

    International Nuclear Information System (INIS)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H.

    1999-01-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO 2 , ThO 2 and Er 2 O 3 , and a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the additives UO 2 + Er 2 O 3 are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO 2 heterogeneous core, further reduction of power peaking seems necessary. (author)

  10. Performance study of ejector cooling cycle at critical mode under superheated primary flow

    International Nuclear Information System (INIS)

    Tashtoush, Bourhan; Alshare, Aiman; Al-Rifai, Saja

    2015-01-01

    Highlights: • The ECC is modeled using EES Software and it is validated with published data. • Detailed analysis of the ECC with different refrigerants is conducted. • The constant pressure mixing is better than constant area mixing ejectors. • R134a is the selected refrigerant for the best cooling cycle performance. • The superheated primary flow at critical mode is achieved with EJ2 ejector used. - Abstract: In this work the performance of the ejector cooling cycle is investigated at critical mode, where, the effects of ejector geometry, refrigerant type, and operating condition are studied. The ejector cooling cycle is modeled with EES Software. The mass, momentum, and energy conservation principles are applied to the secondary and primary flows to investigate the performance of the ejector cooling cycle under superheated primary flow. The refrigerant R134 a is selected based on the merit of its environmental and performance characteristics. The primary working fluid in the refrigeration cycle is maintained at superheated conditions for optimal ejector performance. The solar generator temperature ranges are 80–100 °C. The operating temperature of evaporator range is 8–12 °C and the optimal condensation temperature is in the range of 28–40 °C. It is found that constant-pressure mixing ejector generates higher backpressure than constant-area mixing ejector for the same entrainment ratio and COP. The type of ejector is selected based on the performance criteria of the critical backpressure and choking condition of the primary flow, the so called EJ2 type ejector meets the criteria. The COP is found to be in the range of 0.59–0.67 at condenser backpressure of 24 bar due to higher critical condenser pressure and higher generator temperature

  11. Configuration analysis of pipe support for primary cooling using Ps + Caepipe code

    International Nuclear Information System (INIS)

    Sitandung, Y. B.; Pustandyo, W.; Sujalmo, S.

    1998-01-01

    Pipe stress evaluation and support loads has been analyzed on piping segment of RSG-GAS primary cooling system. This paper describes an analysis method of piping system with the use of computer Code PS + CAEPIPE Version 3.4.05.W. From the selected pipe segment, the data of pipe characteristic, material properties, operation condition, equipment and supports were used input. The final evaluation result of primary cooling pipe segment show that actual stress dead weight and seismic load are less than allowable limits (stress ratio 0.101 for deadweight 0.35 for seismic load). From the above ratio, it can be concluded that ratio of pipe support configuration to stress distribution is acceptable, and based on analysis result, the Code used by INTERATOM was sufficiently accurate

  12. HAZOP-study on heavy water research reactor primary cooling system

    International Nuclear Information System (INIS)

    Hashemi-Tilehnoee, M.; Pazirandeh, A.; Tashakor, S.

    2010-01-01

    By knowledge-based Hazard and Operability (HAZOP) technique, equipment malfunction and deficiencies in the primary cooling system of the generic heavy water research reactor are studied. This technique is used to identify the representative accident scenarios. The related Process Flow Drawing (PFD) is prepared as our study database for this plant. Since this facility is in the design stage, applying the results of HAZOP-study to PFD improves the safety of the plant.

  13. Sizewell: proposed site for Britain's first PWR power station

    International Nuclear Information System (INIS)

    1980-10-01

    The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

  14. ROX PWR

    Energy Technology Data Exchange (ETDEWEB)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO{sub 2}, ThO{sub 2} and Er{sub 2}O{sub 3}, and a heterogeneous core with Zr-ROX and UO{sub 2} assemblies. As a result, the additives UO{sub 2}+ Er{sub 2}O{sub 3} are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO{sub 2} heterogeneous core, further reduction of power peaking seems necessary. (author)

  15. Heat Transfer Modes and their Coefficients for a Passive Containment Cooling System of PWR using a Multi-Pod Heat Pipe

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Gyeongho; Park, Junseok; Kim, Sangnyung [Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-05-15

    If a reactor core is damaged due to a disaster such as happened at TEPCO's Fukushima nuclear power plant, the inevitable rise of super-heated steam that could potentially convert to hydrogen resulting from unimpeded temperature and pressure rises will threaten the integrity of the containment structure. To prevent this, safety and regulatory standards typically specify that the gas vent and external cooling systems be designed to maintain containment up to the level C limit for 24 hours and integrity for 48 hours after any damage to the core. Furthermore, it is recommended that the installation of the exhaust penetration unit have a minimum diameter of 3ft. However, installation of such cooling measures or penetration units is burdensome in terms of operational and maintenance costs not to mention the need to ensure a fleet of fire trucks to be on standby as well as the need to ensure a plentiful supply of water for cooling and a filtration system to clean the water. Therefore, the development of a reliable passive cooling system will be economically advantageous because the extra cost burdens of the external system can be omitted. The Passive Containment Cooling System (PCCS) using a multi-pod heat pipe proposed in this study satisfies these conditions.

  16. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  17. MK-III function tests in JOYO. Primary main cooling pump

    International Nuclear Information System (INIS)

    Isozaki, Kazunori; Saito, Takakazu; Sumino, Kouzo; Karube, Kouji; Terano, Toshihiro; Sakaba, Hideo; Nakai, Satoru

    2004-06-01

    MK-III function test (SKS-1) that was carried out from October 17, 2001 through October 23, 2001 using MK-III transition core configuration and MK-III function tests (SKS-2) was carried out from January 27, 2003 through February 13, 2003 using MK-III core configuration. The major function tests results of primary cooling system were shown as follows; (1) The stability of the primary main pump flow control system was confirmed on both CAS (cascade) mode and Man (manual) mode. Also no divergence of flow and revolution of the pump were observed at step flow change disturbance. (2) The main motor was shifted to run-back flow control operation in about 54 seconds after scram. The flow rate and pump revolution at run-back operation of A and B cooling system were 167 m 3 /h and 117 rpm, 185m 3 /h and 118 rpm respectively. The pump revolution was within the design target revolution 122 rpm ± 8 rpm and the flow was over the 10% of the rated flow. (3) The pony motor was engaged in operation in about 39 seconds after the primary main pump trip. The flow rate and pump revolution at the pony motor operation of A and B cooling system were 180 m 3 /h and 124 rpm, 190 m 3 /h and 123 rpm respectively. These values were satisfied the design low limit of 93 rpm and 10% of the rated flow. (4) Free flow coast down time constant was longer than 10 seconds that was design shortest time at both the primary pump trip and run-back operation. (5) Pump over flow column sodium levels of both A and B cooling system at rated operating condition were NL-1550 mm and, NL-1468 mm respectively and were lower than NL-1581 mm of the design value. This result shows the new IHX pressure loss estimation was conservative. (6) It was confirmed that the primary main pump could operate with out scram for up to 0.6 seconds of external power supply loss. (author)

  18. Integral type small PWR with stand-alone safety

    International Nuclear Information System (INIS)

    Makihara, Yoshiaki

    2001-01-01

    A feasibility study is achieved on an integral type small PWR with stand-alone safety. It is designed to have the following features. (1) The coolant does not leak out at any accidental condition. (2) The fuel failure does never occur while it is supposed on the large scale PWR at the design base accident. (3) At any accidental condition the safety is secured without any support from the outside (stand-alone safety secure). (4) It has self-regulating characteristics and easy controllability. The above features can be satisfied by integrate the steam generator and CRDM in the reactor vessel while the pipe line break has to be considered on the conventional PWR. Several counter measures are planned to satisfy the above features. The economy feature is also attained by several simplifications such as (1) elimination of main coolant piping and pressurizer by the integration of primary cooling system and self-pressurizing, (2) elimination of RCP by application of natural circulating system, (3) elimination of ECCS and accumulator by application of static safety system, (4) large scale volume reduction of the container vessel by application of integrated primary cooling system, (5) elimination of boric acid treatment by deletion of chemical shim. The long operation period such as 10 years can be attained by the application of Gd fuel in one batch refueling. The construction period can be shortened by the standardizing the design and the introduction of modular component system. Furthermore the applicability of the reduced modulation core is also considered. (K. Tsuchihashi)

  19. PWR systems transient analysis

    International Nuclear Information System (INIS)

    Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

    1985-01-01

    Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

  20. Emergency core cooling system

    International Nuclear Information System (INIS)

    Kato, Ken.

    1989-01-01

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

  1. Modeling the cool down of the primary heat transport system using shut down cooling system in normal operation and after events such as LOCA

    International Nuclear Information System (INIS)

    Icleanu, D.L.; Prisecaru, I.

    2015-01-01

    This paper aims at modeling the cooling of the primary heat transport system using shutdown cooling system (SDCS), for a CANDU 6 NPP in all operating modes, normal and abnormal (particularly in case of LOCA accident), using the Flowmaster calculation code. The modelling of heavy water flow through the shutdown cooling system and primary heat transport system was performed to determine the distribution of flows, pressure in various areas of the hydraulic circuit and the pressure loss corresponding to the components but also for the heat calculation of the heat exchangers related to the system. The results of the thermo-hydraulic analysis show that in all cases analyzed, normal operation and for LOCA accident regime, the performance requirements are confirmed by analysis

  2. Preliminary study of radionuclide corrosion products in primary cooling water at RSG-GAS

    International Nuclear Information System (INIS)

    Lestari, D.E.; Pudjojanto, M.S.; Subiharto; Budi, S.

    1998-01-01

    Analysis of radionuclides emitting gamma rays at the primary cooling water at RSG-GAS has been carried out. The water coolant samples was performed using a low level background gamma spectrometer unit, including of high resolution of gamma detector HP-Ge Tennelec and Multichannel Analyzer (MCA) ADCAM 100 ORTEC. The result indicated Na-24 and Mn-56 radionuclides that may be as corrosion product and should studied deeply in the future. The expected activity concentration radionuclide for Mn-56 is lower than those written in the Safety Analysis Report (SAR), while for Na-24 is in agreement

  3. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  4. Safety considerations of PWR's

    International Nuclear Information System (INIS)

    Arnold, W.H. Jr.

    1977-01-01

    The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

  5. Refurbishment of the Primary Cooling System of the Puspati Triga Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramli, S.; Zakaria, M. F.; Masood, Z. [Malaysian Nuclear Agency, Kajang (Malaysia)

    2014-08-15

    The refurbishment of the 27 year old primary cooling system of the 1 MW PUSPATI TRIGA reactor was completed in April 2010 over an eight month outage. The project was implemented with the dual objective of meeting current user needs as well as a future reactor core power upgrade. Hence the cooling system was partly modernized to cater for a 3 MW{sub th} reactor by installing higher capacity heat exchangers and pumps while maintaining the piping and valve sizes. The old 1 MW tube and shell heat exchanger, which had lost 25% of its heat exchange capacity, was replaced with two 1.5 MW plate type heat exchangers. Several manually operated valves were replaced with motorized units to allow remote operation from the control room. The installed cooling system was flushed with distilled water and then subjected to hydrostatic pressure tests. In the cold run test, the system was operated for an hour for every pump and heat exchanger combination while all operating parameters were checked. In the hot run test, the same was done at four levels of increasing reactor power, and dose measurements were also recorded. The paper gives the design, installation, testing and commissioning details of the project. (author)

  6. Cooling methods of station blackout scenario for LWR plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

  7. Stress corrosion cracking growth rate of TT alloy 690 and its weld joint in simulated PWR primary water

    International Nuclear Information System (INIS)

    Yonezawa, T.

    2015-01-01

    Recently, some researchers reported that the SCC growth rate (SCCGR) of cold worked thermally treated (TT) Alloy 690 was significantly different in heat by heat. But, author has hypothesized that these high SCCGRs in cold worked TT Alloy 690 could be due to the metallurgical characteristics of these heats. In order to confirm this hypothesis, this study has been started in the author's laboratory, and the following 4 new evidences were obtained. First, microcracks of carbides and voids were observed in eutectic M 23 C 6 GB carbides (primary carbides) for cold rolled laboratory heat after as cast or lightly forged condition or for chemical composition simulated Bettis'TT Alloy 690 heat, after cold rolling, before SCC test. However, microcracks in primary carbides along grain boundaries and voids were rarely detected in the cold rolled commercial heat of TT Alloy 690 used for CRDM penetrations. Secondly, the SCCGR observed in TT Alloy 690 was different in each hot working process and each heat. Comparing the SCCGRs for all heats of cold worked TT Alloy 690, the SCCGR decreased with increasing of Vickers hardness. However, in same heats of cold worked TT Alloy 690, the SCCGR increased with increasing of Vickers hardness. Thirdly, the SCCGR in cold rolled TT Alloy 690 should be integrated by the effect of hardness or cold working ratio and by the effect of existing ratio of primary M23C6 carbides with cracks and Voids due to chemical composition and the fabrication process of TT Alloy 690. Fourthly, it is argued that the high SCCGRs in highly cold rolled TT Alloy 690 are not representative of the practical situation with TT Alloy 690 in service for CRDM adapter nozzles etc. The high SCCGR of highly cold rolled TT Alloy 690 is not thought to be an accurate tool in predicting the possibility of cracking of TT Alloy 690 for CRDM adapter nozzles. (author)

  8. Stress corrosion cracking of iron-nickel-chromium alloys in primary circuit environment of PWR-type reactors

    International Nuclear Information System (INIS)

    Boursier, Jean-Marie

    1993-01-01

    Stress corrosion cracking of Alloy 600 steam generator tubing is a great concern for pressurized water reactors. The mechanism that controls intergranular stress corrosion cracking of Alloy 600 in primary water (lithiated-borated water) has yet to be clearly identified. A study of stress corrosion cracking behaviour, which can identify the main parameters that control the cracking phenomenon, was so necessary to understand the stress corrosion cracking process. Constant extension rate tests, and constant load tests have evidenced that Alloy 600 stress corrosion cracking involves firstly an initiation period, then a slow propagation stage with crack less than 50 to 80 micrometers, and finally a rapid propagation stage leading to failure. The influence of mechanical parameters have shown the next points: - superficial strain hardening and cold work have a strong effect of stress corrosion cracking resistance (decrease of initiation time and increase of crack growth rate), - strain rate was the most suitable parameter for describing the different stage of propagation. The creep behaviour of alloy 600 has shown an increase of creep rate in primary water compared to air, which implies a local interaction plasticity/corrosion. An assessment of the durations of the initiation and the propagation stages was attempted for the whole uniaxial tensile tests, using the macroscopic strain rate: - the initiation time is less than 100 hours and seems to be an electrochemical process, - the durations of the propagation stage are strongly dependent on the strain rate. The behaviour in high primary water temperature of Alloys 690 and 800, which replace Alloy 600, was studied to appraise their margin, and validate their choice. Then the last chapter has to objective to evaluate the crack tip strain rate, in order to better describe the evolution of the different stages of cracking. (author) [fr

  9. Design and test of ASME strainer for primary cooling system in HANARO

    International Nuclear Information System (INIS)

    Park, Yong-Chul; Ryu, Jeong-Soo

    1999-01-01

    The ASME strainers have been newly installed at the suction side of each reactor coolant pump to get rid of the foreign materials which may damage the pump impeller or interfere with the coolant path of fuel flow tube or primary plate type heat exchanger. The strainer was designed in accordance with ASME SEC. III, DIV. 1, ND and the structural integrity was verified by seismic analysis. The screen was designed in accordance with the effective void area from the result of flow analysis for T-type strainer. After installation of the strainer, it was confirmed through the field test that the flow characteristics of primary cooling system were not adversely affected. The pressure loss coefficient was calculated by Darcy equation using the pressure difference through each strainer and the flow rate measured during the strainer performance test. And these are useful data to predict flow variations by the pressure difference. (author)

  10. Assessment of heat loss for RSG-GAS primary cooling system

    International Nuclear Information System (INIS)

    Dibyo, S.

    1998-01-01

    Heat Loss is part term of energy balance equation of system, therefore heat loss very important thing in the thermal dynamic analysis. Heat energy loosed from the surface pipe to the air in the room was calculated. Heat energy pass through by conduction, convection and radiation. The convection process are caused by moving of air density, i.e up flow of the hot air return to be down flow. The heat transfer phenomenon could be determined by empirical correlation of Heilman. The primary cooling system is consisted to the 3 zone : 1). Zone of (safety valves-heat exchanger), 2). Zone of heat exchanger surfaces, 3). Zone of heat exchanger-reactor pool. By using input data of air temperature are about 25 o C, temperature of primary coolant about 45 o C, The heat Loss along the pipes to the air are 23.9 k watt or 0.1%

  11. Four decades of working experience of Cirus primary cooling water heat exchangers

    International Nuclear Information System (INIS)

    Dubey, P.K.; Ullas, O.P.; Rao, D.V.H.; Zope, A.K.; Kharpate, A.V.

    2006-01-01

    CIRUS is a 40 MW (Th.) research reactor, commissioned in the year 1960. The reactor has natural uranium fuel rods, heavy water as moderator, demineralised water (DM water) as primary coolant, and seawater as secondary coolant. There are six Heat Exchangers in the primary cooling water (PCW) system. Five of them are required for the normal operation of the reactor and one is kept stand by. DM water flows on the shell side of the heat exchanger in two passes. Seawater is used as coolant on the tube side of the heat exchangers in four passes. Cirus has been in operation for around 41 years excluding refurbishment period. During these four decades of reactor operation, PCW heat exchangers have experienced many failures and undergone many modifications in the circuit for ensuring better performance. This paper tries to capture the essence of working experiences with PCW heat exchangers, various problems faced, remedial measures taken during those four decades of reactor operation. (author)

  12. Measurement and calculation of radiation sources in the primary cooling system of JOYO

    International Nuclear Information System (INIS)

    Suzuki, S.; Iizawa, K.; Ohtani, N.; Kobayashi, T.; Horie, J.; Handa, H.

    1987-01-01

    Production and transfer of radiation sources in the primary cooling system are important consideration in the LMFBR plant from the viewpoint of radiation protection and shielding design. These items were evaluated with calculations and/or measurements in the Japanese experimental fast reactor JOYO. In this study, calculations were made with the DOT3.5 0 two-dimensional discrete ordinate transport code to determine the neutron flux and production rate distributions of radiation sources in the reactor vessel. Using the DOT results, the behavior in primary coolant sodium of the CP (radioactive corrosion products) which were released from the reactor structural material was also calculationally analyzed with the PSYCHE code developed by PNC. These analytical results were compared with the measured results to get the verification of analysis methods and to estimate the accuracy of calculations

  13. Vibration monitoring of the primary piping systems during the hot functional tests of the Mulheim-Karlich PWR

    International Nuclear Information System (INIS)

    Bauernfeind, V.; Bloem, T.; Pache, W.; Diederich, H.J.

    1989-01-01

    During the hot functional tests of the Muelheim--Kaerlich first-of-a-kind plant, vibration measurements were made on the reactor pressure vessel and its' internals and on the primary piping system and main coolant pumps. This paper contains results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement program is to confirm that the components, which are of new design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. Good agreement was found. In the course of these comparisons, information on the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained

  14. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 3: nonseismic stress analysis. Final report

    International Nuclear Information System (INIS)

    Chan, A.L.; Curtis, D.J.; Rybicki, E.F.; Lu, S.C.

    1981-08-01

    This volume describes the analyses used to evaluate stresses due to loads other than seismic excitations in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station. The results of the analyses are used as input to a simulation procedure for predicting the probability of pipe fracture in the primary coolant system. Sources of stresses considered in the analyses are pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, and mechanical vibrations. Pressure and thermal transients arising from plant operations are best estimates and are based on actual plant operation records supplemented by specified plant design conditions. Stresses due to dead weight and thermal expansion are computed from a three-dimensional finite element model that uses a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, reactor pressure vessel, reactor coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients are obtained by closed-form solutions. Calculations of residual stresses account for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation are estimated by a dynamic analysis using existing measurements of pump vibrations

  15. Liquid metal cooled nuclear power plant with direct heat transfer from the primary coolant to the working medium

    International Nuclear Information System (INIS)

    Hahn, G.

    1974-01-01

    The cooling systems of the sodium-cooled reactor are entirely inside a containment. The heat transfer from the primary to the secondary coolant - i.e. water - is done in heat exchangers with three-layer tubes. As there is no component cooling heat exchanger, it is advantageous that the layers that are in touch with the primary coolant form part of the wall of the containment. An emergency cooling system inside the containment is also made of three-layer tubes. The tubes of the primary loops have the shape of loops, helices, and spirals surrounding the reactor tank or a biological shield. Between the tubes and the safety wall there are maintenance areas which are accessible from the outside. The three-layer construction prevents a reaction of leaked-out or evaporated sodium with the secondary coolant. (DG) [de

  16. Vibration monitoring of the primary piping system during the hot functional tests of the Muelheim-Kaerlich PWR

    International Nuclear Information System (INIS)

    Bauernfeind, V.; Bloem, T.; Pache, W.; Diederich, H.J.

    1992-01-01

    During the hot functional tests of the Muelheim-Kaerlich plant, which was the first plant of its type, vibration measurements were made on the reactor pressure vessel and its internal parts and on the primary piping system and the main coolant pumps. This paper contains the results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement programs is to confirm that the components, which are of new structural design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. A good correspondence was found. In the course of these comparisons, information about the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained. The vibration of the main coolant pumps was also continuously monitored. The pump surveillance system for each pump includes two non-contacting displacement sensors for measuring the kinetic shaft orbit, as well as velocity sensors for recording the vibrational velocity of the pump motor housing. During the continuous monitoring, it was checked whether the signal amplitudes remained within the allowable limits. In addition the frequency content of the signals was determined periodically. In this way deviations could be detected immediately and be explained by means of subsequent correlation analysis. Thus amplitude changes resulting from resonance effects were identified. (orig.)

  17. Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

    International Nuclear Information System (INIS)

    2003-01-01

    In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

  18. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  19. New cooling system of the FRG-1 two barrier system of the primary coolant cycle

    International Nuclear Information System (INIS)

    Knop, W.; Schreiner, P.

    2003-01-01

    The GKSS research center operates the swimming pool reactor FRG-1 with a thermal power of 5 MW as national neutron source for neutron scattering experiments and sample irradiation as well. Before changing the primary coolant cycle consisted of the reactor core and the closed piping including pumps, heat exchanger and delay tank. The closed cooling circuit was located underneath the reactor pool, in the so-called radioactive cellar. This piping system served secondary coolant system. Due to the location of the primary coolant cycle below the operation pool a postulated 2-F line break and simultaneous failure of the pool slide gate valve could lead to a falling dry of the total reactor core. the new primary coolant system was built in the beginning 2002 in a partitioned cell all within the radioactive cellar, so that the reactor core remains with water with the assumed incident. Due to the new two barrier-inclusion of the primary circuit only the melting of two fuel plates (from total 252 fuel plates) has to be taken into account. This measure and the core compactness in 2000 with a neutron flux gain of a factor of 2 makes the FRG-1 ready for the next 15 years of reactor operation. (author)

  20. PWR cold-leg small break loca with faulty HPI

    International Nuclear Information System (INIS)

    Kumamaru, H.; Kukita, Y.

    1991-01-01

    The ROSA-IV Large Scale Test Facility (LSTF) is a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). At the LSTF are performed cold-leg small-break loss-of-coolant accident (LOCA) tests with faulty high pressure injection (HPI) system for break areas from 0.5% to 10% and an intentional primary system depressurization test following a small-break LOCA test. A simple prediction model is proposed for prediction of times of major events. Test data and calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of 5% or more, and is insufficient for intermediate break areas to maintain adequate core cooling. (author)

  1. The Conceptual Design of Innovative Safe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han-Gon [Centural Research Institute, Daejeon (Korea, Republic of); Heo, Sun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    Most of countries operating NPPs have been performed post-Fukushima improvements as short-term countermeasure to enhance the safety of operating NPPs. Separately, vendors have made efforts on developing passive safety systems as long-term and ultimate countermeasures. AP1000 designed by Westinghouse Electric Company has passive safety systems including the passive emergency core cooling system (PECCS), the passive residual heat removal system (PRHRS), and the passive containment cooling system (PCCS). ESBWR designed by GE-Hitachi also has passive safety systems consisting of the isolation condenser system, the gravity driven cooling system and the PCCS. Other countries including China and Russia have made efforts on developing passive safety systems for enhancing the safety of their plants. In this paper, we summarize the design goals and main design feature of innovative safe PWR, iPOWER which is standing for Innovative Passive Optimized World-wide Economical Reactor, and show the developing status and results of research projects. To mitigate an accident without electric power and enhance the safety level of PWR, the conceptual designs of passive safety system and innovative safe PWR have been performed. It includes the PECCS for core cooling and the PCCS for containment cooling. Now we are performing the small scale and separate effect tests for the PECCS and the PCCS and preparing the integral effect test for the PECCS and real scale test for the PCCS.

  2. Reduction of PWR containment pressure after hypothetical accidents by water-cooling of the outer containment surface - annular space spray system

    International Nuclear Information System (INIS)

    Cremer, J.; Dietrich, D.P.; Roedder, P.

    1980-12-01

    The consequences of a core melt-out accident in the vicinity of a nuclear power station are determined by the integrity of the safety containment. This can be adversely affected by different events during the course of the core melt-out accident. The most important phenomenon is the contact between the melt and sump water. Due to the evaporation of the sump water, there is a continuous rise in pressure of the safety containment, which finally leads to failure due to excess pressure. In order to reduce the fission product release due to the resulting leakage, one must try to reduce the pressure as quickly as possible. As heat cannot be removed from the steel containment to the environment because of the thick concrete containment, it is best to bypass the insulating effect of the concrete by cooling the steel containment from outside. The aim of this investigation is therefore to work out a technically relatively simple system, which offers the possibility of backfitting, setting to work and repair. Such a system is an annular space spray system, by which the annular space between the concrete and steel containment has water pumped to the level of the dome and evenly sprayed over the top hemisphere. Mobile pumps on fire engines belonging to the fire brigade are sufficient to supply the cooling water and these will be available some hours after the accident occurs. The used spray water without any radioactive components is collected outside the reactor building and/or drained off. (orig./GL) [de

  3. Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

    International Nuclear Information System (INIS)

    Rahardjo, H.P.

    2011-01-01

    Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipment. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogeneously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occurred in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code) was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs) of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA) in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, expansion load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (re-routing) is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node 30, also a

  4. Maturity of the PWR

    International Nuclear Information System (INIS)

    Bacher, P.; Rapin, M.; Aboudarham, L.; Bitsch, D.

    1983-03-01

    Figures illustrating the predominant position of the PWR system are presented. The question is whether on the basis of these figures the PWR can be considered to have reached maturity. The following analysis, based on the French program experience, is an attempt to pinpoint those areas in which industrial maturity of the PWR has been attained, and in which areas a certain evolution can still be expected to take place

  5. Water chemical control of the TRIGA IPR-R1 reactor primary cooling system

    International Nuclear Information System (INIS)

    Auler, Lucia M.L.A; Chaves, Renata D.A.; Palmieri, Helena E.L.; Menezes, Maria Angela de B.C.; Oliveira, Paulo F.; Kastner, Geraldo F.; Damazio, Ilza; Fagundes, Oliene dos R.; Cintra, Maria Olivia C.; Andrade, Geraldo V. de; Amaral, Angela M.; Franco, Milton B.; Fortes, Flavio; Gomes, Nilton Carlos; Vidal, Andrea; Maretti Junior, Fausto; Knupp, Eliana A.N.; Souza, Wagner de; Guedes, Joao B.; Furtado, Renato C.S.

    2013-01-01

    The TRIGA Mark I IPR-R1 reactor located at CDTN/CNEN has been in operation and contributed to research and with services to society since 1960. Is has been used in several activities such as nuclear power plant operation, graduate and post-graduate training courses, isotope production, and as an analytical irradiation tool of different types of samples. Among the several structural and operational safety requirements is the chemical quality control of the primary circuit cooling water. The aim of this work was to check the cooling water quality from the pool reactor. A water sampling plan was proposed (May, 2011 - June, 2012) and presents the results obtained in this period. The natural radioactivity level as gross alpha and gross beta activity and other chemical parameters (pH and electric conductivity) of the samples were analyzed. Some instrumental techniques were used: potentiometric methods (pH), conductometric methods (electrical conductivity, EC) and gross α and gross β proportional counting system). (author)

  6. Estimation of maximum pressure in small containments of PWR reactors due to loss of coolant accident in primary circuit; Estimativa da pressao maxima em contencoes de reatores PWR de pequeno porte devido a um acidente de perda de refrigerante no circuito primario

    Energy Technology Data Exchange (ETDEWEB)

    Mendes Neto, Teofilo [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil); Moreira, Joao Manoel Losada [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2000-07-01

    This work studies the problem of containment pressurization after a LOCA in reactors with small containment free volumes. The relationship between the reactor power and the containment free volume is described with the ratio between the volumes of the primary circuit and of the containment. The maximum pressure in a containment, following a LOCA, obtained after a correlation based on large containment PWR, is around 185 psia for a primary circuit and containment volumes ratio of 0.025. For the same problem, calculations with the CONTEMPT-LT code produced a maximum pressure of 162 psia. The behavior of the temperature after a LOCA to the containment, as a function of the ratio between the primary circuit and containment volume, is such that it increases reaching asymptotically to a maximum; differently, the pressure increases almost linearly with the ratio of volumes. (author)

  7. Reverse primary-side flow in steam generators during natural circulation cooling

    International Nuclear Information System (INIS)

    Stumpf, H.; Motley, F.; Schultz, R.; Chapman, J.; Kukita, Y.

    1987-01-01

    A TRAC model of the Large Scale Test Facility with a 3-tube steam-generator model was used to analyze natural-circulation test ST-NC-02. For the steady state at 100% primary mass inventory, TRAC was in excellent agreement with the natural-circulation flow rate, the temperature distribution in the steam-generator tubes, and the temperature drop from the hot leg to the steam-generator inlet plenum. TRAC also predicted reverse flow in the long tubes. At reduced primary mass inventories, TRAC predicted the three natural-circulation flow regimes: single phase, two phase, and reflux condensation. TRAC did not predict the cyclic fill-and-dump phenomenon seen briefly in the test. TRAC overpredicted the two-phase natural-circulation flow rate. Since the core is well cooled at this time, the result is conservative. An important result of the analysis is that TRAC was able to predict the core dryout and heatup at approximately the same primary mass inventory as in the test. 4 refs., 8 figs., 2 tabs

  8. Cooling the intact loop of primary heat transport system using shut down cooling system after events such as LOCA

    International Nuclear Information System (INIS)

    Icleanu, D.L.

    2015-01-01

    The purpose of this paper is to model the Shutdown Cooling System operation for CANDU 6 NPP in case of LOCA accident, using Flowmaster calculation code by delimiting models and setting calculation assumptions and input data for hydraulic analysis, and and assumptions for the calculation and input data for calculating thermal performance check heat exchangers that are part of this system. The Flowmaster V7.8 code provides system engineers with a powerful tool to investigate pressure surge, pressure drop, flow rate, temperature and system response times - removing the uncertainty from fluid flow systems. Flowmaster is a one-dimensional thermal-hydraulic calculation code for dimensioning, analyzing and verifying the pipeline systems operation. Each component of Flowmaster is a mathematical model for an equipment that is included in a facility. Selected components are connected via nodes in order to form a network, which constitutes a computerized model of the system. Analyzing the parameters of the cooling system for all cooling processes considered it was found that the values obtained for thermal-hydraulic parameters, as well as the duration up to reaching specified limits fall within the design values of the system. This document is made up of an abstract and the slides of the presentation

  9. Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

    International Nuclear Information System (INIS)

    Carvalho, F. de A.T. de.

    1985-01-01

    Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

  10. PWR station blackout transient simulation in the INER integral system test facility

    International Nuclear Information System (INIS)

    Liu, T.J.; Lee, C.H.; Hong, W.T.; Chang, Y.H.

    2004-01-01

    Station blackout transient (or TMLB' scenario) in a pressurized water reactor (PWR) was simulated using the INER Integral System Test Facility (IIST) which is a 1/400 volumetrically-scaled reduce-height and reduce-pressure (RHRP) simulator of a Westinghouse three-loop PWR. Long-term thermal-hydraulic responses including the secondary boil-off and the subsequent primary saturation, pressurization and core uncovery were simulated based on the assumptions of no offsite and onsite power, feedwater and operator actions. The results indicate that two-phase discharge is the major depletion mode since it covers 81.3% of the total amount of the coolant inventory loss. The primary coolant inventory has experienced significant re-distribution during a station blackout transient. The decided parameter to avoid the core overheating is not the total amount of the coolant inventory remained in the primary core cooling system but only the part of coolant left in the pressure vessel. The sequence of significant events during transient for the IIST were also compared with those of the ROSA-IV large-scale test facility (LSTF), which is a 1/48 volumetrically-scaled full-height and full-pressure (FHFP) simulator of a PWR. The comparison indicates that the sequence and timing of these events during TMLB' transient studied in the RHRP IIST facility are generally consistent with those of the FHFP LSTF. (author)

  11. Primary coolant pipe rupture event in liquid metal cooled reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2004-08-01

    In liquid-metal cooled fast reactors (LMFR) the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). However, the primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors (Indira Gandhi Centre for Atomic Research, Kalpakkam, India, 13-17 January 2003) was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the technical meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the technical meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

  12. An experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V test SB-PV-04)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2006-03-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-04, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. It is clarified that AM actions with steam generator (SG) rapid depressurization by fully opening relief valves and auxiliary feedwater supply are effective to avoid core uncovery by actuating the low pressure injection (LPI) system though the primary depressurization is degraded by non-condensable gas inflow to the primary loops from the accumulator injection system. The effective core cooling was established by the rapid depressurization which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as AM actions. (author)

  13. Reliability Analysis of RSG-GAS Primary Cooling System to Support Aging Management Program

    Science.gov (United States)

    Deswandri; Subekti, M.; Sunaryo, Geni Rina

    2018-02-01

    Multipurpose Research Reactor G.A. Siwabessy (RSG-GAS) which has been operating since 1987 is one of the main facilities on supporting research, development and application of nuclear energy programs in BATAN. Until now, the RSG-GAS research reactor has been successfully operated safely and securely. However, because it has been operating for nearly 30 years, the structures, systems and components (SSCs) from the reactor would have started experiencing an aging phase. The process of aging certainly causes a decrease in reliability and safe performances of the reactor, therefore the aging management program is needed to resolve the issues. One of the programs in the aging management is to evaluate the safety and reliability of the system and also screening the critical components to be managed.One method that can be used for such purposes is the Fault Tree Analysis (FTA). In this papers FTA method is used to screening the critical components in the RSG-GAS Primary Cooling System. The evaluation results showed that the primary isolation valves are the basic events which are dominant against the system failure.

  14. Study on a corrosion products behavior with zinc injection in PWR primary circuit based on the crud investigation in actual plant

    International Nuclear Information System (INIS)

    Hisamune Kenji; Nambu Tohru; Akutagawa Daisuke; Nagata Nobuaki; Shimizu, Yuichi; Nagamine, Kunitaka; Kogawa Noritaka

    2012-09-01

    Zinc injection into the reactor coolant of Pressurized Water Reactor (PWR) is one of the most effective techniques for removing radioactive cobalt from the chromite type spinel oxides in the inner layer of reactor coolant component surfaces. Many documents have reported that zinc injection applied plants have generally achieved positive dose rate reduction effects for the reactor coolant components and piping. Tsuruga Nuclear Power Plant unit 2 (hereafter Tsuruga unit 2) which applied zinc injection in 2005 has also achieved good dose reduction effects in most primary coolant piping. However, some piping without dose-rate reduction effects exist, it has come to an issue to plan radiation exposure reduction measures for the works around relevant piping. To plan a reasonable dose-rate reduction measures for respective components and piping in primary systems, it is considered necessary to understand the dose-rate reduction effects of zinc injection mechanically. Therefore, the surface oxides/deposits on piping and fuel cladding tubes with operational histories under zinc injection were investigated. As a result of the investigation, the changes in specific aspects of oxide layers and CRUDs are found as follow: - Owing to long-term zinc injection, deposits and oxides on the primary coolant piping are extremely reduced, the boundary between the outer oxide layer and the inner oxide layer become unclear with the thinning of the outer oxide layers. - At a Pressurizer Spray Piping, the dose-rate increases than before zinc injection. The oxide layers have the same 3-layer construction as before zinc injection, and the surface concentration of the radioactive cobalt is high. - At the fuel cladding tube surfaces, what is observed is the amount of metal nickel decrease and the chromium composition increase in the spinel type oxide From these results, it is suggested that, the oxide layers composed principally of zinc chromite spinel grow up on the pipe surface with zinc

  15. Cooling methods of station blackout scenario for LWR plants

    International Nuclear Information System (INIS)

    2012-01-01

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

  16. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject D. Study on water film cooling for PWR's passive containment cooling system. Final report

    International Nuclear Information System (INIS)

    Huang, Xi

    2016-07-01

    In the present study, a new phenomenological model was developed, to describe the water film flow under conditions of a passive containment cooling system (PCCS). The new model takes two different flow regimes into consideration, i.e. continuous water film and rivulets. For water film flow, the traditional Nusselt's was modified, to consider orientation angle and surface sheer stress. The transition from water film to rivulet as well as the structure of the stable rivulet at its onset point was modeled by using the minimum energy principle (MEP) combined with conservation equations. In addition, two different contact angles, i.e. advancing angle and retreating angle, were applied to take the hysteresis effect into consideration. The models of individual processes were validated as far as possible based on experimental data selected from open literature and from collaboration partner as well. With the models a new program module was developed and implemented into the COCOSYS program. The extended COCOSYS program was applied to analyze the containment behavior of the European generic containment and the performance of the passive containment cooling system ofthe AP1000. The results indicate clearly the importance of the new model and provide information for the optimization of the PCCS of AP1000.

  17. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject D. Study on water film cooling for PWR's passive containment cooling system. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Xi

    2016-07-15

    In the present study, a new phenomenological model was developed, to describe the water film flow under conditions of a passive containment cooling system (PCCS). The new model takes two different flow regimes into consideration, i.e. continuous water film and rivulets. For water film flow, the traditional Nusselt's was modified, to consider orientation angle and surface sheer stress. The transition from water film to rivulet as well as the structure of the stable rivulet at its onset point was modeled by using the minimum energy principle (MEP) combined with conservation equations. In addition, two different contact angles, i.e. advancing angle and retreating angle, were applied to take the hysteresis effect into consideration. The models of individual processes were validated as far as possible based on experimental data selected from open literature and from collaboration partner as well. With the models a new program module was developed and implemented into the COCOSYS program. The extended COCOSYS program was applied to analyze the containment behavior of the European generic containment and the performance of the passive containment cooling system ofthe AP1000. The results indicate clearly the importance of the new model and provide information for the optimization of the PCCS of AP1000.

  18. Monte Carlo Simulation of Quantitative Electron Probe Microanalysis of the PWR Spent Fuel with a Pt Coating

    International Nuclear Information System (INIS)

    Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum

    2012-01-01

    The PWR spent fuel sample should be coated with conducting material in order to provide a path for electrons and to prevent charging. Generally, the ZAF method has been used for quantitative electron probe microanalysis of conducting samples. However, the coated samples are not applicable for the ZAF method. Probe current, primary electron energy and x-ray produced by the primary beam are attenuated within the coating films. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program [2] to evaluate the x-ray attenuation within the Pt coating films. The target samples are the PWR spent fuels with 50 GWd/tU of burnup , 6 years of cooling time and a Pt coating film (3, 5, 7, 10 and 15 nm thickness)

  19. Monte Carlo Simulation of Quantitative Electron Probe Microanalysis of the PWR Spent Fuel with a Pt Coating

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The PWR spent fuel sample should be coated with conducting material in order to provide a path for electrons and to prevent charging. Generally, the ZAF method has been used for quantitative electron probe microanalysis of conducting samples. However, the coated samples are not applicable for the ZAF method. Probe current, primary electron energy and x-ray produced by the primary beam are attenuated within the coating films. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program [2] to evaluate the x-ray attenuation within the Pt coating films. The target samples are the PWR spent fuels with 50 GWd/tU of burnup , 6 years of cooling time and a Pt coating film (3, 5, 7, 10 and 15 nm thickness)

  20. Habitable planets around white and brown dwarfs: the perils of a cooling primary.

    Science.gov (United States)

    Barnes, Rory; Heller, René

    2013-03-01

    White and brown dwarfs are astrophysical objects that are bright enough to support an insolation habitable zone (IHZ). Unlike hydrogen-burning stars, they cool and become less luminous with time; hence their IHZ moves in with time. The inner edge of the IHZ is defined as the orbital radius at which a planet may enter a moist or runaway greenhouse, phenomena that can remove a planet's surface water forever. Thus, as the IHZ moves in, planets that enter it may no longer have any water and are still uninhabitable. Additionally, the close proximity of the IHZ to the primary leads to concern that tidal heating may also be strong enough to trigger a runaway greenhouse, even for orbital eccentricities as small as 10(-6). Water loss occurs due to photolyzation by UV photons in the planetary stratosphere, followed by hydrogen escape. Young white dwarfs emit a large amount of these photons, as their surface temperatures are over 10(4) K. The situation is less clear for brown dwarfs, as observational data do not constrain their early activity and UV emission very well. Nonetheless, both types of planets are at risk of never achieving habitable conditions, but planets orbiting white dwarfs may be less likely to sustain life than those orbiting brown dwarfs. We consider the future habitability of the planet candidates KOI 55.01 and 55.02 in these terms and find they are unlikely to become habitable.

  1. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  2. Vibration test on KMRR reactor structure and primary cooling system piping

    International Nuclear Information System (INIS)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

  3. Primary design of Si cooling arm structure in ICF cryogenic target

    International Nuclear Information System (INIS)

    Zhang Yong; Yi Yong; Tang Changhuan; Zhang Jicheng

    2013-01-01

    According to the requirement of the cryogenic target system to the Si cooling arm structure, the Si cooling arm was primarily designed based on the USA National Ignition Facility (NIF) target. A new three-dimensional model of Si cooling arm was developed by SolidWorks software, and the simulation and analysis of Si cooling arm in aspect of mechanical property, thermal response and assembly were made based on the model. A law about the effect of the arm length of Si cooling arm and the width and the length of bifurcation on Si cooling arm was achieved. The research may provide the theoretical foundation and reference for the further improvement of cryogenic target. (authors)

  4. Primary energy savings in desiccant and evaporative cooling-assisted 100% outdoor air system combined with a fuel cell

    International Nuclear Information System (INIS)

    Kim, Min-Hwi; Dong, Hae-Won; Park, Joon-Young; Jeong, Jae-Weon

    2016-01-01

    Highlights: • A LD-IDECOAS integrated with a PEMFC was proposed. • A pilot system was installed and tested during cooling operation. • The proposed system powered by the PEMFC saved 21% of the primary energy consumption during cooling. - Abstract: The main purpose of this study involved investigating the primary energy saving potential of a liquid desiccant and evaporative cooling-assisted 100% outdoor air system (LD-IDECOAS) integrated with a proton exchange membrane fuel cell (PEMFC). During the cooling season, the heat produced by the PEMFC was used to regenerate a weak desiccant solution, and the electricity generated was used to operate the LD-IDECOAS. A pilot LD-IDECOAS powered by a PEMFC was installed and operated in an office space to experimentally verify the annual operating energy savings of the proposed system. The findings indicated that the heat reclaimed from the PEMFC saved 42% of the desiccant solution regenerating energy when compared to that in the case of a conventional gas-fired water heater. The results also suggested that the LD-IDECOAS combined with a PEMFC consumed 21% less primary energy when compared with that of a system powered by grid electricity and a conventional gas-fired water heater.

  5. Characteristic Study of the Al 6061 T-6 used in RTP Primary Cooling System Using Scanning Electron Microscope (SEM)

    International Nuclear Information System (INIS)

    Tonny Anak Lanyau; Yusof Abdullah; Tom, P.P.

    2011-01-01

    Reactor TRIGA PUSPATI (RTP) is the only nuclear research reactor in Malaysia. Since the first criticality on 28th June 1982, RTP has been going through the safe operation and well maintenance. Along the period of operation almost 30 years, some of the reactor system and component has been refurbished, upgraded and replaced to ensure the functionality and safety to the reactor itself as well as to protect personnel and environment. Primary cooling system is to provide the sufficient cooling to the reactor by removal of the heat generated in the reactor core through the heat transfer process in the heat exchanger. In 2009, RTP has been undergoing the primary cooling system upgrades. Primary cooling system components including aluminium pipes has been dismantled and replaced with the new system. As a part of the ageing management programme and radiation damage study, the disposed aluminum pipes were taken and used in this study. Scanning Electron Microscope (SEM) is used to study the surface topography and elemental composition in conjunction of energy dispersive x-ray spectroscopy analysis. This paper presents the study that has been conducted. (author)

  6. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  7. Improvement of PWR reliability by corrosion prevention

    International Nuclear Information System (INIS)

    Takamatsu, Hiroshi

    1999-01-01

    Since first PWR in Japan started commercial operation in 1970, we have encountered the various modes of corrosion on primary and secondary side components. We have paid much efforts for resolving these corrosion problems, that is, investigating the causes of corrosion and establishing the countermeasures for these corrosion. We summarize these efforts in this article. (author)

  8. Habitable Planets Around White and Brown Dwarfs: The Perils of a Cooling Primary

    Science.gov (United States)

    Heller, René

    2013-01-01

    Abstract White and brown dwarfs are astrophysical objects that are bright enough to support an insolation habitable zone (IHZ). Unlike hydrogen-burning stars, they cool and become less luminous with time; hence their IHZ moves in with time. The inner edge of the IHZ is defined as the orbital radius at which a planet may enter a moist or runaway greenhouse, phenomena that can remove a planet's surface water forever. Thus, as the IHZ moves in, planets that enter it may no longer have any water and are still uninhabitable. Additionally, the close proximity of the IHZ to the primary leads to concern that tidal heating may also be strong enough to trigger a runaway greenhouse, even for orbital eccentricities as small as 10−6. Water loss occurs due to photolyzation by UV photons in the planetary stratosphere, followed by hydrogen escape. Young white dwarfs emit a large amount of these photons, as their surface temperatures are over 104 K. The situation is less clear for brown dwarfs, as observational data do not constrain their early activity and UV emission very well. Nonetheless, both types of planets are at risk of never achieving habitable conditions, but planets orbiting white dwarfs may be less likely to sustain life than those orbiting brown dwarfs. We consider the future habitability of the planet candidates KOI 55.01 and 55.02 in these terms and find they are unlikely to become habitable. Key Words: Extrasolar terrestrial planets—Habitability—Habitable zone—Tides—Exoplanets. Astrobiology 13, 279–291. PMID:23537137

  9. Study of water radiolysis in relation with the primary cooling circuit of pressurized water reactors; Etude sur la radiolyse de l`eau en relation avec le circuit primaire de refroidissement des reacteurs nucleaires a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Pastina, B

    1997-07-01

    This memorandum shows a fundamental study on the water radiolysis in relation with the cooling primary circuit of PWR type reactors. The water of the primary circuit contains boric acid a soluble neutronic poison and also hydrogen that has for role to inhibit the water decomposition under radiation effect. In the aim to better understand the mechanism of dissolved hydrogen action and to evaluate the impact of several parameters on this mechanism, aqueous solutions with boric acid and hydrogen have been irradiated in a experimental nuclear reactor, at 30, 100 and 200 Celsius degrees. It has been found that, with hydrogen, the water decomposition under irradiation is a threshold phenomenon in function of the ratio between the radiation flux `1` B(n, )`7 Li and the gamma flux. When this ratio become too high, the number of radicals is not sufficient to participate at the chain reaction, and then water is decomposed in O{sub 2} and H{sub 2}O{sub 2} in a irreversible way. The temperature has a beneficial part on this mechanism. The iron ion and the copper ion favour the water decomposition. (N.C.). 83 refs.

  10. Corrosion inhibition measures in primary cooling water system during refurbishment of Cirus, re-commissioning and subsequent operation

    International Nuclear Information System (INIS)

    Rai, K.K.; Ramesh, N.; Sharma, R.C.

    2008-01-01

    Cirus is a 40 MWth, heavy water moderated, demineralized light water cooled, natural uranium fuelled research reactor. Reactor was commissioned in year 1960 and operated satisfactorily till 1990. After that availability factor started decreasing mainly due to equipment outage exhibiting signs of ageing. Based upon systematic ageing studies and assessment of condition of systems, structures and components, a refurbishment plan including safety upgrades was drawn up. Reactor was shut down in October 1997 for execution of jobs. After completion of refurbishment jobs reactor was started back in October 2002 and power operation was achieved in 2003. Primary cooling water (PCW) system consists of re-circulating pumps, heat exchangers, expansion tank, piping, valves, emergency storage reservoir (Ball Tank) and other components. Normally the fission heat from fuel is removed by re-circulating coolant in closed loop and transferred to seawater via heat exchangers. In case of outage of pumps, shut down cooling is provided by flow of water from Ball Tank under gravity to the underground dump tanks. The dissolved oxygen is maintained below 2 ppm and pH is maintained neutral to minimize corrosion of fuel cladding (Aluminum). This paper highlights the experience gained during segmentation of primary cooling water pipelines for pressure testing, measures taken to corrosion inhibition of primary cooling water lines to permit execution of refurbishment jobs, inspections and actions taken to repair/replace the corroded PCW pipe line segments, observations regarding corrosion related failures, re-commissioning of the system after refurbishment, assessment for safe reactor operation and experience during power operation. (author)

  11. A study on effective system depressurization during a PWR vessel bottom break LOCA with HPI failure and gas inflow prevention. ROSA-V/LSTF test SB-PV-05

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2006-11-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-05, simulated a PWR vessel bottom SBLOCA with a rupture of nine instrument tubes, which is equivalent to 0.18% cold leg break. It is clarified that AM actions with steam generator (SG) depressurization to achieve a primary loop cooling rate at -55 K/h and auxiliary feedwater supply for 30 minutes are effective to avoid core uncovery by actuating the low pressure injection (LPI) system. It is also shown through the comparison with the previous experiment of SB-PV-03 that prevention of non-condensable gas inflow from the accumulator injection system (AIS) is very important to actuate the LPI to achieve adequate core cooling. This report presents experiment results of SB-PV-05 in detail and shows the effects of gas inflow prevention on core cooling through the estimation of primary coolant mass and energy balance in the primary system. (author)

  12. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  13. A Comparison of Models Describing Heat Transfer in the Primary Cooling Zone of a Continuous Casting Machine

    Directory of Open Access Journals (Sweden)

    Miłkowska-Piszczek K.

    2015-04-01

    Full Text Available This paper presents the findings of research conducted concerning the determination of thermal boundary conditions for the steel continuous casting process within the primary cooling zone. A cast slab - with dimensions of 1100 mm×220 mm - was analysed, and models described in references were compared with the authors’ model. The presented models were verified on the basis of an industrial database. The research problem was solved with the finite element method using the ProCAST software package.

  14. Influence of localized deformation on A-286 austenitic stainless steel stress corrosion cracking in PWR primary water; Influence de la localisation de la deformation sur la corrosion sous contrainte de l'acier inoxydable austenitique A-286 en milieu primaire des REP

    Energy Technology Data Exchange (ETDEWEB)

    Savoie, M

    2007-01-15

    Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels is known to be a critical issue for structural components of nuclear reactor cores. The deformation of irradiated austenitic stainless steels is extremely heterogeneous and localized in deformation bands that may play a significant role in IASCC. In this study, an original approach is proposed to determine the influence of localized deformation on austenitic stainless steels SCC in simulated PWR primary water. The approach consists in (i) performing low cycle fatigue tests on austenitic stainless steel A-286 strengthened by {gamma}' precipitates Ni{sub 3}(Ti,Al) in order to shear and dissolve the precipitates in intense slip bands, leading to a localization of the deformation within and in (ii) assessing the influence of these {gamma}'-free localized deformation bands on A-286 SCC by means of comparative CERT tests performed on specimens with similar yield strength, containing or not {gamma}'-free localized deformation bands. Results show that strain localization significantly promotes A-286 SCC in simulated PWR primary water at 320 and 360 C. Moreover, A-286 is a precipitation-hardening austenitic stainless steel used for applications in light water reactors. The second objective of this work is to gain insights into the influence of heat treatment and metallurgical structure on A-286 SCC susceptibility in PWR primary water. The results obtained demonstrate a strong correlation between yield strength and SCC susceptibility of A-286 in PWR primary water at 320 and 360 C. (author)

  15. Aspects of PWR nuclear power plant secondary cycle relating to reactor safety

    International Nuclear Information System (INIS)

    Mueller, A.E.F.; Leal, M.R.L.V.; Dominguez, D.

    1981-01-01

    A safety study of the main steam system, condensate and feedwater systems and water treatment system that belong to the secondary cooling circuits of a PWR nuclear power plant is presented. (E.G.) [pt

  16. PWR secondary water chemistry study

    International Nuclear Information System (INIS)

    Pearl, W.L.; Sawochka, S.G.

    1977-02-01

    Several types of corrosion damage are currently chronic problems in PWR recirculating steam generators. One probable cause of damage is a local high concentration of an aggressive chemical even though only trace levels are present in feedwater. A wide variety of trace chemicals can find their way into feedwater, depending on the sources of condenser cooling water and the specific feedwater treatment. In February 1975, Nuclear Water and Waste Technology Corporation (NWT), was contracted to characterize secondary system water chemistry at five operating PWRs. Plants were selected to allow effects of cooling water chemistry and operating history on steam generator corrosion to be evaluated. Calvert Cliffs 1, Prairie Island 1 and 2, Surry 2, and Turkey Point 4 were monitored during the program. Results to date in the following areas are summarized: (1) plant chemistry variations during normal operation, transients, and shutdowns; (2) effects of condenser leakage on steam generator chemistry; (3) corrosion product transport during all phases of operation; (4) analytical prediction of chemistry in local areas from bulk water chemistry measurements; and (5) correlation of corrosion damage to chemistry variation

  17. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

  18. Maximization of primary energy savings of solar heating and cooling systems by transient simulations and computer design of experiments

    International Nuclear Information System (INIS)

    Calise, F.; Palombo, A.; Vanoli, L.

    2010-01-01

    In this paper, the simulation of the performance of solar-assisted heating and cooling systems is analyzed. Three different plant layouts are considered: (i) the first one consists of evacuated solar collectors and a single-stage LiBr-H 2 O absorption chiller; here in order to integrate the system in case of insufficient solar radiation, an electric water-cooled chiller is activated; (ii) configuration of the secondly considered system is similar to the first one, but the absorption chiller and the solar collector area are sized for balancing about 30% of the building cooling load only; (iii) the layout of the thirdly considered system differs from the first one since the auxiliary electric chiller is replaced by a gas-fired heater. Such system configurations also include: circulation pumps, storage tanks, feedback controllers, mixers, diverters and on/off hysteresis controllers. All such devices are modelled for maximizing the system energy efficiency. In order to simulate the systems' performance for dynamic heating/cooling loads, a single-lumped capacitance building is also modelled and implemented in the computer code. A cost model is also developed in order to calculate the systems' operating and capital costs. All the models and the relative simulations are carried out by TRNSYS. A design of experiment procedure is also included. By such tool the effects of the system operating parameters' variation on the relative energy efficiency are analyzed. In addition, the set of synthesis/design variables maximizing the system's energetic performance can be also identified. The annual primary energy saving is chosen as the optimization objective function, whereas collector slope, pump flows, set-point temperatures and tank volume are selected as optimizing system design variables. A case study was developed for an office building located in South Italy. Here, the energetic and the economic analysis for all the three considered system layouts are carried out. The

  19. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  20. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

    Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

  1. Analysis of breaks in pipe connections to the hot leg and to the loop seal in the primary system of Ringhals 2 PWR

    International Nuclear Information System (INIS)

    Nilsson, L.; Sjoeberg, A.

    1987-01-01

    Analysis has been made of seven different cases of breaks in pipes connected to the hot leg and to the loop seal in Ringhals 2 PWR. The pipes, in which guillotine breaks are postulated, have nominal diameters ranging from 1 to 14 inches. The purpose of the analysis is to supplement the basis for a review of the inspection procedures for the safety of pressure vessels prescribed by SKI. A similar analysis already exists concerning breaks in the cold leg connections. The analysis has been made using the thermal hydraulic computer code RELAPS/MOD2 and with best estimate assumptions. This means that normal operating conditions have been adopted for the input to the calculations. However, the capacity of the safety injection system was assumed to be reduced by having one pump not operating each of the low pressure and high pressure safety injection system. The results of the analysis are presented in tables and as computer plots. The analysis shows that the consequences with respect to increased fuel rod and cladding temperatures are quite harmless. Only the two cases with the largest break sizes lead to critical heat flux (CHF) and increased temperatures for the hottest rods in the core. The peak cladding temperature was 636 degrees C for the 12 inch break. In both cases rewetting occurred within 25 s of accident initiation. In the cases with breaks in connections of 6 inch diameter and smaller the RELAP5 calculations indicated a substantial margin to CHF throughout the transient. (authors)

  2. Evaluation Of Radioactivity Concentration In The Primary Cooling Water System Of The RSG-GAS During Operation With 30% Silicide Fuels

    International Nuclear Information System (INIS)

    Hartoyo, Unggul; Udiyani, P.M.; Setiawanto, Anto

    2001-01-01

    The evaluating radioactivity concentration in the primary cooling water of the RSG-GAS during operation with 30% silicide fuels has been performed. The method of the research is sampling of primary cooling water during operation of the reactor and calculation of its radioactivity concentration. Based on the data obtained from calculation, the identified nuclides in the water are, Mn-56, Sb-124, Sb-122 and Na-24, under the limit of safety value

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  4. Dresden Nuclear Power Station, Unit No. 1: Primary cooling system chemical decontamination: Draft environmental statement (Docket No. 50-10)

    International Nuclear Information System (INIS)

    1980-05-01

    The staff has considered the environmental impact and economic costs of the proposed primary cooling system chemical decontamination at Dresden Nuclear Power Station, Unit 1. The staff has focused this statement on the occupational radiation exposure associated with the proposed Unit 1 decontamination program, on alternatives to chemical decontamination, and on the environmental impact of the disposal of the solid radioactive waste generated by this decontamination. The staff has concluded that the proposed decontamination will not significantly affect the quality of the human environment. Furthermore, any impacts from the decontamination program are outweighed by its benefits. 2 figs., 7 tabs

  5. Thermodynamic data for selected gas impurities in the primary coolant of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Feber, R.C.

    1976-12-01

    The literature of thermodynamic data for selected fission-product species is reviewed and supplemented in support of complex chemical equilibrium calculations applied to fission-product distributions in the primary coolant of high-temperature gas-cooled reactors. Thermodynamic functions and heats and free energies of formation are calculated and tabulated to 3000 0 K for CsI (s,l,g), Cs 2 I 2 (g), CH 3 I(g), COI 2 (g), and CsH(g). 79 references

  6. Ventilation and air-conditioning system for PWR nuclear power plant

    International Nuclear Information System (INIS)

    Ohmoto, Kenji

    1987-01-01

    This report outlines the ventilation and air conditioning facilities for PWR nuclear power plant as well as design re-evaluation and optimization of ventilation and air-conditioning. The primary PWR installations are generally housed in the nuclear reactor building, auxiliary buildings and control building, which are equipped with their own ventilation and air-conditioning systems to serve for their specific purposes. A ventilation/air-conditioning system should be able to work effectively not only for maintaining the ordinary reactor operation but also for controlling the environmental temperature in the event of an accident. Designing of a ventilation/air-conditioning system relied on empirical data in the past, but currently it is performed based on information obtained from various analyses to optimize the system configuration and ventilation capacity. Design re-evaluation of ventilation/air-conditioning systems are conducted widely in various areas, aiming at the integration of safety systems, optimum combination of air-cooling and water-cooling systems, and optimization of the ventilation rate for controlling the concentrations of radioactive substances in the atmosphere in the facilities. It is pointed out that performance evaluation of ventilation/air-conditioning systems, which has been conducted rather macroscopically, should be carried out more in detal in the future to determine optimum air streams and temperature distribution. (Nogami, K.)

  7. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  8. Probable variations of a passive safety containment for a 1700 MWe class PWR with passive safety systems

    International Nuclear Information System (INIS)

    Sato, Takashi; Fujiki, Yasunobu; Oikawa, Hirohide; Ofstun, Richard P.

    2009-01-01

    The paper presents probable variations of a passive safety containment for a PWR. The passive safety containment is named Mark P containment tentatively. It is a pressure suppression type containment for a large scale PWR with a BWR type passive containment cooling system (PCCS). More than 3-day grace period can be achieved even for a 1700 MWe class large scale PWR owing to the PCCS. The containment is a reinforced concrete containment vessel (RCCV). The design pressure of the RCCV can be low owing to the suppression pool (S/P) and no prestressed tendon is necessary. It is a single barrier CV that can withstand a large airplane crash by itself. This simple configuration results in good economy and short construction term. The BWR type passive safety systems also include the Passive Cooling and Depressurization System (PCDS). The PCDS has 3-day grace period for the SBO induced by a giant earthquake and can practically eliminate the residual risk of a giant earthquake beyond the design basis earthquake of Ss. It also has a safety function to automatically depressurize the primary system at accidents such as SGTR and eliminate the need for operator actions. It is a large 1700 MWe passive safety PWR that has more than 3-day grace period for extremely severe natural disasters including a giant earthquake, a mega hurricane, tsunami and so on; no containment failure at a SA establishing a no evacuation plant; protection for a large airplane crash with the RCCV single barrier; good economy and short construction term. (author)

  9. Differential roles of galanin on mechanical and cooling responses at the primary afferent nociceptor

    Directory of Open Access Journals (Sweden)

    Hulse Richard P

    2012-06-01

    Full Text Available Abstract Background Galanin is expressed in a small percentage of intact small diameter sensory neurons of the dorsal root ganglia and in the afferent terminals of the superficial lamina of the dorsal horn of the spinal cord. The neuropeptide modulates nociception demonstrating dose-dependent pro- and anti-nociceptive actions in the naïve animal. Galanin also plays an important role in chronic pain, with the anti-nociceptive actions enhanced in rodent neuropathic pain models. In this study we compared the role played by galanin and its receptors in mechanical and cold allodynia by identifying individual rat C-fibre nociceptors and characterising their responses to mechanical or acetone stimulation. Results Mechanically evoked responses in C-fibre nociceptors from naive rats were sensitised after close intra-arterial infusion of galanin or Gal2-11 (a galanin receptor-2/3 agonist confirming previous data that galanin modulates nociception via activation of GalR2. In contrast, the same dose and route of administration of galanin, but not Gal2-11, inhibited acetone and menthol cooling evoked responses, demonstrating that this inhibitory mechanism is not mediated by activation of GalR2. We then used the partial saphenous nerve ligation injury model of neuropathic pain (PSNI and the complete Freund’s adjuvant model of inflammation in the rat and demonstrated that close intra-arterial infusion of galanin, but not Gal2-11, reduced cooling evoked nociceptor activity and cooling allodynia in both paradigms, whilst galanin and Gal2-11 both decreased mechanical activation thresholds. A previously described transgenic mouse line which inducibly over-expresses galanin (Gal-OE after nerve injury was then used to investigate whether manipulating the levels of endogenous galanin also modulates cooling evoked nociceptive behaviours after PSNI. Acetone withdrawal behaviours in naive mice showed no differences between Gal-OE and wildtype (WT mice. 7-days after

  10. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  11. The study on radioactivity reduction of spent PWR cladding hull

    International Nuclear Information System (INIS)

    Jung, I. H.; Kim, J. H.; Park, C. J.; Jung, Y. H.; Song, K. C.; Lee, J. W.; Park, J. J.; Yang, M. S.

    2003-01-01

    Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull-from PWR spent fuel of 32,000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4-7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radioactive waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull

  12. Improvement of Corrosion Inhibitors of Primary and Secondary Closed Cooling Water System

    International Nuclear Information System (INIS)

    Choi, Byung Seon; Kim, K. M.; Kim, K. H.

    2010-08-01

    In nuclear power plants, the Closed Cooling Water (CCW) system provide cooling to both safety-related and non-safety-related heat exchange equipment. Various chemicals are used to mitigate corrosion, fouling, and microbiological growth in the CCW systems. In nuclear plants, these inhibitors have included chromates, nitrites, molybdates, hydrazine, and silicate. In the case of the CCW of some domestic nuclear power plants, there is during the overhaul period, a saturation of ion exchange resin caused by an inhibitor which has high conductivity for an increase in radiation exposure and radioactive waste. The objective of this study is to evaluate the corrosion behavior of structural materials with various corrosion inhibitors. In the present study, more than 50 ppm hydrazine concentration is needed to reduce the corrosion rate of carbon steel to satisfy the CCW operational guidelines. However, if hydrazine is continuously injected into the CCW system, the critical concentration of hydrazine will be lower. Hydrazine might be an alternative corrosion inhibitor for nitrite in the CCW system of nuclear power plant

  13. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors; Modelisation en dynamique rapide d'accidents dans le circuit primaire des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  14. Scaling studies - PWR

    International Nuclear Information System (INIS)

    Sonneck, G.

    1983-05-01

    A RELAP 4/MOD 6 study was made based on the blowdown phase of the intermediate break experiment LOFT L5-1. The method was to set up a base model and to vary parametrically some areas where it is known or suspected that LOFT differs from a commercial PWR. The aim was not to simulate LOFT or a PWR exactly but to understand the influence of the following parameters on the thermohydraulic behaviour of the system and the clad temperature: stored heat in the downcomer (LOFT has rather large filler blocks in this part of the pressure vessel); bypass between downcomer and upper plenum; and core length. The results show that LOFT is prototypical for all calculated blowdowns. As the clad temperatures decrease with decreasing stored energy in the downcomer, increased bypass and increased core length, LOFT results seem to be realistic as long as realistic bypass sizes are considered; they are conservative in the two other areas. (author)

  15. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

    Youinou, G.; Girieud, R.; Guigon, B.

    2000-01-01

    Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)

  16. The integrated PWR

    International Nuclear Information System (INIS)

    Gautier, G.M.

    2002-01-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  17. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  18. Requirements on cast steel for the primary coolant circuit of water cooled reactors

    International Nuclear Information System (INIS)

    The most important requirements placed on the structural components of water cooled nuclear reactors include corrosion resistance and mechanical materials properties. Intercrystalline corrosion resistance was tested using the Strauss Test in compliance with the DIN 50914 Standard. Following sensitization between 600 to 700 degC with a dwell time between 15 minutes and 100 hours, a specimen homogeneously annealed with the casting and rapidly water cooled showed no intercrystalline corrosion. Specimens cooled from 1050 degC at a rate of 100 degC per hour showed no unambiguous tendency for intercrystalline corrosion after sensitization; in some cases, however, an initial attack of intercrystalline corrosion was found. It was found that austenitic Cr-Ni cast steel containing 2.5% Mo and about 15% ferrite showed the sensitive intercrystalline corrosion range at higher temperatures and longer dwell times than rolled Cr-Ni steels. In plating the ferritic cast steel with a corrosion resistant plating material, annealing temperature after welding must not exceed 600 to 620 degC otherwise the resistance of the plated layer against intercrystalline corrosion would not be safeguarded, and following annealing for stress removal at a temperature of 600 to 620 degC all requirements must be satisfied by the weld metal and weld transition placed on the initial material. Martensite materials are used for the manufacture of components which are not used under pressure, such as alloys with 13% Cr and 1% to 6% Ni and alloys with 17% Cr and 4% Ni. Carbon content is maintained below 0.10% to guarantee good weldability and the highest corrosion resistance. Cast steels with 13% Cr and 4% Ni after a dwell of 2500 hours in fully desalinated water without oxygen and with 3600 ppm of boron at a test temperature of 95 to 300 degC showed a surface reduction of 0.005 mm annually. In identical conditions except for the water containing oxygen the reduction in surface was 0.05 mm per year. (J.B.)

  19. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    International Nuclear Information System (INIS)

    Valle H, J.; Hidago H, F.; Morales S, J.B.

    2007-01-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  20. PWR: 10 years after and perspectives

    International Nuclear Information System (INIS)

    1990-01-01

    These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

  1. Basic information about development and construction of a PWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1977-01-01

    1.0) Plant layout of a PWR; 2.0) principle design of a PWR and the reactor coolant system; 3.0) reactor auxiliary and ancillary systems; 3.1) volume control system; 3.2) boric acid control and chemical feeding system; 3.3) coolant purification and degassing system; 3.4) coolant storage and treatment system; 3.5) nuclear component cooling system; 3.6) liquid waste processing system; 3.7) gaseous waste processing system; 4.0) residual heat removal system; 5.0) emergency feedwater system; 6.0) containment design; 7.0) fuel handling, storage and transport system in a PWR. (orig.) [de

  2. Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

    A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950 0 K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated

  3. Measurement of transient hydrodynamic characteristics of the reactor RA primary cooling system

    International Nuclear Information System (INIS)

    Jovic, L.; Majstorovic, D.; Zeljkovic, I.

    1987-01-01

    Experimental study of transient hydrodynamic characteristics of the research nuclear reactor RA by simultaneous measurements of fluid flow and pressure on several locations of the RA primary coolant system is done. Loss of electric power transient on the main circulation pumps is simulated. measurement methodology, data processing and results of measured data analysis are given. (author)

  4. Materials performance in operating PWR steam generators

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

  5. Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water; Facteurs metallurgiques et mecaniques controlant l'amorcage de defauts de corrosion sous contrainte dans l'alliage 718 en milieu primaire des reacteurs a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Deleume, J

    2007-11-15

    Improving the performance and reliability of the fuel assemblies of the pressurized water reactors requires having a perfect knowledge of the operating margins of both the components and the materials. The choice of alloy 718 as reference material for this study is justified by the industrial will to identify the first order parameters controlling the excellent resistance of this alloy to Stress Corrosion Cracking (SCC). For this purpose, a specific slow strain rate (SSR) crack initiation test using tensile specimen with a V-shaped hump in the middle of the gauge length was developed and modeled. The selectivity of such SSR tests in simulated PWR primary water at 350 C was clearly established by characterizing the SCC resistance of nine alloy 718 thin strip heats. Regardless of their origin and in spite of a similar thermo-mechanical history, they did not exhibit the same susceptibility to SCC crack initiation. All the characterized alloy 718 heats develop oxide scale of similar nature for various exposure times to PWR primary medium in the temperature range [320 C - 360 C]. {delta} phase precipitation has no impact on alloy 718 SCC initiation behavior when exposed to PWR primary water, contrary to interstitial contents and the triggering of plastic instabilities (PLC phenomenon). (author)

  6. Proposals for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Bolt, P.R.

    1976-01-01

    Design and operational experience of CEGB gas cooled reactors and certain overseas reactor plant is reviewed in relation to in-service inspection and monitoring capabilities. Design guidelines and preliminary proposals are given for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors. Specific comments are made on the items of further design and development work believed to be necessary

  7. Mechanical Properties of Post Irradiation Primary Cooling Piping of Bandung Research Reactor

    International Nuclear Information System (INIS)

    Histori; Renaningsih S; Sri Nitiswati; Ari Triyadi

    2003-01-01

    Testing on primary coolant piping of research reactor Bandung have been done. Primary coolant piping were made from Al 6061-T6. The goal of this activity is to investigate the mechanical properties changes caused by aging process after 33 years in irradiated. Type of testing i.e visual examination, thickness measurement, tensile and hardness test were done. The test data shown that there was a deposit at the inside surface of pipe, thickness decreased about 0.2 mm, tensile strength is 293 MPa, yield strength is 262 MPa, while the hardness is about 83 HRE (mean value). The test data than compared with ASTM standard. As the conclusion tensile and yield strength of pipe still fulfill the ASTM requirements, except the hardness is unsignificantly less/decreased. (author)

  8. Gravity driven emergency core cooling experiments with the PACTEL facility

    International Nuclear Information System (INIS)

    Munther, R.; Kalli, H.; Kouhia, J.

    1996-01-01

    PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs

  9. Gravity driven emergency core cooling experiments with the PACTEL facility

    Energy Technology Data Exchange (ETDEWEB)

    Munther, R; Kalli, H [University of Technology, Lappeenranta (Finland); Kouhia, J [Technical Research Centre of Finland, Lappeenranta (Finland)

    1996-12-01

    PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs.

  10. Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2005-06-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

  11. Measurement of tritium activity in the aluminum pipe of JRR-2 heavy water primary cooling system using imaging plate

    International Nuclear Information System (INIS)

    Motoishi, Shoji; Kobayashi, Katsutoshi

    2000-12-01

    JRR-2 is the heavy water cooling type nuclear reactor, which has been operated for 36 years (1960-1976) and in the process of decommissioning at present. For this reason, evaluation of tritium quantity permeated into the pipe and apparatus of the primary coolant heavy water circulating system is important. In the Radioisotope Production Division, activity of tritium in aluminum pipe was measured with imaging plate (IP), liquid scintillation analyzer and high purity germanium detector (HPGe). After acrylic paints was applied for the region except for tritium contamination on the surface of aluminum pipe, only the oxidized contaminated part was dissolved by 1.5%(1.21M) HF for 3 minutes, and measured with IP. As a result, the tritium was found to permeate in the depth of 25 μm. Moreover, 90% of it was found to be distributed within 7 μm. (author)

  12. Method for evaluating the system instrumentation for loose part detection in the primary cooling circuit of French PWRs

    International Nuclear Information System (INIS)

    Gerardin, J.P.; Donnette, J.E.

    1995-05-01

    The purpose of the loose part detection system is to trigger an alarm whenever it is warranted, to localize, and to provide information on the type of loose part involved and the damages it may provoke. It is therefore indispensable to have efficient instrumentation, beginning with the sensors which must provide us with a response to all mechanical impacts in natural trapping areas (reactor vessel and steam generator water box). A series of mass- and energy-calibrated impacts have been generated on 45 points in the primary cooling system of a nuclear plant unit in the startup phase. This test provided insights into the relationship between sensor signals and various impact parameters such as velocity of impact or loose part mass. Once these parameters were known, it was possible to define a method for evaluating the detection threshold of sensors depending on the way they are mounted. (author)

  13. Thermal-hydraulic study of integrated steam generator in PWR

    International Nuclear Information System (INIS)

    Osakabe, Masahiro

    1989-01-01

    One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. (author)

  14. Development of technique for estimating primary cooling system break diameter in predicting nuclear emergency event sequence

    International Nuclear Information System (INIS)

    Tatebe, Yasumasa; Yoshida, Yoshitaka

    2012-01-01

    If an emergency event occurs in a nuclear power plant, appropriate action is selected and taken in accordance with the plant status, which changes from time to time, in order to prevent escalation and mitigate the event consequences. It is thus important to predict the event sequence and identify the plant behavior resulting from the action taken. In predicting the event sequence during a loss-of-coolant accident (LOCA), it is necessary to estimate break diameter. The conventional method for this estimation is time-consuming, since it involves multiple sensitivity analyses to determine the break diameter that is consistent with the plant behavior. To speed up the process of predicting the nuclear emergency event sequence, a new break diameter estimation technique that is applicable to pressurized water reactors was developed in this study. This technique enables the estimation of break diameter using the plant data sent from the safety parameter display system (SPDS), with focus on the depressurization rate in the reactor cooling system (RCS) during LOCA. The results of LOCA analysis, performed by varying the break diameter using the MAAP4 and RELAP5/MOD3.2 codes, confirmed that the RCS depressurization rate could be expressed by the log linear function of break diameter, except in the case of a small leak, in which RCS depressurization is affected by the coolant charging system and the high-pressure injection system. A correlation equation for break diameter estimation was developed from this function and tested for accuracy. Testing verified that the correlation equation could estimate break diameter accurately within an error of approximately 16%, even if the leak increases gradually, changing the plant status. (author)

  15. PWR type reactor plant

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1993-01-01

    A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

  16. French PWR safety philosophy

    International Nuclear Information System (INIS)

    Conte, M.

    1986-05-01

    Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach, each of them having possibilities and limits. As a consequence of the global risk objective set in 1977 for nuclear reactors, safety analysis was extended to the evaluation of events more complex than the conventional ones, and later to the evaluation of the feasibility of the offsite emergency plans in case of severe accidents

  17. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  18. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    Silliman, P.L.

    1978-01-01

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations

  19. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  20. Modelling the transport of radionuclides released in the Ilha Grande bay (Brazil) after a Large Break Loca ion the primary system of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Soares, Abner Duarte; Lapa, Celso Marcelo Franklin, E-mail: flamego@ien.gov.b, E-mail: asoares@cnen.gov.b, E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear (LIMA/IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    It was postulated, in the cooling system of the core, a LOCA, where 431 m{sup 3} of soda almost instantaneously was lost. This inventory contained 1.87x10{sup 10} Bq/m{sup 3} of tritium, 2.22x10{sup 7} Bq/m{sup 3} of cobalt,3.48x10{sup 8} Bq/m{sup 3} of cesium and 3.44x10{sup 10} Bq/m{sup 3} of iodine and was released in liquid form near the Itaorna cove, Angra dos Reis - RJ. Applying the model in the proposed scenario (Angra 1 and 2 in operation and Angra 3 progressively reducing the capture and discharge after the accident), the simulated dilution of the specific activity of radionuclide spots, reached values much lower than report levels for seawater (1,1x10{sup 6} Bq/m{sup 3}, 1,11x10{sup 4} Bq/m{sup 3} and 1,85x10{sup 3} Bq/m{sup 3}) after 22 hours, respectively for {sup 3}H, {sup 60}Co, {sup 131}I and {sup 137}Cs. From the standpoint of public exposure to radionuclide dispersion, the results of activity concentration obtained by the model suggest that the observed radiological impact is negligible. Based on these findings, we conclude that there would be no radiological impact related to a further release of controlled effluent discharges into Itaorna cove. (author)

  1. Modelling the transport of radionuclides released in the Ilha Grande bay (Brazil) after a Large Break Loca ion the primary system of a PWR

    International Nuclear Information System (INIS)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Soares, Abner Duarte; Lapa, Celso Marcelo Franklin

    2011-01-01

    It was postulated, in the cooling system of the core, a LOCA, where 431 m 3 of soda almost instantaneously was lost. This inventory contained 1.87x10 10 Bq/m 3 of tritium, 2.22x10 7 Bq/m 3 of cobalt,3.48x10 8 Bq/m 3 of cesium and 3.44x10 10 Bq/m 3 of iodine and was released in liquid form near the Itaorna cove, Angra dos Reis - RJ. Applying the model in the proposed scenario (Angra 1 and 2 in operation and Angra 3 progressively reducing the capture and discharge after the accident), the simulated dilution of the specific activity of radionuclide spots, reached values much lower than report levels for seawater (1,1x10 6 Bq/m 3 , 1,11x10 4 Bq/m 3 and 1,85x10 3 Bq/m 3 ) after 22 hours, respectively for 3 H, 60 Co, 131 I and 137 Cs. From the standpoint of public exposure to radionuclide dispersion, the results of activity concentration obtained by the model suggest that the observed radiological impact is negligible. Based on these findings, we conclude that there would be no radiological impact related to a further release of controlled effluent discharges into Itaorna cove. (author)

  2. Monte Carlo simulation of the electron and X-ray depth distribution for quantitative electron probe microanalysis of PWR spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Electron probe microanalysis requires several corrections to quantify an element of a specimen. The X-rays produced by the primary beam are created at some depth in the specimen. This distribution is usually represented as the function {Phi}(pz), and it is possible to calculate the correction factors for atomic number and absorption effects. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to quantify some elements of the PWR spent fuel with 50 GWd/tU of burnup and 2 years of cooling time

  3. Monte Carlo simulation of the electron and X-ray depth distribution for quantitative electron probe microanalysis of PWR spent fuels

    International Nuclear Information System (INIS)

    Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum

    2011-01-01

    Electron probe microanalysis requires several corrections to quantify an element of a specimen. The X-rays produced by the primary beam are created at some depth in the specimen. This distribution is usually represented as the function Φ(pz), and it is possible to calculate the correction factors for atomic number and absorption effects. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to quantify some elements of the PWR spent fuel with 50 GWd/tU of burnup and 2 years of cooling time

  4. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    Henshaw, Jim; McGurk, John; Dickinson, Shirley; Burrows, Robert; Hinds, Kelvin; Hussey, Dennis; Deshon, Jeff; Barrios Figueras, Joan Pau; Maldonado Sanchez, Santiago; Fernandez Lillo, Enrique; Garbett, Keith

    2012-09-01

    The activation and transport of corrosion products around a PWR circuit is a major concern to PWR plant operators as these may give rise to high personnel doses. The understanding of what controls dose rates on ex-core surfaces and shutdown releases has improved over the years but still several questions remain unanswered. For example the relative importance of particle and soluble deposition in the core to activity levels in the plant is not clear. Wide plant to plant and cycle to cycle variations are noted with no apparent explanations why such variations are observed. Over the past few years this group have been developing models to simulate corrosion product transport around a PWR circuit. These models form the basis for the latest version of the BOA code and simulate the movement of Fe and Ni around the primary circuit. Part of this development is to include the activation and subsequent transport of radioactive species around the circuit and this paper describes some initial modelling work in this area. A simple model of activation, release and deposition is described and then applied to explain the plant behaviour at Sizewell B and Vandellos II. This model accounts for activation in the core, soluble and particulate activity movement around the circuit and for activity capture ex-core on both the inner and outer oxides. The model gives a reasonable comparison with plant observations and highlights what controls activity transport in these plants and importantly what factors can be ignored. (authors)

  5. Generic study on the relation between contamination if primary coolants and occupational radiation exposure in nuclear power plants with PWR. Final report; Generische Studie zum Zusammenhang zwischen Kontamination von Primaerkreislaufmedien und beruflicher Strahlenexposition bei Kernkraftwerken mit Druckwasserreaktor. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Artmann, Andreas; Bruhn, Gerd; Schneider, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany); Strub, Erik [Koeln Univ. (Germany)

    2016-01-15

    A generic model for the primary cooling system contamination in pressurized water reactors and the resulting radiological consequences has been developed. The functional capability was demonstrated by means of three examples concerning manipulation procedures during revision outages. Activities at the main reactor coolant pumps were studied and the influence of the coolant contamination on the resulting dose rates and collective doses were calculated. The effect of a Co-90 hot spot in a more remote area on the radiation exposure during the specific action at the reactor pumps was considered.

  6. Failure of PWR-RHRS under cold shutdown conditions: Experimental results from the PKL test facility

    International Nuclear Information System (INIS)

    Mandl, R.M.; Umminger, K.J.; Logt, J.V.D.

    1991-01-01

    The Residual Heat Removal System (RHRS) of a PWR is designed to transfer thermal energy from the core after plant shutdown and maintain the plant in cold shutdown or refuelling conditions for extended periods of time. Initial reactor cooling after shutdown is achieved by dissipating heat through the steam generators (SGs) and discharging steam to the condenser by means of the Turbine Bypass System (TBS). When the reactor coolant temperature has dropped to about 160C and pressure has been reduced to 30 bar the RHRS is placed into operation. it reduces the coolant temperature to 50C within 20 hours after shutdown. The time margin for establishing alternate methods of heat removal following a failure of the RHRS depends on the Reactor Coolant System (RCS) temperature, the decay heat rate and the amount of RCS inventory. During some shutdown operations the RCS may be partially drained (e. g. to perform SG inspections). Decreased primary system inventory can significantly reduce the time available to recover the RHRS's function prior to bulk boiling and possible core uncovery. In the PKL test facility, which simulates a 1,300 MWe 4-loop PWR on a scale 1:145, a failure of RHRS under cold shutdown conditions was performed. This presentation gives a brief description of the test facility followed by the test objectives and results of this experiment

  7. Natural circulation in a scaled PWR integral test facility

    International Nuclear Information System (INIS)

    Kiang, R.L.; Jeuck, P.R. III

    1987-01-01

    Natural circulation is an important mechanism for cooling a nuclear power plant under abnormal operating conditions. To study natural circulation, we modeled a type of pressurized water reactor (PWR) that incorporates once-through steam generators. We conducted tests of single-phase natural circulations, two-phase natural circulations, and a boiler condenser mode. Because of complex geometry, the natural circulations observed in this facility exhibit some phenomena not commonly seen in a simple thermosyphon loop

  8. PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Bell, M.J.; Blomgren, J.C.; Fackelmann, J.M.

    1982-10-01

    Steam generators in pressurized water reactor (PWR) nuclear power plants have experienced tubing degradation by a variety of corrosion-related mechanisms which depend directly on secondary water chemistry. As a result of this experience, the Steam Generator Owners Group and EPRI have sponsored a major program to provide solutions to PWR steam generator problems. This report, PWR Secondary Water Chemistry Guidelines, in addition to presenting justification for water chemistry control parameters, discusses available analytical methods, data management and surveillance, and the management philosophy required to successfully implement the guidelines

  9. Radioprotection and safety for a dry storage module for bare PWR fuel elements

    International Nuclear Information System (INIS)

    Tzontlimatzin, E.

    1983-01-01

    A module for dry storage of spent fuel from PWR, after a previous cooling time of 2 years, is examined. Biological protection is obtained by 185 cm of concrete. The safety study shows the impossibility of a fast increase in temperature in case of cooling system failure because in this case the module will be cooled by natural convection or thermosiphon. A project for a storage installation consisting of 5 modules for 1500 irradiated fuel assemblies is described [fr

  10. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

    1995-01-01

    The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

  11. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  12. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 9: PRAISE computer code user's manual. Final report

    International Nuclear Information System (INIS)

    Lim, E.Y.

    1981-08-01

    The PRAISE (Piping Reliability Analysis Including Seismic Events) computer code estimates the influence of earthquakes on the probability of failure at a weld joint in the primary coolant system of a pressurized water reactor. Failure, either a through-wall defect (leak) or a complete pipe severance (a large-LOCA), is assumed to be caused by fatigue crack growth of an as-fabricated interior surface circumferential defect. These defects are assumed to be two-dimensional and semi-elliptical in shape. The distribution of initial crack sizes is a function of crack depth and aspect ratio. Crack propagation rates are governed by a Paris-type relationship with separate RMS cyclic stress intensity factors for the depth and length. Both uniform through the wall and radial gradient thermal stresses are included in the calculation of the stress intensity factors. The failure probabilities are estimated by applying Monte Carlo methods to simulate the life histories of the selected weld joint. In order to maximize computational efficiency, a stratified sampling procedure is used to select the initial crack size. Hydrostatic proof test, pre-service inspection, and in-service inspection can be simulated. PRAISE treats the inter-arrival times of operating transients either as a constant or exponentially distributed according to observed or postulated rates. Leak rate and leak detection models are also included. The criterion for complete pipe severance is exceedance of a net section critical stress. Earthquakes of various intensity and arbitrary occurrence times can be modeled. PRAISE presently assumes that exactly one initial defect exists in the weld and that the earthquake of interest is the first earthquake experienced at the reactor

  13. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  14. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  15. Recent development in PWR zinc injection

    International Nuclear Information System (INIS)

    Ocken, H.; Fruzzetti, K.; Frattini, P.; Wood, C.J.

    2002-01-01

    Zinc injection to the reactor coolant system (RCS) of PWRs holds the promise to alleviate two key challenges facing PWR plant operators: (1) reducing degradation of coolant system materials, including nickel-base alloy tubing and lower alloy penetrations due to stress corrosion cracking, and (2) lowering shutdown dose rates. Primary water stress corrosion cracking (PWSCC) is a dominant tube failure mode at many plants. This paper summarizes recent observations from U. S. and international PWRs that have implemented zinc injection, focusing primarily on coolant chemistry and dose rate issues. It also provides a look at the future direction of EPRI-sponsored projects on this topic. (authors)

  16. Contribution to study and design of PWR plant simulation code

    International Nuclear Information System (INIS)

    Delourme, Didier.

    1980-11-01

    This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants [fr

  17. Proposed method of the modeling and simulation of corrosion product behavior in the primary cooling system of fast breeder reactors

    International Nuclear Information System (INIS)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    2011-01-01

    Radioactive corrosion products (CP) are main cause of personal radiation exposure during maintenance without fuel failure in FBR plants. In order to establish the techniques of radiation dose estimation for worker in radiation-controlled area, Program SYstem for Corrosion Hazard Evaluation code 'PSYCHE' has been developed. The PSYCHE is based on the Solution-Precipitation model. The CP transfer calculation using the Solution-Precipitation model needs a fitting factor for the calculation of the precipitation of CP. This fitting factor must be determined based on the measured values in reactors that have operating experience. For this reason, the inability to make accurate predictions for reactor without measured values is a major issue. In this study, in addition to existing Solution-Precipitation model in PSYCHE, a transfer-model of CP species in particle form was applied to calculations of CP behavior in the primary cooling system of fast breeder reactor MONJU. Based on the calculated results, we estimated the contribution of CP deposition in the particle-form. It was suggested that the improved model including transfer-model of CP species in particle-form could be used for evaluation of CP transfer and radiation-source distribution in place of conventional Solution-Precipitation model with fitting factor in the PSYCHE. Moreover, it was predicted that CP particles would tend to be deposited in region with high-flow rate of coolant. (author)

  18. Approximation for maximum pressure calculation in containment of PWR reactors

    International Nuclear Information System (INIS)

    Souza, A.L. de

    1989-01-01

    A correlation was developed to estimate the maximum pressure of dry containment of PWR following a Loss-of-Coolant Accident - LOCA. The expression proposed is a function of the total energy released to the containment by the primary circuit, of the free volume of the containment building and of the total surface are of the heat-conducting structures. The results show good agreement with those present in Final Safety Analysis Report - FSAR of several PWR's plants. The errors are in the order of ± 12%. (author) [pt

  19. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    Bernard, Patrice; Dupraz, Remy; Vasile, Alfredo.

    1979-11-01

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE [fr

  20. Leak before break application in French PWR plants under operation

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  1. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  2. Transient performance of flow in PWR reactor circuits

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.; Carajilescov, P.

    1988-12-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  3. Modeling on a PWR power conversion system with system program

    International Nuclear Information System (INIS)

    Gao Rui; Yang Yanhua; Lin Meng

    2007-01-01

    Based on the power conversion system of nuclear and conventional islands of Daya Bay Power Station, this paper models the thermal-hydraulic systems of primary and secondary loops for PWR by using the PWR best-estimate program-RELAP5. To simulate the full-scope power conversion system, not only the traditional basic system models of nuclear island, but also the major system models of conventional island are all considered and modeled. A comparison between the calculated results and the actual data of reactor demonstrates a fine match for Daya Bay Nuclear Power Station, and manifests the feasibility in simulating full-scope power conversion system of PWR by RELAP5 at the same time. (authors)

  4. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  5. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    International Nuclear Information System (INIS)

    Virpi Kouhia, V.; Purhonen, H.; Riikonen, V.; Puustinen, M.; Kyrki-Rajamaki, R.; Vihavainen, J.

    2012-01-01

    This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  6. Measurement of the residual stresses in a PWR Control Rod Drive Mechanism nozzle

    OpenAIRE

    Coules, Harry; Smith, David

    2018-01-01

    Residual stress in the welds that attach Control Rod Drive Mechanism nozzles into the upper head of a PWR reactor vessel can influence the vessel's structural integrity and initiate Primary Water Stress Corrosion Cracking. PWSCC at Alloy 600 CRDM nozzles has caused primary coolant leakage in operating PWRs. We have used Deep Hole Drilling to characterise residual stresses in a PWR vessel head. Measurements of the internal cladding and nozzle attachment weld showed that although modest tensile...

  7. French PWR Safety Philosophy

    International Nuclear Information System (INIS)

    Conte, M. M.

    1986-01-01

    The first 900 MWe units, built under the American Westinghouse licence and with reference to the U. S. regulation, were followed by 28 standardized units, C P1 and C P2 series. Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. As early as 1976, this experience was taken into account by French Safety organisms to discuss, with Electricite de France, the safety options for the planned 1300 MWe units, P4 and P4 series. In 1983, the new reactor scheduled, Ni4 series 1400 MWe, is a totally French design which satisfies the French regulations and other French standards and codes. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach each of them having possibilities and limits. Increasing knowledge and lessons learned from operating experience have contributed to the French safety philosophy improvement. The methodology now applied to safety evaluation develops a new facet of the in depth defense concept by taking highly unlikely events into consideration, by developing the search of safety consistency of the design, and by completing the deterministic approach by the probabilistic one

  8. Corrosion of high temperature alloys in the primary circuit helium of high temperature gas cooled reactors. Pt. 2

    International Nuclear Information System (INIS)

    Quadakkers, W.J.

    1985-01-01

    The reactive impurities H 2 O, CO, H 2 and CH 4 which are present in the primary coolant helium of high temperature gas-cooled reactors can cause scale formation, internal oxidation and carburization or decarburization of the high temperature structural alloys. In Part 1 of this contribution a theoretical model was presented, which allows the explanation and prediction of the observed corrosion effects. The model is based on a classical stability diagram for chromium, modified to account for deviations from equilibrium conditions caused by kinetic factors. In this paper it is shown how a stability diagram for a commercial alloy can be constructed and how this can be used to correlate the corrosion results with the main experimental parameters, temperature, gas and alloy composition. Using the theoretical model and the presented experimental results, conditions are derived under which a protective chromia based surface scale will be formed which prevents a rapid transfer of carbon between alloy and gas atmosphere. It is shown that this protective surface oxide can only be formed if the carbon monoxide pressure in the gas exceeds a critical value. Psub(CO), which depends on temperature and alloy composition. Additions of methane only have a limited effect provided that the methane/water ratio is not near to, or greater than, a critical value of around 100/1. The influence of minor alloying additions of strong oxide forming elements, commonly present in high temperature alloys, on the protective properties of the chromia surface scales and the kinetics of carbon transfer is illustrated. (orig.) [de

  9. Composition and concentration of soluble and particulate matter in the coolant of the reactor primary cooling system of the Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Garcia Rodenas, Luis; La Gamma, Ana M.; Villegas, Marina; Fernandez, Alberto N.; Allemandi, Walter; Manera, Raul; Rosales, Hugo

    2000-01-01

    Nuclear power plants type PWR and PHWR (pressurized water reactor and pressurized heavy water reactor) have three coolant circuits which only exchange energy among them. The primary circuit, whose coolant extracts the reactor energy, the secondary circuit or water-steam cycle and the tertiary circuit which could be lake, river or sea water. The chemistry of the primary and secondary coolants is carefully controlled with the aim of minimizing the corrosion of structural materials. However, very low rates of corrosion are inevitable and one of the consequences of the corrosion processes is the presence of soluble and particulate matter in the coolant from where several problems associated with mass transfer arisen. In this way radioactive nuclides are transported out of the core to the steam generators, hydraulic resistance increases and heat transfer capability degrades. In the present paper some alternative techniques are proposed for the quantification of both, the particulate and soluble matter present in the coolant and their correspondent composition. Some results are also included and discussed. (author)

  10. Transient Analysis for Evaluating the Potential Boiling in the High Elevation Emergency Cooling Units of PWR Following a Hypothetical Loss of Coolant Accident (LOCA) and Subsequent Water Hammer Due to Pump Restart

    International Nuclear Information System (INIS)

    Husaini, S. Mahmood; Qashu, Riyad K.

    2004-01-01

    The Generic Letter GL-96-06 issued by the U.S. Nuclear Regulatory Commission (NRC) required the utilities to evaluate the potential for voiding in their Containment Emergency Cooling Units (ECUs) due to a hypothetical Loss Of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB) accompanied by the Loss Of Offsite Power (LOOP). When the offsite power is restored, the Component Cooling Water (CCW) pumps restart causing water hammer to occur due to cavity closure. Recently EPRI (Electric Power Research Institute) performed a research study that recommended a methodology to mitigate the water hammer due to cavity closure. The EPRI methodology allows for the cushioning effects of hot steam and released air, which is not considered in the conventional water column separation analysis. The EPRI study was limited in scope to the evaluation of water hammer only and did not provide any guidance for evaluating the occurrence of boiling and the extent of voiding in the ECU piping. This paper presents a complete methodology based on first principles to evaluate the onset of boiling. Also, presented is a methodology for evaluating the extent of voiding and the water hammer resulting from cavity closure by using an existing generalized computer program that is based on the Method of Characteristics. The EPRI methodology is then used to mitigate the predicted water hammer. Thus it overcomes the inherent complications and difficulties involved in performing hand calculations for water hammer. The heat transfer analysis provides an alternative to the use of very cumbersome modeling in using CFD (computational fluid dynamics) based computer programs. (authors)

  11. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

    Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

  12. Investigation of primary cooling water chemistry following the partial meltdown of Pu-Be neutron source in Tehran Research Reactor Core (TRR)

    Energy Technology Data Exchange (ETDEWEB)

    Aghoyeh, Reza Gholizadeh [School of Research and Development of Nuclear Reactors and Accelerators, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Khalafi, Hossein, E-mail: hkhalafi@aeoi.org.i [School of Research and Development of Nuclear Reactors and Accelerators, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of)

    2011-03-15

    Research highlights: Effect of Pu-Be neutron source meltdown in core on reactor water chemistry. Water chemistry of primary cooling before, during and after of above incident was compared. Training importance. Management of nuclear incident and accident. - Abstract: Effect of Pu-Be neutron source meltdown in core on reactor water chemistry was main aim of this study. Leaving the neutron source in the core after reactor power exceeds a few hundred Watts was the main reason for its partial meltdown. Water chemistry of primary cooling before, during and after of above incident was compared. Activity of some radio-nuclides such as Ba-140, La-140, I-131, I-132, Te-132 and Xe-135 increased. Other radio-nuclides such as Nd-147, Xe-133, Sr-91, I-133 and I-135 are also detected which were not existed before this incident.

  13. Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

    Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)

  14. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR); Implementacion en software libre del simulador universitario de nucleoelectrica tipo PWR (SU-PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Hidago H, F.; Morales S, J.B. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

    2007-07-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  15. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    1982-04-01

    The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

  16. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  17. Advanced high conversion PWR: preliminary analysis

    International Nuclear Information System (INIS)

    Golfier, H.; Bellanger, V.; Bergeron, A.; Dolci, F.; Gastaldi, B.; Koberl, O.; Mignot, G.; Thevenot, C.

    2007-01-01

    In this paper, physical aspects of a HCPWR (High Conversion Light Water Reactor), which is an innovative PWR fuelled with mixed oxide and having a higher conversion ratio due to a lower moderation ratio. Moderation ratios lower than unity are considered which has led to low moderation PWR fuel assembly designs. The objectives of this parametric study are to define a feasibility area with regard to the following neutronic aspects: moderation ratio, Pu loading, reactor spectrum, irradiation time, and neutronic coefficients. Important thermohydraulic parameters are the pressure drop, the critical heat flux, the maximum temperature in the fuel rod and the pumping power. The thermohydraulic analysis shows that a range of moderation ratios from 0.8 to 1.2 is technically possible. A compromise between improved fuel utilization and research and development effort has been found for the moderation ration of about 1. The parametric study shows that there are 2 ranges of interest for the moderation ratio: -) moderation ratio between 0.8 and 1.2 with reduced fissile heights (> 3 m), hexagonal arrangement fuel assembly and square arrangement fuel assembly are possible; and -) moderation between 0.6 and 0.7 with a modification of the reactor operating conditions (reduction of the primary flow and of the thermal power), the fuel rods could be arranged inside a hexagonal fuel rod assembly. (A.C.)

  18. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  19. Experimental study on reflooding in advanced tight lattice PWR

    International Nuclear Information System (INIS)

    Hori, K.; Kodama, J.; Teramae, T.

    2000-01-01

    This paper is related to the experimental study on the feasibility of core cooling by re-flooding in a large break loss of coolant accident (LOCA) for the advanced tight lattice pressurized water reactor (PWR). The tight lattice core design should be adopted to improve the conversion ratio. Major one of the key questions of such tight lattice core is the cooling capability under the re-flood condition in a large break LOCA. Forced feed bottom re-flooding experiments have been performed by use of a 4x4 triangular array rod bundle. The rod gap is 0.5 mm, 1.0 mm, or 1.5 mm. The measured peak temperature is below around 1273 K even in case of 1.0/0.5 mm rod gap. And, the evaluation based on the experimental results of rod temperatures and core pressure drop also shows that the core cooling under re-flooding condition is feasible. (author)

  20. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

  1. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    International Nuclear Information System (INIS)

    Bandini, G.; Caira, M.; Naviglio, A.; Sorabella, L.

    1995-01-01

    The MARS nuclear plant is equipped with a 600 MWth PWR type nuclear steam supply system, with completely innovative engineered core safeguards. The most relevant innovative safety system of this plant is its Emergency Core Cooling System, which is completely passive (with only one non static component). The Emergency Core Cooling System (ECCS) of the MARS reactor is natural-circulation, passive-type, and its intervention follows a core flow decrease, whatever was the cause. The operation of the system is based on a cascade of three fluid systems, functionally interfacing through heat exchangers; the first fluid system is connected to the reactor vessel and the last one includes an atmospheric-pressure condenser, cooled by external air. The infinite thermal capacity of the final heat sink provides the system an unlimited autonomy. The capability and operability of the system are based on its integrity and on the integrity of the primary coolant boundary (both of them are permanently enclosed in a pressurized containment; 100% redundancy is also foreseen) and on the operation of only one non static component (a check valve), with 400% redundancy. In the paper, all main thermal hydraulic transients occurring as a consequence of postulated accidents are analysed, to verify the capability of the passive-type ECCS to intervene always in time, without causing undue conditions of reduced coolability of the core (DNB, etc.), and to verify its capability to guarantee a long-term (indefinite) coolability of the core without the need of any external intervention. (author)

  2. Prevention against fragile fracture in PWR pressure vessel in the presence of pressurized thermal shock

    International Nuclear Information System (INIS)

    Carmo, E.G.D. do; Oliveira, L.F.S. de; Roberty, N.C.

    1984-01-01

    A method for the determination of operational limit curves (primary pressure versus temperature) for PWR is presented. Such curves give the operators indications related to the safety status of the plant concerning the possibility of a pressurized thermal shock. The method begins by a thermal analysis for several postulated transients, followed by the determination of the thermomechanical stresses in the vessel and finally it makes use of the linear elasticity fracture mechanics. Curves are shown for a typical PWR. (Author) [pt

  3. Influences of boric acid and lithium hydroxide on oxide film of type 316 stainless steel in PWR simulated primary water; PWR 1次冷却材模擬環境中の316ステンレス鋼に生成した皮膜性状に及ぼすほう酸および水酸化リチウムの影響

    Energy Technology Data Exchange (ETDEWEB)

    Fukumura, Takuya; Fukuya, Koji; Arioka, Koji [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan)

    2012-06-15

    In order to understand the influences of boric acid and lithium hydroxide on the IGSCC of type 316 stainless steel, an oxide film was analyzed in simulated PWR primary water while varying the boric acid and lithium hydroxide concentrations. It was found that, although boric acid and lithium hydroxide did not affect the structure and chemical composition of the surface oxide film remarkably, a lower boric acid concentration or a higher lithium concentration produced an oxide film with a thicker surface. It was considered that the lower boric acid concentration and higher lithium hydroxide concentration caused a higher magnetite solubility at the surface of the material and that the higher magnetite solubility caused a higher iron concentration gradient, which promoted iron dissolution from the material and the formation of a thicker oxide film. It was found that the thicker oxide film caused a higher IGSCC susceptibility and that the corrosion was the dominant factor of the IGSCC mechanism. No significant change was found in the morphologies of crack tip oxide in different bulk water chemistry systems, thus producing CT specimens with similar crack growth rates. (author)

  4. Thermohydraulic behavior in a primary cooling system during a loss-of-coolant accident of a light-water reactor

    International Nuclear Information System (INIS)

    Shimamune, Hiroji; Shiba, Masayoshi; Adachi, Hiromichi; Suzuki, Norio; Okubo, Kaoru

    1975-12-01

    With ROSA-I (Rig of Safety Assessment - I), 61 runs of the LWR blowdown experiment have been carried out under the conditions: model reactor type, BWR and PWR; reactor core, none, no-heating and heating; rupture position, upper and lower pressure vessel nozzle; initial discharge pressure, 40, 70 and 100 kg/cm 2 G; and rupture diameter, 25, 50, 70, 100 and 125 mm. The purpose was to obtain the data of thermal and hydrodynamic behavior in the reactor pressure vessel during a blowdown, including in-vessel pressure, coolant temperature, discharge flow rate, model fuel rod surface temperature and shock wave. Analysis was also made with the codes RELAP-2 and -3 developed by NRTS of the United States, to verify the calculation model used. In addition, the results of calculation with the shockwave analysis code DEPCO developed in JAERI were compared with those by experiment. The experimental facility ROSA-I and the results obtained with it and also the analyses made in this connection, are described in detail. (auth.)

  5. Results of safety analysis on PWR type nuclear power plants with two and three loops

    International Nuclear Information System (INIS)

    1979-01-01

    The results of safety analysis on PWR type nuclear power plants with two and three loops are presented, which was conducted by the Resource and Energy Agency, in June, 1979. This analysis was made simulating the phenomenon relating to the pressurizer level gauge at the time of the TMI accident. The model plants were the Ikata nuclear power plant with two loops and the Takahama No. 1 nuclear power plant with three loops. The premise conditions for this safety analysis were as follows: 1) the main feed water flow is totally lost suddenly at the full power operation of the plants, and the feed water pump is started manually 15 minutes after the accident initiation, 2) the relief valve on the pressurizer is kept open even after the pressure drop in the primary cooling system, and the primary cooling water flows out into the containment vessel through the rupture disc of the pressurizer relief tank, and 3) the electric circuit, which sends out the signal of safety injection at the abnormal low pressure in the reactor vessel, is added from the view-point of starting the operation of the emergency core cooling system as early as possible. Relating to the analytical results, the pressure in the reactor vessels changes less, the water level in the pressurizers can be regulated, and the water level in the steam generators is recovered safely in both two and three-loop plants. It is recognized that the plants with both two- and three loops show the safe transient phenomena, and the integrity of the cores is kept under the premise conditions. The evaluation for each analyzed result was conducted in detail. (Nakai, Y.)

  6. A primary study on the increasing of efficiency in the computer cooling system by means of external air

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Kim, M. H. [Silla University, Busan (Korea, Republic of)

    2009-07-01

    In recent years, since the continuing increase in the capacity of in personal computer such as the optimal performance, high quality and high resolution image, the computer system's components produce large amounts of heat during operation. This study analyzes and investigates an ability and efficiency of the cooling system inside the computer by means of Central Processing Unit (CPU) and power supply cooling fan. This research was conducted for increasing an ability of the cooling system inside the computer by making a structure which produces different air pressures in an air inflow tube. Consequently, when temperatures of the CPU and room inside computer were compared with a general personal computer, temperatures of the tested CPU, the room and the heat sink were as low as 5 .deg. C, 2.5 .deg. C and 7 .deg. C respectively. In addition to, Revolution Per Minute (RPM) was shown as low as 250 after 1 hour operation. This research explored the possibility of enhancing the effective cooling of high-performance computer systems.

  7. Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Hsu, C.J.

    1983-01-01

    In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization

  8. ASN takes position in the in-service follow-up programs of primary and secondary loops of EdF's nuclear power plants

    International Nuclear Information System (INIS)

    2002-01-01

    This decision from the French authority of nuclear safety (ASN) aims at fixing the conditions to be respected by Electricite de France (EdF) during its in-service follow-up programs for the monitoring and preventive maintenance of the primary and secondary cooling loops of EdF's PWR reactors. The components and the particular points to be controlled are listed in appendixes. (J.S.)

  9. Safety Technology Research Program in the field of pressurized water reactors. 1. Technical report on advancement project RS 36/2. Emergency cooling program service life experiments: reflooding experiments involving the primary loop systems

    International Nuclear Information System (INIS)

    Schweickert, H.; Kremin, H.; Mandl, R.; Riedle, V.; Ruthrof, K.; Sarkar, J.; Schmidt, H.

    The reflooding of the hot reactor core is to be examined for a pressurized water reactor (PWR), using a model of the entire primary loop system. The scale of the model is to be 1:340 in cross-section, with the heights represented full-scale. In addition to the goals of the project, a description of the test facility, including data collection and control equipment is presented. The instrumentation, the planned test program and the test procedure are briefly set forth

  10. For sale: 7 AGR stations and a brand new PWR

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    Britain's seven AGR stations and the Sizewell B PWR will pass to private ownership under the UK government's plan to privatise the two nuclear generators, Nuclear Electric and Scottish Nuclear, sometime next year. Under the new set-up, the two generators will become operating subsidiaries of a holding company which will be headquartered in Scotland. The companies' ageing Magnox gas-cooled reactors will remain in a separate public sector company before being transferred to British Nuclear Fuels (BNFL) at the time of privatisation. (author)

  11. Emergency cooling process and device for nuclear reactor containment

    International Nuclear Information System (INIS)

    Costes, D.

    1985-01-01

    The emergency cooling system of a PWR containment, according to the principal patent, comprises a turbine fed by the humid air of the containment, a condenser in which the air flowing out of the turbine is dryed and cooled by an external coolant and a compressor actuated by the turbine and returning the dryed air in the containment [fr

  12. PWR-to-PWR fuel cycle model using dry process

    International Nuclear Information System (INIS)

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  13. A PWR reactor downcomer modification for reduction of ECC bypass flow during LOCA

    International Nuclear Information System (INIS)

    Popov, N.; Bosevski, T.

    1986-01-01

    The ECC bypass phenomenon in the PWR reactor down-comer, which delays the reactor vessel refilling, after cold leg large break LOCA accident, has been subject of analysis in this paper. In the paper, a particular construction modification of the reactor down-comer has been suggested by inserting vertical ribs, aimed to intensify the reactor ECC refilling following the LOCA accident, and to advance the thermal-hydraulics safety of post-accidental cooling of the PWR reactors. To verify the effectiveness of the suggested down-comer construction modification, some properly selected results, obtained by corresponding verified mathematical model, have been presented in this paper. (author)

  14. The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

    Science.gov (United States)

    Cockeram, B. V.; Kammenzind, B. F.

    2018-02-01

    Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

  15. PWR plant construction in Japan

    International Nuclear Information System (INIS)

    Tamura, Toshifumi

    2002-01-01

    The construction methods based on the experiences on the Nuclear Island, which is a critical path in the total construction schedule, have been studied and reconsidered in order to construct by more reliable and economical method. So various improved construction method are being applied and the duration of construction is being reduced continuously. So various improved construction method are being applied and the duration of construction is being reduced continuously. In this paper, the history of construction of twenty-three (23) PWR Plant, the actual construction methods and schedule of Ohi-3/4, to which the many improved methods were applied during their construction, are introduced mainly with the improved points for previously constructed plants. And also the situation of construction method for the next PWR Plant is simply explained

  16. A comparison of fuzzy logic-PID control strategies for PWR pressurizer control

    International Nuclear Information System (INIS)

    Kavaklioglu, K.; Ikonomopoulos, A.

    1993-01-01

    This paper describes the results obtained from a comparison performed between classical proportional-integral-derivative (PID) and fuzzy logic (FL) controlling the pressure in a pressurized water reactor (PWR). The two methodologies have been tested under various transient scenarios, and their performances are evaluated with respect to robustness and on-time response to external stimuli. One of the main concerns in the safe operation of PWR is the pressure control in the primary side of the system. In order to maintain the pressure in a PWR at the desired level, the pressurizer component equipped with sprayers, heaters, and safety relief valves is used. The control strategy in a Westinghouse PWR is implemented with a PID controller that initiates either the electric heaters or the sprayers, depending on the direction of the coolant pressure deviation from the setpoint

  17. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  18. PWR system reliability improvement activities

    International Nuclear Information System (INIS)

    Yoshikawa, Yuichiro

    1985-01-01

    In Japan lacking in energy resources, it is our basic energy policy to accelerate the development program of nuclear power, thereby reducing our dependence. As referred to in the foregoing, every effort has been exerted on our part to improve the PWR system reliability by dint of the so-called 'HOMEMADE' TQC activities, which is our brain-child as a result of applying to the energy industry the quality control philosophy developed in the field of manufacturing industry

  19. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  20. Cooling techniques

    International Nuclear Information System (INIS)

    Moeller, S.P.

    1994-01-01

    After an introduction to the general concepts of cooling of charged particle beams, some specific cooling methods are discussed, namely stochastic, electron and laser cooling. The treatment concentrates on the physical ideas of the cooling methods and only very crude derivations of cooling times are given. At the end three other proposed cooling schemes are briefly discussed. (orig.)

  1. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Young; Chang, M H; Yu, K J; Lee, D J; Cho, B H; Kim, H Y; Yoon, J H; Lee, Y J; Kim, J P; Park, C T; Seo, J K; Kang, H S; Kim, J I; Kim, Y W; Kim, Y H

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  2. Method of starting up PWR type reactor

    International Nuclear Information System (INIS)

    Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

    1988-01-01

    Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

  3. Conceptual study on advanced PWR system

    International Nuclear Information System (INIS)

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. 1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. 2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. 3) Control rod drive mechanism for fine control : type and function were surveyed. 4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. 5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. 6) Steam injector concepts: analysis and experiment were conducted. 7) Fluidic diode concepts : analysis and experiment were conducted. 8) Wet thermal insulator : tests for thin steel layers and assessment of materials. 9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs

  4. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  5. Results of water chemistry control in the in-pile ''Callisto'' loop (an experimental PWR rig installed in the BR2 reactor)

    International Nuclear Information System (INIS)

    Weber, M.; Benoit, P.; Dekeyser, J.; Verwimp, A.

    1994-01-01

    Since June 1992, a new experimental facility, called CALLISTO, is being irradiated in the BR2 materials testing reactor at Mol, Belgium. The main objective of the present test campaign is to study the behaviour of advanced fuel to high burn-up rates in a realistic PWR environment. Three in-pile sections, containing each 9 fuel rods, are loaded inside the reactor vessel and are connected to a common out-of-pile pressurized water circulation loop (ref.1). The later is branched-off into a purification circuit (feed-bleed concept) and further equipped with safety and auxiliary systems. To cope with the test programme, the equipments are designed so that the guidelines of a PWR primary water chemistry can be followed (ref.2). Real steady-state conditions cannot be observed because the typical BR2 cycle (3 weeks running/3 weeks shut-down) is much shorter and because the rig is cooled down during each reactor shut-down. The purpose of this poster is to provide results of chemical parameters recorded during the cycling behaviour of the CALLISTO primary water. (authors). 4 figs., 1 tab., 2 refs

  6. Material property changes of stainless steels under PWR irradiation

    International Nuclear Information System (INIS)

    Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

    2009-01-01

    Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

  7. Industrial assessment of nonbackfittable PWR design modifications. Final report

    International Nuclear Information System (INIS)

    Matzie, R.A.; Daleas, R.S.; Miller, D.D.

    1980-11-01

    As part of the US Department of Energy's Advanced Reactor Design Study, various nonbackfittable PWR design modifications were evaluated to determine their potential for improved uranium utilization and commercial viability. Combustion Engineering, Inc. contributed to this effort through participation in the Battelle Pacific Northwest Laboratory industrial assessment of such design modifications. Seven modifications, including the use of higher primary system temperatures and pressures, rapid-frequent refueling, end-of-cycle stretchout, core periphery modifications, radial blankets, low power density cores, and small PWR assemblies, were evaluated with respect to uranium utilization, economics, technical and operational complexity, and several other subjective considerations. Rapid-frequent refueling was judged to have the highest potential although it would probably not be economical for the majority of reactors with the design assumptions used in this assessment

  8. CAREM: an innovative-integrated PWR

    International Nuclear Information System (INIS)

    Mazzi, R.

    2005-01-01

    A promising future in view of the increasing worldwide acknowledgment of the Nuclear Power as a bulk-environmentally friendly energy source is envisaged; nevertheless the widespread concerns about nuclear safety means the uppermost challenge to the nuclear designers to achieve massive public acceptance of NPP. CAREM is an Argentine project aimed to achieve the development, design and construction of an innovative, small and integrated Nuclear Power Plant (NPP). The reactor has an indirect cycle with some distinctive features that greatly simplify the design, and also contributes to a higher safety level. Some of the design highlight are: integrated primary cooling system, primary cooling by natural circulation, self-pressurised primary system and safety systems relying on passive features. The innovative solutions are embraced in the 'CAREM Concept' aimed to enhance safety by using simpler and more reliable solutions to tackle major safety design challenges of the nuclear generation industry. The goal is achieved by drastically reducing the conceivable list of initiating events jointly with a large primary water/power ratio that results in spontaneous slow and mild transients even after most severe system or component failure. In addition all Safety Systems are also based on simple and reliable solutions that increase sharply the overall plant reliability at reduced costs. The concept has been engineering developed for the CAREM 25 (prototype, 100 MWth, 27 MWe) considered an appropriate size to display the performance related with the reactor core cooling and safety systems. This module while not cost effective if compared with major sized NPP's installation and operating cost, results appropriate for applications such as supplying domestic or industrial electricity and/or steam (i.e. for a water demineralising plant) at isolated or difficult to access, mid size, populations. A promising market is envisaged for the evolution of the CAREM Concept towards higher

  9. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  10. Electrochemistry of Water-Cooled Nuclear Reactors

    International Nuclear Information System (INIS)

    Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy; Amit Jain Han Sang Kim; Vishisht Gupta; Jonathan Pitt

    2006-01-01

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or ''radiation fields'' around the primary loop and the vessel, as a function of the operating parameters and the water chemistry

  11. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    Tabata, Hiroaki

    1996-01-01

    The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

  12. Possibilities for the reduction of the activity build-up in the primary circuit of water-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Sachse, G.; Mittag, I.

    1985-01-01

    Basing upon the international literature in a review are refered: research and development efforts; release, transport and deposition of radioactive corrosion products under primary circuit conditions; experimental results in test and technical systems; and possibilities to control radiation fields in nuclear power plants by waterchemical measures, decontamination, and high temperature filtration. Relevant patents are summarized. (author)

  13. Predictive analysis of the radiation exposure for the primary cooling system of the rated power operation of MONJU

    International Nuclear Information System (INIS)

    Matuo, Youichirou; Miyahara, Shinya; Hasegawa, Masanori; Maegawa, Yoshiharu

    2011-01-01

    Radioactive corrosion products (CP) are main source of personal radiation exposure during maintenance without fuel-failure accident in the Liquid-Metal Fast Breeder Reactor (LMFBR) plants. In order to establish the techniques of radiation dose estimation for personnel, program system 'DORE' has been developed. The DORE system is constructed by PSYCHE code and QAD code system. The density of each deposited CP of primary coolant system in MONJU was estimated by using the PSYCHE. Moreover, the QAD-CGGP2R code is applied to dose rate calculations for the primary coolant system in MONJU. The dose rate around primary piping system was visualized using AVS software. The predicted values were estimated to be saturated at 2-3 mSv/h in twenty years after the start of operation, and the dose rate reaches 4 mSv/h in domains near the IHX and the cold-leg piping. It has been assumed that the main radiation source is 54 Mn in the IHX, primary pump and cold-leg piping region. On the other hand, it was indicated that the contribution to dose rate of the 60 Co accounted for approximately 23% in the hot-leg piping region. (author)

  14. Surveillance of vibrations in PWR

    International Nuclear Information System (INIS)

    Espefaelt, R.; Lorenzen, J.; Aakerhielm, F.

    1980-07-01

    The core of a PWR - including fuel elements, internal structure, control rods and core support structure inside the pressure vessel - is subjected to forces which can cause vibrations. One sensitive means to detect and analyse such vibrations is by means of the noise from incore and excore neutron detector signals. In this project noise recordings have been made on two occasions in the Ringhals 2 plant and the obtained data been analysed using the Studsvik Noise Analysis Program System (SNAPS). The results have been intepreted and a detailed description of the vibrational status of the core and pressure vessel internals has been produced. On the basis of the obtained results it is proposed that neutron signal noise analysis should be performed at each PWR plant in the beginning, middle and end of each fuel cycle and an analysis be made using the methods developed in the project. It would also provide a contribution to a higher degree of preparedness for diagnostic tasks in case of unexpected and abnormal events. (author)

  15. Acceptance test for 900 MWe PWR unit replacement steam generators

    International Nuclear Information System (INIS)

    Gourguechon, B.

    1993-01-01

    During the first half of 1994, the Gravelines 1 steam generators will be replaced (SG replacement procedure). The new SG's differ from the former components notably by the alloy used for the tube bundle, in this case, the high chromium content Inconel 690. So, from this standpoint, they are to be considered as PWR 900 replacement SG first models and their thermal efficiency has consequently to be assessed. This will provide an opportunity of ensuring that the performance of the components delivered is in compliance with requirements and of making the necessary provisions if significant deviations are observed. The EFMT branch, which has been in charge of the instrumentation and acceptance of the different SG first models since the first PWR plants were commissioned, will be responsible for the acceptance tests and the ultimate validation of a performance assessment procedure applicable to the future replacement steam generators. The methods and tests proposed for SG expert appraisal are based on consideration of the importance of primary measurement quality for satisfactory SG assessment and of the new test facilities with which the 900 and 1 300 PWR plants are gradually being equipped. These facilities provide an on-site computer environment for tests compatible with the tools (PATTERN, etc.) used at EFMT and in other departments. This test is the first of this kind performed by EFMT and the test facility of a nuclear power plant. (author). 6 figs

  16. Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

    International Nuclear Information System (INIS)

    Muhammad Subekti

    2009-01-01

    Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

  17. Vibration behavior of PWR reactor internals Model experiments and analysis

    International Nuclear Information System (INIS)

    Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

    1975-01-01

    In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

  18. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design

    International Nuclear Information System (INIS)

    Fernandez, J. L.; Lopez, J.

    1986-01-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs

  19. Primary coolant chemistry of the Peach Bottom and Fort St. Vrain high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Burnette, R.D.; Baldwin, N.L.

    1980-11-01

    The chemical impurities in the primary coolants of the Peach Bottom and Fort St. Vrain reactors are discussed. The impurity mixtures in the two plants were quite different because the sources of the impurities were different. In the Peach Bottom reactor, the impurities were dominated by H 2 and CH 4 , which are decomposition products of oil. In the Fort St. Vrain reactor, there were high levels of CO, CO 2 , and H 2 O. Although oil ingress at Peach Bottom created carbon deposits on virtually all surfaces, its effect on reactor operation was negligible. Slow outgassing of water from the thermal insulation at Fort St. Vrain caused delays in reactor startup. The overall graphite oxidation in both plants was negligible

  20. Primary coolant chemistry of the Peach Bottom and Fort St. Vrain high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Burnette, R.D.; Baldwin, N.L.

    1981-01-01

    The chemical impurities in the primary coolants of the Peach Bottom and Fort St. Vrain reactors are discussed. The impurity mixtures in the two plants were quite different because the sources of the impurities were different. In the Peach Bottom reactor, the impurities were dominated by H 2 and CH 4 , which are decomposition products of oil. In the Fort St. Vrain reactor, there were high levels of CO, CO 2 , and H 2 O. Although oil ingress at Peach Bottom created carbon deposits on virtually all surfaces, its effect on reactor operation was negligible. Slow outgassing of water from the thermal insulation at Fort St. Vrain caused delays in reactor startup. The overall graphite oxidation in both plants was negligible. (author)

  1. Evaluation method of corrosive conditions in cooling systems of nuclear power plants by combined analyses of flow dynamics and corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Shunsuke [Nuclear Power Engineering Corporation (NUPEC), Tokyo (Japan); Atomic Energy Society of Japan (AESJ) (Japan). Research Committee on Water Chemistry Standard; Naitoh, Masanori [Nuclear Power Engineering Corporation (NUPEC), Tokyo (Japan); Atomic Energy Society of Japan (AESJ) (Japan). Computational Science and Engineering Div.; Uehara, Yasushi; Okada, Hidetoshi [Nuclear Power Engineering Corporation (NUPEC), Tokyo (Japan); Hotta, Koji [ITOCHU Techno-Solutions Corporation (Japan); Ichikawa, Ryoko [Mizuho Information and Research Inst., Inc. (Japan); Koshizuka, Seiichi [Tokyo Univ. (Japan)

    2007-03-15

    Problems in major components and structural materials in nuclear power plants have often been caused by flow induced vibration, corrosion and their overlapping effects. In order to establish safe and reliable plant operation, it is necessary to predict future problems for structural materials based on combined analyses of flow dynamics and corrosion and to mitigate them before they become serious issues for plant operation. The analysis models are divided into two types. 1. Prediction models for future problems with structural materials: Distributions of oxidant concentrations along flow paths are obtained by solving water radiolysis reactions in the boiling water reactor (BWR) primary cooling water and hydrazine-oxygen reactions in the pressurized water reactor (PWR) secondary cooling water. Then, the electrochemical corrosion potential (ECP) at the point of interest is also obtained by the mixed potential model using oxidant concentration. Higher ECP enhances the possibility of intergranular stress corrosion cracking (IGSCC) in the BWR primary system, while lower ECP enhances flow accelerated corrosion (FAC) in the PWR secondary system. 2. Evaluation models of wall thinning caused by flow accelerated corrosion: The degree of wall thinning is evaluated at a location with a higher possibility of FAC occurrence, and lifetime is estimated for preventive maintenance. General features of models are reviewed in this paper and the prediction models for oxidant concentrations are briefly introduced. (orig.)

  2. Evaluation method of corrosive conditions in cooling systems of nuclear power plants by combined analyses of flow dynamics and corrosion

    International Nuclear Information System (INIS)

    Uchida, Shunsuke; Hotta, Koji; Ichikawa, Ryoko; Koshizuka, Seiichi

    2007-01-01

    Problems in major components and structural materials in nuclear power plants have often been caused by flow induced vibration, corrosion and their overlapping effects. In order to establish safe and reliable plant operation, it is necessary to predict future problems for structural materials based on combined analyses of flow dynamics and corrosion and to mitigate them before they become serious issues for plant operation. The analysis models are divided into two types. 1. Prediction models for future problems with structural materials: Distributions of oxidant concentrations along flow paths are obtained by solving water radiolysis reactions in the boiling water reactor (BWR) primary cooling water and hydrazine-oxygen reactions in the pressurized water reactor (PWR) secondary cooling water. Then, the electrochemical corrosion potential (ECP) at the point of interest is also obtained by the mixed potential model using oxidant concentration. Higher ECP enhances the possibility of intergranular stress corrosion cracking (IGSCC) in the BWR primary system, while lower ECP enhances flow accelerated corrosion (FAC) in the PWR secondary system. 2. Evaluation models of wall thinning caused by flow accelerated corrosion: The degree of wall thinning is evaluated at a location with a higher possibility of FAC occurrence, and lifetime is estimated for preventive maintenance. General features of models are reviewed in this paper and the prediction models for oxidant concentrations are briefly introduced. (orig.)

  3. EDF program for cleaning of PWR high dosimetric NPPs results and perspectives of radiological cleanup engineering

    International Nuclear Information System (INIS)

    Ider, Samir

    2012-09-01

    In 2002 EDF's PWRs presented significant disparities in their integrated dose results. In order to treat the PWR with highest dose per shutdown ratio, EDF/DPN (Nuclear Production Division) decided to develop an engineering project for 'PWR high dosimetry cleaning'. A specific financial support and staff resource was allocated to each reactor on this project. On every reactor, equipment is decontaminated using chemical process. The average dose reduction is estimated at 0,5 Man.mSv per reactor treated over the five following years. During the power operation of a nuclear reactor, corrosion products from the primary circuit undergo activation through their circulation in the reactor core. These activated products can be integrated into the oxide layer that builds up during the power operation on the primary piping. Additionally, highly radioactive particles can fall by gravity to the bottom of the reactor vessel. Being highly mobile, these can be dispersed throughout the circuit system by the water movement during the shutdown of the Unit. These particles will become hot spots when trapped in specific features of the primary cooling circuit and its auxiliaries. This contamination of the primary circuit leads to high dosimetry of the personnel during the maintenance of the Unit. To decrease the dose received during the maintenance operations, curative actions must be taken to eliminate hot spots and reduce contamination, as well as preventatives measures to curb future new contamination. For each PWR concerned, the studies begin with a diagnostic of the origins of the contamination, the type of pollution, its form (mobile hot spots, oxide layers) and its location. Many types of measuring equipment are used to make a radiological diagnostic (portable gamma-ray spectroscopy, video cartography, and gamma camera system). Experience and knowledge of the underlying chemistry in decontamination process, leads to a Decontamination Factor (DF) per system for

  4. Methodology to monitor and diagnostic vibrations of the motor-pumps used in the primary cooling system of IEAR-1 nuclear research reactor

    International Nuclear Information System (INIS)

    Benevenuti, Erion de Lima

    2004-01-01

    The objectives of this study are to establish a strategy to monitor and diagnose vibrations of the motor pumps used in the primary reactor cooling system of the IEA-R1 nuclear research reactor, to verify the possibility of using the existing installed monitoring vibration system and to implement such strategy in a continuous way. Four types of mechanical problems were considered: unbalancing, misalignment, gaps and faults in bearings. An adequate set of analysis tools, well established by the industry, was selected. These are: global measurements of vibration, velocity spectrum and acceleration envelope spectrum. Three sources of data and information were used; the data measured from the primary pumps, experimental results obtained with a Spectra Quest machine used to simulate mechanical defects and data from the literature. The results show that, for the specific case of the motor-pumps of IEA-R1 nuclear research reactor, although the technique using the envelope of acceleration, which is not available in the current system used to monitor the vibration of the motor pumps, is the one with best performance, the other techniques available in the system are sufficient to monitor the four types of mechanical problems mentioned. The proposed strategy is shown and detailed in this work. (author)

  5. THALES, Thermohydraulic LOCA Analysis of BWR and PWR

    International Nuclear Information System (INIS)

    ABE, Kiyoharu

    1990-01-01

    reactor coolant system, combustible gas burning, atmosphere- structure heat transfer, ventilation, containment spray cooling, etc. After the molten core penetrates the reactor bottom head, steam generation, concrete disintegration and noncondensable gas generation are calculated in the reactor cavity or the pedestal. 2 - Method of solution: Each of the THALES member codes first establishes the steady state conditions after reading input data. Then iterative time-dependent calculation is continued, taking account of various phenomena and events and their interactions which will occur in the course of a postulated severe accident. The transient calculations are iterated by the physical times specified by input. Generally the RCS thermal hydraulic analysis with the THALES-PM or THALES-BM code is first carried out and its results are transferred to the following containment analysis with the THALES-CV code. Then both results are transferred to a code for analyzing fission product release and transport behavior. Automatic data transfer is possible in the case the JAERI's ART code is used for fission product behavior analysis. In overall thermal hydraulic analysis, a new method is adopted aiming at sufficiently accurate estimation of mixture levels in the reactor coolant system and the containment in a reasonable computer time. The heat transfer calculation in the core is carried out based on the backward method. 3 - Restrictions on the complexity of the problem: Restrictions relating to storage allocation are: (1) Maximum number of radial regions in the core : 10; (2) Maximum number of axial increments in the fuel rods : 50; (3) Maximum number of loops in the PWR primary system : 4; (4) Maximum number of volumes in the PWR primary system : 11; (5) Number of BWR recirculation loops: 2 (fixed); (6) Number of volumes in the BWR reactor coolant system : 7 (fixed); (7) Maximum number of compartments in the containment : 10. There is another restriction, which relates to time step

  6. PWR thermocouple mechanical sealing structure

    International Nuclear Information System (INIS)

    Shen Qiuping; He Youguang

    1991-08-01

    The PWR in-core temperature detection device, which is one of measures to insure reactor safety operation, is to monitor and diagnose reactor thermal power output and in-core power distribution. The temperature detection device system uses thermocouples as measuring elements with stainless steel protecting sleeves. The thermocouple has a limited service time and should be replaced after its service time has reached. A new sealing device for the thermocouples of reactor in-core temperature detection system has been developed to facilitate replacement. The structure is complete tight under high temperature and pressure without any leakage and seepage, and easy to be assembled or disassembled in radioactive environment. The device is designed to make it possible to replace the thermocouple one by one if necessary. This is a new, simple and practical structure

  7. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    Bacher, P.E.

    1987-01-01

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  8. Maintenance technologies for SCC of PWR

    International Nuclear Information System (INIS)

    Okimura, Koji; Hori, Nobuyuki; Kanzaki, Hiroshi; Tokuhisa, Kiichi; Kamo, Kazuhiko; Kurokawa, Masaaki

    2007-01-01

    The recent technologies of test, relaxation of deterioration, repairing and change of materials are explained for safe and stable operation of pressurized water reactor (PWR). Stress corrosion cracking (SCC) is originated by three factors such as materials, stress and environment. The eddy current test (ECT) method for the stream generator pipe and the ultrasonic test method for welding part of pipe were developed as the test technologies. Primary water stress corrosion cracking (PWSCC) of Inconel 600 in the welding part is explained. The shot peening of instrument in the gas, the water jet peening of it in water, and laser irradiation on the surface are illustrated as some examples of improvement technology of stress. The cladding of Inconel 690 on Inconel 600 is carried out under the condition of environmental cut. Total or some parts of the upper part of reactor, stream generator and structure in the reactor are changed by the improvement technologies. Changing Inconel 600 joint in the exit pipe of reactor with Inconel 690 is illustrated. (S.Y.)

  9. Beta and gamma dose calculations for PWR and BWR containments

    International Nuclear Information System (INIS)

    King, D.B.

    1989-07-01

    Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 x 10 8 rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 x 10 8 rad equipment qualification test region. 8 refs., 23 figs., 12 tabs

  10. Model for calculating the boron concentration in PWR type reactors

    International Nuclear Information System (INIS)

    Reis Martins Junior, L.L. dos; Vanni, E.A.

    1986-01-01

    A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

  11. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gorman, J.A.; Turner, A.P.L.

    1997-01-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses

  12. The N4 plant: culmination of French PWR experience

    International Nuclear Information System (INIS)

    Bellet, J.; Houyez, A.; Journet, J.; Pierrard, J.H.

    1985-01-01

    The model N4 series of 1400MWe class PWR plants has been developed in France from a unique base of technical and operating experience. It meets the French government's requirement for a reactor free of constraints due to licensing agreements with overseas companies, with enhanced safety features and incorporating the lessons of Three Mile Island. In particular, improvements have been made to the reactor vessel, the steam generators, the primary pumps and control systems. The units are capable of daily load following and extended operation between refuelling. The N4 plant includes a new design of turbine-generator. (author)

  13. Safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release

    International Nuclear Information System (INIS)

    Pointner, W.; Broecker, A.

    2012-01-01

    The report on safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release covers the following issues: assessment of the relevant status for PWR, evaluation of the national and international (USA, Canada, France) status, actualization of recommendations, transferability from PWR to BWR. Generic studies on the core cooling capability in case of insulation material release in BWR-type reactors were evaluated.

  14. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  15. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

    The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

  16. Role of electromagnetic filter in limitating radioactivity in the primary circuits of light water reactors

    International Nuclear Information System (INIS)

    Dolle, L.

    1978-01-01

    High temperature electromagnetic filtration of particulate corrosion products can be carried out with discharges up to 5% of the cooling flow rate. It allows efficient extraction of particulate matter which rate constants required for considerable reduction of activable crud deposition in the core. The paper holds a review of the preventing operation in the primary circuit of a PWR, and reports experimental results of efficiency measurments with an electromagnetic filter set in out-of-pile and in-pile pressurized water loops. The notable efficiencies towards radioactive fine grain and colloidal matter justify more extensive reactor scale application experiments. (author)

  17. Role of electromagnetic filter in limitating radioactivity in the primary circuits of light water reactors

    International Nuclear Information System (INIS)

    Dolle, L.

    1978-01-01

    High temperature electromagnetic filtration of particulate corrosion products can be carried out with discharges up to 5% of the cooling flow rate. It allows efficient extraction of particulate matter with rate constants required for considerable reduction of activable crud deposition in the core. The paper holds a review of the preventing operation in the primary circuit of a PWR, and reports experimental results of efficiency measurements with an electromagnetic filter set in out-of-pile and in-pile pressurized water loops. The notable efficiencies towards radioactive fine grain and colloidal matter justify more extensive reactor scale application experiments

  18. Cooling Tower Losses in Industry

    OpenAIRE

    Barhm Mohamad

    2017-01-01

    Cooling towers are a very important part of many chemical plants. The primary task of a cooling tower is to reject heat into the atmosphere. They represent a relatively inexpensive and dependable means of removing low-grade heat from cooling water. The make-up water source is used to replenish water lost to evaporation. Hot water from heat exchangers is sent to the cooling tower. The water exits the cooling tower and is sent back to the exchangers or to other units for further cooling.

  19. Recommendations of the MRP-139: Inspection of Welds dissimilar in Nozzles PWR reactor vessel in Spain

    International Nuclear Information System (INIS)

    Gadea, J. R.; Willke, A.; Regidor, J. J.; Tecnatom, S. A.

    2010-01-01

    The guide EPRI MRP-139, which provides the way forward for the inspection and evaluation of dissimilar butt welds, the primary system of PWR reactors, indicating the type of nondestructive testing to be done in these areas, based on discovered several cases of default in lnconel alloys 600 and 182 in American and European plants. The phenomenon of cracking.

  20. Experiments on natural circulation during PWR severe accidents and their analysis

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Stewart, W.A.; Sha, W.T.

    1988-01-01

    Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

  1. Application on electrochemistry measurement of high temperature high pressure condition in PWR nuclear power plants

    International Nuclear Information System (INIS)

    Li Yuchun; Xiao Zhongliang; Jiang Ya; Yu Xiaowei; Pang Feifei; Deng Fenfang; Gao Fan; Zhou Nianguang

    2011-01-01

    High temperature high pressure electrochemistry testing system was comprehensively analyzed in this paper, according to actual status for supervision in primary and secondary circuits of PWR nuclear power plants. Three research methods were reviewed and discussed for in-situ monitor system. By combination with ECP realtime measurement it was executed for evaluation and water chemistry optimization in nuclear power plants. It is pointed out that in-situ electrochemistry measurement has great potential application for water chemistry evaluation in PWR nuclear power plants. (authors)

  2. Study on quality control measures of static casting main pipe in PWR nuclear power plant

    International Nuclear Information System (INIS)

    Jiang Zhenbiao; Li Guanying; Liu Zhicheng

    2013-01-01

    This study analyzes the main reasons which impact the quality of primary pipe static casting elbows in PWR-M310 nuclear power plant. The quality control measures are developed from the election and inspection of material, improving sand production and casting process, improving lean management of personnel. The static casting defects of primary pipe elbows for Fuqing Unit 1 and 2 were down to less than 50% of the former project. The quality of static casting for the primary pipe elbows was significantly improved. Moreover, the implementation saves human resources and financing to repair casting defects, and also helps to win the delivery schedule. The quality control measures are good reference for improving primary pipe casting process. This study provides valuable experience for further study of improving the quality of static casting for the primary pipe of PWR nuclear power plant. (authors)

  3. Activity incorporation into zinc doped PWR oxides

    International Nuclear Information System (INIS)

    Maekelae, Kari

    1998-01-01

    Activity incorporation into the oxide layers of PWR primary circuit constructional materials has been studied in Halden since 1993. The first zinc injection tests showed that zinc addition resulted in thinner oxide layers on new metal surfaces and reduced further incorporation of activity into already existing oxides. These tests were continued to find out the effects of previous zinc additions on the pickup of activity onto the surface oxides which were subsequently exposed to zinc-free coolant. The results showed that previous zinc addition will continue to reduce the rate of Co-60 build-up on out-of-core surfaces in subsequent exposure to zinc-free coolants. However, the previous Zn free test was performed for a relatively short period of time and the water chemistry programme was continued to find out the long term effects for extended periods without zinc. The activity incorporation into the stainless steel oxides started to increase as soon as zinc dosing to the coolant was stopped. The Co-60 concentration was lowest on all of the coupons which were first oxidised in Zn containing primary coolant. After the zinc injection period the thickness of the oxides increased, but activity in the oxide films did not increase at the same rate. This could indicate that zinc in the oxide blocks the adsorption sites for Co-60 incorporation. The Co-60 incorporation rate into the oxides on Inconel 600 seemed to be linear whether the oxide was pre-oxidised with or without Zn. The results indicate that zinc can either replace or prevent cobalt transport in the oxides. The results show that for zinc injection to be effective it should be carried out continuously. Furthermore the actual mechanism by which Zn inhibits the activity incorporation into the oxides is still not clear. Therefore, additional work has to follow with specified materials to verify the conclusions drawn in this work. (author)

  4. Diagnostic technology of PWR plant equipment failures

    International Nuclear Information System (INIS)

    Nakamura, Tetsuo; Tanaka, Mamoru; Okamachi, Masao; Taguchi, Shozo; Nagashima, Kazuhiro; Ishikawa, Satoshi

    1985-01-01

    To confirm the soundness of the important facilities in a nuclear power plant contributes to the reliability of the plant operations and improvement of its operation rate. For this purpose, the following diagnostic techniques have been developed. (1). Vibration and loose parts monitoring: Detection of abnormal structural vibrations in the reactor, estimation of its mode, detection of loose parts in the primary system, and estimation of the position and energy of their collisions against the reactor vessel or the like. (2). Valve leak monitoring: Detection of leaks from primary valves in the primary cooling boundary, such as the pressurizer relief valve and safety valve, and estimation of the form of the leaks. (3). Detector noise response diagnosis: Diagnosis of degradation of principal process detectors during plant operation. Furthermore, a diagnostic system incorporating the above diagnostic technology applicable to actual plants has been experimentally manufactured and successfully verified. (author)

  5. The design of a compact integral medium size PWR

    International Nuclear Information System (INIS)

    Shirvan, Koroush; Hejzlar, Pavel; Kazimi, Mujid S.

    2012-01-01

    Highlights: ► We model the IRIS reactor in RELAP5 and VIPRE codes. ► We use Printed Circuit Heat Exchangers and internally and externally cooled fuel pins in IRIS. ► We increase the IRIS power by 50% and demonstrate adequate safety performance. ► We show significant potential gain in economics for any integral PWR reactor design. - Abstract: Integral reactors have been proposed in recent years as a means to eliminate loss of coolant events, and reduce the number of large vessels of a nuclear power plant. In this paper the focus on how to further increase the power that can be derived from a given vessel volume. The example is applied to the International Reactor Innovative and Secure (IRIS), a medium size, light water reactor rated at 1000 MWt. The IRIS is an integral design containing all pumps and steam generators along with a traditional PWR core inside the reactor vessel. IRIS was designed with 8 Once-Through Helically Coiled Steam Generators (OTHSG), located above the core, in an annular region between the riser and the pressure vessel wall. This work examines ideas to increase its power output in the same vessel size while maintaining or improving the safety margins. The combination of Printed Circuit Heat Exchangers (PCHE) and Internally and EXternally cooled Annular Fuel (IXAF) is proposed to implement such improvement in otherwise the reference IRIS design. Safety implications of such steam generator and fuel design changes for the same reactor size are examined, under both steady state and transients, using the RELAP5 and VIPRE codes. It is found that the IRIS reactor power can be increased by 50% by using the PCHE and IXAF. The proposed design is found to be less expensive per unit electric power produced, these improvements and analyses can be applied to any integral reactor design.

  6. Application of the complex equilibrium code QUIL to cesium-impurity equilibria in the primary coolant of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Feber, R.D.; Lunsford, J.L.; Stark, W.A. Jr.

    1976-05-01

    An equilibrium analysis has been made of the fission-product cesium in the primary coolant loop of the high-temperature gas-cooled reactor (HTGR). The species distributions that result at equilibrium have been calculated for various conditions of reactor operation. The cesium species considered were the monomer, dimer, oxides, hydroxides, and the hydride. The effect of cesium sorption isotherms on graphite also was included in the analysis. During normal reactor operations, the abundant species of cesium were calculated to be elemental cesium, Cs, and the monomeric hydroxide, CsOH. Under most conditions of steam ingress, the abundant species was calculated to be CsOH. Cesium adsorbed onto graphite was stable under all steam-ingress conditions considered. Thermal transients above 1500 0 K were required for equilibrium transport of cesium from the core to the coolant. The analysis was carried out using the complex equilibrium code QUIL, designed and written with special emphasis on features that make it applicable to the fission-product problem

  7. Power plant cooling systems: trends and challenges

    International Nuclear Information System (INIS)

    Rittenhouse, R.C.

    1979-01-01

    A novel design for an intake and discharge system at the Belle River plant is described followed by a general discussion of water intake screens and porous dikes for screening fish and zooplankton. The intake system for the San Onofre PWR plant is described and the state regulations controlling the use of water for power plants is discussed. The use of sewage effluent as a source of cooling water is mentioned with reference to the Palo Verde plant. Progress in dry cooling and a new wet/dry tower due to be installed at the San Juan plant towards the end of this year, complete the survey

  8. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  9. Upgrading of PWR plant simulators

    International Nuclear Information System (INIS)

    Wada, Tomonori; Sasaki, Kazunori; Nakaishi, Hirokazu.

    1989-01-01

    For the education and training of operators in electric power plants, simulators have been employed, and it is well known that their effect is great. There are operation training simulators which simulate the dynamic characteristics of plants and all the machinery and equipment that operators handle, and train the procedure of restoration at the time of abnormality in plants, education simulators which can analyze the dynamic characteristics of plants efficiently in a short time, and offer information by visualizing phenomena with three-dimensional display and others so as to be easily understandable, and forecast simulators which do the analysis forecasting plant behavior at the time of abnormality in plants, and investigate the necessity of the guide for operation procedure and the countermeasures at the time of emergency. In this explanation, the upgrading of operation training simulators which have been put already in training is discussed. The constitution of simulator system and the instructor function, the outline of PWR plant simulation models comprising thermal flow model, pump model, leak model and so on, the techniques of increasing simulator speed, and the example of analysis using the NUPAC code are reported. (K.I.)

  10. Cooling towers

    International Nuclear Information System (INIS)

    Boernke, F.

    1975-01-01

    The need for the use of cooling systems in power plant engineering is dealt with from the point of view of a non-polluting form of energy production. The various cooling system concepts up to the modern natural-draught cooling towers are illustrated by examples. (TK/AK) [de

  11. Study of colloidal particles behaviour in the PWR primary circuit conditions; Etude du comportement des particules colloidales dans les conditions physicochimiques du circuit primaire des reacteurs a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Barale, M

    2006-12-15

    EDF wants to understand, model and limit primary circuit contamination of Pressurized Water Reactors by colloidal particles resulting from corrosion. The electrostatic behaviour of representative oxide particles (cobalt ferrite, nickel ferrite and magnetite) has been studied in primary circuit conditions with the influence of boric acid and lithium hydroxide. The isoelectric point (IEP) and the point of zero charge (PZC) of particles, measured between 5 C and 320 C, exhibit a minimum towards 200 C. The thermodynamic constants of the protonation equilibrium of surface sites were calculated. When boric acid is added, zeta potential and IEP decrease because of borate ions sorption. On the contrary, there is not effect of lithium ions. The modelling of these results under conditions representative of primary circuit shows that these oxides exhibit a negative surface charge, explaining their sorption and adhesion behaviour. (author)

  12. Experimental research progress on passive safety systems of Chinese advanced PWR

    International Nuclear Information System (INIS)

    Xiao Zejun; Zhuo Wenbin; Zheng Hua; Chen Bingde; Zong Guifang; Jia Dounan

    2003-01-01

    TMI and Chernobyl accidents, having pronounced impact on nuclear industries, triggered the governments as well as interested institutions to devote much attention to the safety of nuclear power plant and public's requirements on nuclear power plant safety were also going to be stricter and stricter. It is obvious that safety level of an ordinary light water reactor is no longer satisfactory to these requirements. Recently, the safety authorities have recommended the implementation of passive system to improve the safety of nuclear reactors. Passive safety system is one of the main differences between Chinese advanced PWR and other conventional PWR. The working principle of passive safety system is to utilize the gravity, natural convection (natural circulation) and stored energy to implement the system's safety function. Reactors with passive safety systems are not only safer, but also more economical. The passive safety system of Chinese advanced PWR is composed of three independent systems, i.e. passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system. This paper is a summary of experimental research progress on passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system

  13. Analyses of plant behaviors at the secondary side depressurization during LOCA of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kawabe, Yasuharu; Tamaki, Tomohiko; Kohriyama, Tamio; Ohtani, Masanori [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    When high pressure injection systems failed during a small break loss-of-coolant-accident (LOCA) for a PWR, main steam relief valves are opened to operate accumulator systems. However, it is pointed out that the core can be exposed since so-called counter current flow limitation (CCFL) occurs in steam generator (SG) tubes. The possibility of the core exposure by CCFL in a PWR plant was evaluated. First, RELAP5/MOD2 code was modified to be able to calculate CCFL. And then the code was applied to evaluate a 4-loop PWR plant. The LOCA with a rupture 3 inches were analyzed with the following two cases: (1) Only the main steam relief valve of the loop with the rupture is opened. (2) all of the relief valves are opened. It is seen that the CCFL phenomenon occurs in the case (1), however, the core cooling was maintained by the accumulator systems that actuated during the core exposure. On the other hand, the core exposure by CCFL is not observed in the case (2). It is shown that core cooling is promoted by operation of main steam relief valves. (author)

  14. Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios

    International Nuclear Information System (INIS)

    Zhou, Shengcheng; Wu, Hongchun; Zheng, Youqi

    2018-01-01

    Highlights: •ADS reloading scheme is optimized to raise discharge burnup and lower reactivity loss. •ADS is flexible to be combined with various pyro-chemical reprocessing technologies. •ADS is flexible to transmute MAs from different spent PWR fuels. -- Abstract: A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order.

  15. Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

    International Nuclear Information System (INIS)

    Sayyaadi, Hoseyn; Sabzaligol, Tooraj

    2010-01-01

    A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

  16. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  17. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  18. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  19. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  20. Occupational dose for the water cooling system of the SEAFP project

    International Nuclear Information System (INIS)

    Sandri, S.; Di Pace, L.

    1996-01-01

    The Occupational Radiation Exposure (ORE) for the water primary cooling system (PCS) of the SEAFP was assessed taking into account the first wall/blanket section only. All the potential radiological sources were considered and the analysis was restricted to the most important source at the PCS, the activated corrosion products (ACP). The relevant dose rate was evaluated using the computer code MCNP. Comparison of the results with the respective values measured at the fission PWR plants made it possible to transfer the parameters relevant to the working activities to the SEAFP PCS. Maintenance and inspection were found to be the only working tasks applicable to the SEAFP circuit and the worker access was considered to be allowed 24 h after the plant shut down only. 12 refs., 11 tabs

  1. Analysis of accidental loss of pool coolant due to leakage in a PWR SFP

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-01-01

    Highlights: • Accidental loss of pool coolant due to leakage in a PWR SFP was studied using MAAP5. • The effect of emergency ventilation on the accident progression was investigated. • The effect of emergency injection on the accident progression was discussed. - Abstract: A large loss of pool coolant/water accident may be caused by extreme accidents such as the pool wall or bottom floor punctures due to a large aircraft strike. The safety of SFP under this circumstance is very important. Large amounts of radioactive materials would be easily released into the environment if a severe accident happened in the SFP, because the spent fuel pool (SFP) in a PWR nuclear power station (NPS) is often located in the fuel handing building outside the reactor containment. To gain insight into the loss of pool coolant accident progression for a pressurized water reactor (PWR) SFP, a computational model was established by using the Modular Accident Analysis Program (MAAP5). Important factors such as Zr oxidation by air, air natural circulation and thermal radiation were considered for partial and complete drainage accidents without mitigation measures. The calculation indicated that even if the residual water level was in the active fuel region, there was a chance to effectively remove the decay heat through axial heat conduction (if the pool cooling system failed) or steam cooling (if the pool cooling system was working). For sensitivity study, the effects of emergency ventilation and water injection on the accident progression were analyzed. The analysis showed that for the current configuration of high-density storage racks, it was difficult to cool the spent fuels by air natural circulation. Enlarging the space between the adjacent assemblies was a way of increasing air natural circulation flow rate and maintaining the coolability of SFP. Water injection to the bottom of the SFP helped to recover water inventory, quenching the high temperature assemblies to prevent

  2. Spray cooling

    International Nuclear Information System (INIS)

    Rollin, Philippe.

    1975-01-01

    Spray cooling - using water spraying in air - is surveyed as a possible system for make-up (peak clipping in open circuit) or major cooling (in closed circuit) of the cooling water of the condensers in thermal power plants. Indications are given on the experiments made in France and the systems recently developed in USA, questions relating to performance, cost and environmental effects of spray devices are then dealt with [fr

  3. Steam turbines for PWR stations

    International Nuclear Information System (INIS)

    Muscroft, J.

    1989-01-01

    The thermodynamic cycle requirements and mechanical design features applying to modern GEC 3000 rev/min steam turbines for pressurised water reactor power stations are reviewed. The most recent developments include machines of 630 MW and 985 MW output which are currently under construction. The importance of service experience with nuclear wet steam turbines associated with a variety of types of water cooled reactor and its relevance to the design of modern 3000 rev/min turbines for pressurised water reactor applications is emphasised. (author)

  4. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

    International Nuclear Information System (INIS)

    MAJI, A. K.; MARSHALL, B.

    2000-01-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation FR-om nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  5. Gas cooled reactors

    International Nuclear Information System (INIS)

    Kojima, Masayuki.

    1985-01-01

    Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

  6. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J L; Lopez, J

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  7. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL) [de

  8. Reliability of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Ribeiro, A.A.T.; Muniz, A.A.

    1978-12-01

    Results of the analysis of factors influencing the reliability of international nuclear power plants of the PWR type are presented. The reliability factor is estimated and the probability of its having lower values than a certain specified value is discussed. (Author) [pt

  9. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  10. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    Larsen, N.

    1987-03-01

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  11. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  12. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  13. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.

    1980-01-01

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.) [pt

  14. Survey of PWR water chemistry

    International Nuclear Information System (INIS)

    Gorman, J.

    1989-02-01

    This report surveys available information regarding primary and secondary water chemistries of pressurized water reactors (PWRs) and the impact of these water chemistries on reactor operation. The emphasis of the document is on aspects of water chemistry that affect the integrity of the primary pressure boundary and the radiation dose associated with maintenance and operation. The report provides an historical overview of the development of primary and secondary water chemistries, and describes practices currently being followed. Current problems and areas of research associated with water chemistry are described. Recommendations for further research are included. 183 refs., 9 figs., 19 tabs

  15. Design Development and Verification of a System Integrated Modular PWR

    International Nuclear Information System (INIS)

    Kim, S.-H.; Kim, K. K.; Chang, M. H.; Kang, C. S.; Park, G.-C.

    2002-01-01

    An advanced PWR with a rated thermal power of 330 MW has been developed at the Korea Atomic Energy Research Institute (KAERI) for a dual purpose: seawater desalination and electricity generation. The conceptual design of SMART ( System-Integrated Modular Advanced ReacTor) with a desalination system was already completed in March of 1999. The basic design for the integrated nuclear desalination system is currently underway and will be finished by March of 2002. The SMART co-generation plant with the MED seawater desalination process is designed to supply forty thousand (40,000) tons of fresh water per day and ninety (90) MW of electricity to an area with approximately a ten thousand (100,000) population or an industrialized complex. This paper describes advanced design features adopted in the SMART design and also introduces the design and engineering verification program. In the beginning stage of the SMART development, top-level requirements for safety and economics were imposed for the SMART design features. To meet the requirements, highly advanced design features enhancing the safety, reliability, performance, and operability are introduced in the SMART design. The SMART consists of proven KOFA (Korea Optimized Fuel Assembly), helical once-through steam generators, a self-controlled pressurizer, control element drive mechanisms, and main coolant pumps in a single pressure vessel. In order to enhance safety characteristics, innovative design features adopted in the SMART system are low core power density, large negative Moderator Temperature Coefficient (MTC), high natural circulation capability and integral arrangement to eliminate large break loss of coolant accident, etc. The progression of emergency situations into accidents is prevented with a number of advanced engineered safety features such as passive residual heat removal system, passive emergency core cooling system, safeguard vessel, and passive containment over-pressure protection. The preliminary

  16. Improvements to PWR type reactors

    International Nuclear Information System (INIS)

    Ailloud, Jean; Monteil, Marcel.

    1978-01-01

    Improvements to pressurized water nuclear reactors are described, where the core coolant, called primary fluid, flows under the effect of a circulating pump in a primary loop between a steam generator and a pressure vessel containing the reactor core. The steam generator includes a bundle of tubes through which flows the primary fluid which exchanges calories with a secondary fluid, generally water, entering the generator as a liquid and issuing from it as steam. After expansion in turbines and recovery in a condenser, this steam is returned to the inside of the generator. Each primary fluid circulating pump is powered by a back-pressure turbine located in parallel with the high pressure section of the main turbine and hence fed with steam taken directly from the steam generator or the main steam pipe outside it [fr

  17. Core cooling system for reactor

    International Nuclear Information System (INIS)

    Kondo, Ryoichi; Amada, Tatsuo.

    1976-01-01

    Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

  18. An integrated PWR for marine propulsion

    International Nuclear Information System (INIS)

    Letouze, A.; Marecaux, A.; Rollason, J.; Heap, S.; Foster, A.; Jewer, S.; Thompson, A. C.; Williams, A. M.; Beeley, P. A.

    2008-01-01

    Results from a design study for a nuclear propulsion plant utilising a small integrated PWR using many of the inherent safety features of the IRIS design. The design consists of a single pass, low enrichment core housed, together with all associated primary circuit components, within a reactor pressure vessel 10.3 m high and 4.1 m in diameter. Reactor physics calculations were conducted with the codes WIMS9a and MONK8b. The core design contains 21 fuel assemblies each containing 264 UO 2 fuel pins. Each fuel module has a cluster of 24 boron carbide control rods and a central instrumentation channel. The fuel enrichment was 9% in order to achieve the core lifetime requirement of 3000 EFPD at a reactor power of 120 MWth. This gives a discharge burnup of 51,000 MWd/t. To control excess reactivity, two forms of burnable poison are employed: a zirconium dibromide (ZrB 2 ) coating on the fuel compacts, and gadolinium oxide homogeneously mixed in the fuel. Thermal hydraulic calculations were performed using TRAC-P(ND) for steady-state operation and for a number of fault transients. The helical once through steam generators were modelled using heat structure and pipe components and their performance compared to independent calculations including heat transfer correlations for the helical coiled geometry. Intact circuit calculations for steady state were followed by a small break LOCA calculation including the effect of a containment volume which reproduced the gain of coolant effect reported for IRIS. It was demonstrated that the thermal limits were not exceeded for the identified key transients. The dynamic response of the reactor plant to typical power demands was modelled using AcslXtreme software. Several schemes for limiting the power overshoot that was found on rapid increase to full power were examined. It was concluded that the SG must be operated with variable secondary pressure and the best means of reducing power overshoot is to step back the throttle opening

  19. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Doyen, J.J.; Lebey, J.

    1979-01-01

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 280 0 C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 280 0 C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  20. Aqueous Boric acid injection facility of PWR type reactor

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi; Iwami, Masao.

    1996-01-01

    If a rupture should be caused in a secondary system of a PWR type reactor, pressure of a primary coolant recycling system is lowered, and a back flow check valve is opened in response to the lowering of the pressure. Then, low temperature aqueous boric acid in the lower portion of a pressurized tank is flown into the primary coolant recycling system based on the pressure difference, and the aqueous boric acid reaches the reactor core together with coolants to suppress reactivity. If the injection is continued, high temperature aqueous boric acid in the upper portion boils under a reduced pressure, further urges the low temperature aqueous boric acid in the lower portion by the steam pressure and injects the same to the primary system. The aqueous boric acid stream from the pressurized tank flowing by self evaporation of the high temperature aqueous boric acid itself is rectified by a rectifying device to prevent occurrence of vortex flow, and the steam is injected in a state of uniform stream. When the pressure in the pressurized tank is lowered, a bypass valve is opened to introduce the high pressure fluid of primary system into the pressurized tank to keep the pressure to a predetermined value. When the pressure in the pressurized tank is elevated to higher than the pressure of the primary system, a back flow check valve is opened, and high pressure aqueous boric acid is flown out of the pressurized tank to keep the pressure to a predetermined value. (N.H.)

  1. VAMCIS, a new measuring channel for continuous monitoring of leak rates inside PWR steam generators

    International Nuclear Information System (INIS)

    Champion, G.; Dubail, A.; Lefevre, F.

    1988-01-01

    In order to assess the primary to secondary leakage, radioactive isotopes, formed in the primary coolant as a result of fission or neutron capture, are usually monitored in the pressurized water reactor (PWR) secondary coolant. Conventional methods mainly based on the detection of 133 Xe, tritium, and 41 Ar are widely used on French Electricite de France (EdF) PWRs. Some years ago, it appeared necessary to improve both leak rate assessments and steam generator tube rupture (SGTR) detection. A volumetric activity measuring channel inside steam (VAMCIS) has been developed for this purpose. The SGTR that occurred at the North Anna PWR has focused the attention of safety authorities on this new measuring channel. It is planned to implement VAMCIS at North Anna in order to check the leak rate variations more accurately

  2. START - glass model of PWR

    International Nuclear Information System (INIS)

    Marn, J.; Ramsak, M.

    1998-01-01

    Recognizing the importance of nuclear engineering in the area of process engineering the University of Maribor, Faculty of Mechanical Engineering has invested in procuring and erecting glass model of pressurized water reactor. This paper deals with description of the model, its capabilities, and plans for its use within nuclear engineering community of Slovenia. The model, made primarily of glass, serves three purposes: educational, professional development and research. As an example, medium break loss of coolant accident is presented in the paper. Temperatures within primary and secondary side, and pressure on primary side of reactor coolant system are followed. The characteristic points are emphasized, and commented.(author)

  3. Electrochemical measurements in PWR steam generators to follow crevice chemistry

    International Nuclear Information System (INIS)

    Feron, D.

    1991-01-01

    In PWR steam generator crevices, the evolution of chemistry is important for the understanding of corrosion phenomena. Electrochemical measurements have been performed in high temperature simulated crevice environments in order to follow hideout processes and remedial actions (on-line addition of boric acid). Reported tests have been conducted with model boilers of AJAX facilities. Eccentric and concentric tube support plate crevices have been instrumented with platinum electrodes. Electrochemical measurements have been collected when model boiler was under nominal conditions (primary temperature: 335 deg C, secondary temperature: 280 deg C). They include Electrochemical Impedance Spectroscopy (EIS) and potential measurements: with EIS, sodium and boric acid hideouts have been detected and followed. Potential measurements have been performed in an attempt to measure crevice PH evolution

  4. Waste management aspects of entire PWR LOOP decontamination

    International Nuclear Information System (INIS)

    Murray, A.P.; Roesmer, J.

    1988-01-01

    The waste management parameters for decontamination of an entire PWR primary circuit have been determined for dilute alkaline-permanganate/citric acid (APCA), LOMI, ozone and cerium acid process variations. APCA processes generate the largest waste volumes; over 140 m 3 (5000 ft 3 ) in some cases. This represents a potential disposal cost of one million dollars. The cation regeneration column makes the greatest contribution to the disposal volume. In contrast, the LOMI process generates approximately half as much waste, but it is expected to contain relatively high metal concentrations (200-800 ppm). The ozone and cerium acid processes product the least waste, usually under 45 m 3 . These waste volume estimates represent considerable fractions of a utility's annual disposal volume. Consequently, improved waste processing technology is required, and several approaches are suggested

  5. Numerical regulation of a test facility of materials for PWR

    International Nuclear Information System (INIS)

    Zauq, M.H.

    1982-02-01

    The installation aims at testing materials used in nuclear power plants; tests consists in simulations of a design basis accident (failure of a primary circuit of a PWR type reactor) for a qualification of these materials. A description of the test installation, of the thermodynamic control, and of the control system is presented. The organisation of the software is then given: description of the sequence chaining monitor, operation, list and function of the programs. The analog information processing is also presented (data transmission). A real-time microcomputer and clock are used for this work. The microprocessor is the 6800 of MOTOROLA. The microcomputer used has been built around the MC 6800; its structure is described. The data acquisition include an analog data acquisition system and a numerical data acquisition system. Laboratory and on-site tests are finally presented [fr

  6. PWR-blowdown heat transfer separate effects program

    International Nuclear Information System (INIS)

    Thomas, D.G.

    1976-01-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results are obtained from the Thermal-Hydraulic Test Facility (THTF). Supporting experiments are carried out in several additional test loops - the Forced Convection Test Facility (FCTF), an air-water loop, a transient steam-water loop, and a low-temperature water mockup of the THTF heater rod bundle. The studies to date are described

  7. Transient analysis of multifailure conditions by using PWR plant simulator

    International Nuclear Information System (INIS)

    Morisaki, Hidetoshi; Yokobayashi, Masao.

    1984-11-01

    This report describes results of the analysis of abnormal transients caused by multifailures using a PWR plant simulator. The simulator is based on an existing 822MWe power plant with 3 loops, and designed to cover wide range of plant operation from cold shutdown to full power at the end of life. Various malfunctions to simulate abnormal conditions caused by equipment failures are provided. In this report, features of abnormal transients caused by concurrence of malfunctions are discussed. The abnormal conditions studied are leak of primary coolant, loss of charging and feedwater flows, and control systems failure. From the results, it was observed that transient responses caused by some of the malfunctions are almost same as the addition of behaviors caused by each single malfunction. Therefore, it can be said that kinds of malfunctions which are concurrent may be estimated from transient characteristics of each single malfunction. (author)

  8. Behaviour of organic iodides under pwr accident conditions

    International Nuclear Information System (INIS)

    Alm, M.

    1982-01-01

    Laboratory experiments were performed to study the behaviour of radioactive methyl iodide under PWR loss-of-coolant conditions. The pressure relief equipment consisted of an autoclave for simulating the primary circuit and of an expansion vessel for simulating the conditions after a rupture in the reactor coolant system. After pressure relief, the composition of the CH 3 sup(127/131)I-containing steam-air mixture within the expansion vessel was analysed at 80 0 C over a period of 42 days. On the basis of the values measured and of data taken from the literature, both qualitative and quantitative assessments have been made as to the behaviour of radioactive methyl iodide in the event of loss-of-coolant accidents. (author)

  9. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  10. Critical heat flux correlation analysis for PWR reactors with low mass flow

    International Nuclear Information System (INIS)

    Carajilescov, Pedro

    1996-01-01

    The major limit in the thermalhydraulic design of water cooled reactors consists in the occurrence of critical heat flux, which is verified by correlation of large range of validity. In the present work, the major design correlations were analyzed, through comparisons with experimental data, for utilization in PWR with low mass flux in the core. The results show that the EPRI correlation, with modifications, gives conservative results, from the safety point of view, with lower data spreading, being the most indicated for the reactor thermal design. (author)

  11. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  12. A device for emergency cooling visualization

    International Nuclear Information System (INIS)

    Rezende, Hugo Cesar; Ladeira, Luiz Carlos Duarte

    1995-01-01

    A test facility for rewetting experiments, Emergency Cooling Visualization Device, has been erected at CDTN, with the objective of Emergency Cooling visualization device performing visual observations of basic phenomena that occur during the reflood phase of a Loss of Coolant Accident (LOCA), in a Pressurised Water Reactor (PWR), utilizing annular test sections. It permits to film or photograph the advance of a wetting front and the flow and heat transfer conditions. Then it is possible to observe the heat transfer regions and flow zones: steam convection, fog cooling, film boiling, nucleate boiling and fluid convection. Finally, this facility is the first test facility, in the Thermohydraulics Laboratory of CDTN, that uses a indirectly heated fuel rod simulator. (author). 3 refs, 5 figs

  13. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    International Nuclear Information System (INIS)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

    2017-01-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  14. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  15. Fluid-structure coupled dynamic response of PWR core barrel during LOCA

    International Nuclear Information System (INIS)

    Lu, M.W.; Zhang, Y.G.; Shi, F.

    1991-01-01

    This paper is engaged in the Fluid-Structure Interaction LOCA analysis of the core barrel of PWR. The analysis is performed by a multipurpose computer code SANES. The FSI inside the pressure vessel is treated by a FEM code including some structural and acoustic elements. The transient in the primary loop is solved by a two-phase flow code. Both codes are coupled one another. Some interesting conclusions are drawn. (author)

  16. Safety aspects of the design of a PWR gaseous radwaste treatment system using hydrogen recombiners

    International Nuclear Information System (INIS)

    Glibert, R.; Nuyt, G.; Herin, S.; Fossion, P.

    1978-01-01

    PWR Gaseous radwaste treatment system is essential for the reduction of impact on environment of the nuclear power plants. Decay tank system has been used for the retention of the radioactive gaseous fission products generated in the primary coolant. The use of a system combining decay tanks and hydrogen recombiner units is described in this paper. Accent is put on the safety aspects of this gaseous radwaste treatment facilitystudied by BN for a Belgian Power Plant. (author)

  17. Thermal analysis of a one-element PWR spent fuel shipping cask

    International Nuclear Information System (INIS)

    Fields, S.R.

    1979-06-01

    The transient thermal behavior of a typical one-element PWR spent fuel shipping cask, following a hypothetical accident and fire, has been simulated. The objectives of the study were to determine the transient behavior of the cask and its spent fuel primary coolant through the pressure relief system and possible fuel pin clad failure due to overheating following loss of coolant. 15 figures, 7 tables

  18. Zinc injection in German PWR plants

    International Nuclear Information System (INIS)

    Streit, K.

    2004-01-01

    Operating experience acquired at PWR NNPs shows that zinc injection at low concentrations of 5 ppb is a very effective source term reduction measure. This method does not lead to any operating restrictions or other negative effects on plant systems and components. The nuclear industry has been very successful in reducing radiation exposures within the past two decades. Annual exposures could be significantly decreased and are now at a level of around 1 man-Sv per plant and year. This great success can mainly be attributed to the general commitment of plant operators to maintaining radiation exposures of workers in the controlled access area as low as reasonably achievable (ALARA principle). The ALARA principle, of course, also implies evaluation of the economic benefit of radiation protection measures. Radiation source term reduction has drawn increasing attention of plant operators in recent years. For the new PWRs cobalt-based alloys in the primary system have successively been eliminated already at the design and construction phase within the last decade. Use of wear-resistant cobalt-free substitute materials in combination with the general use of advanced alloys for the steam generator tubing of PWRs resulted in low values for the two most common sources of plant radiation fields, namely 58 Co and 60 Co. Investigations showed that the beneficial effect of zinc can be related to its high affinity for mixed spinel oxide phases, resulting in the following two basic effects: -Zinc is incorporated preferentially into the oxide layer on primary system surfaces and thus reduces pickup of 58 Co and 60 Co and - Zinc can displace cobalt isotopes from existing oxide layers. In German PWRs with Incoloy 800 steam generator tubing material (Ni-content -32%), the observed reductions correspond to a decrease in dose rates of around 10 to 15% per year and thus follow, as predicted, the half-life time of 60 Co. Overall reductions in high radiation areas are now in the range of

  19. PWR reactor vessel in-service-inspection according to RSEM

    International Nuclear Information System (INIS)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

    2006-01-01

    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high requirements of

  20. Simulation of nonlinear dynamics of a PWR core by an improved lumped formulation for fuel heat transfer

    International Nuclear Information System (INIS)

    Su, Jian; Cotta, Renato M.

    2000-01-01

    In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant. An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling. Transient temperature response of fuel, cladding and coolant is analysed. (author)

  1. Ventilative Cooling

    DEFF Research Database (Denmark)

    Heiselberg, Per Kvols; Kolokotroni, Maria

    This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state-of-the-art of ventil......This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state......-of-the-art of ventilative cooling potentials and limitations, its consideration in current energy performance regulations, available building components and control strategies and analysis methods and tools. In addition, the report provides twenty six examples of operational buildings using ventilative cooling ranging from...

  2. PWR blowdown heat transfer separate-effects program: thermal-hydraulic test facility experimental data report for test 104

    International Nuclear Information System (INIS)

    Leon, D.M.; White, M.D.; Moore, P.A.; Hedrick, R.A.

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 104, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in the PWR system. Test 104 was conducted to obtain CHF in bundle 1 under blowdown conditions. The primary purpose of this report is to make the reduced instrument responses during test 104 available

  3. Sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hokkyo, N; Inoue, K; Maeda, H

    1968-11-21

    In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

  4. The PWR spectral code GELS. Pt. 1

    International Nuclear Information System (INIS)

    Penndorf, K.; Schult, F.; Schulz, G.

    1976-01-01

    The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

  5. Fuel management optimization for a PWR

    International Nuclear Information System (INIS)

    Dumas, M.; Robeau, D.

    1981-04-01

    This study is aimed to optimize the refueling pattern of a PWR. Two methods are developed, they are based on a linearized form of the optimization problem. The first method determines a feasible solution in two steps; in the first one the original problem is replaced by a relaxed one which is solved by the Method of Approximation Programming. The second step is based on the Branch and Bound method to find the feasible solution closest to the solution obtained in the first step. The second method starts from a given refueling pattern and tries to improve this pattern by the calculation of the effects of 2 by 2, 3 by 3 and 4 by 4 permutations on the objective function. Numerical results are given for a typical PWR refueling using the two methods

  6. RSK-guidelines for PWR reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The RSK guidelines for PWA reactors of April 24, 1974, have been revised and amended in this edition. The RSK presents a summary of safety requirements to be observed in the design, construction, and operation of PWR reactors in the form of guidelines. From January 1979 onwards these guidelines will be the basis of siting and safety considerations for new PWR reactors, and newly built nuclear power plants will have to form these guidelines. They are not binding for existing nuclear power plants under construction or in operation. It will be a matter of individual discussion whether or not the guidelines will be applied in these plants. The main purpose of the guidelines is to facilitate discussion among RSK members and to give early information on necessary safety requirements. If the guidelines are observed by producers and operators, the RSK will make statements on individual projects at short notice. (orig./HP) [de

  7. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  8. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  9. EDF/CIDEN - ONECTRA: PWR decontamination

    International Nuclear Information System (INIS)

    Fayolle, P.; Orcel, H.; Wertz, L.

    2010-01-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  10. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    Neubauer, W.

    1979-07-01

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

  11. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    Stutzmann, A.

    1997-01-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  12. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  13. Shielding design for PWR in France

    International Nuclear Information System (INIS)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983

  14. Organization patterns of PWR power plants

    International Nuclear Information System (INIS)

    Leicman, J.

    1980-01-01

    Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)

  15. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    Bruel, Renata Nunes

    1997-01-01

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  16. T Plant removal of PWR Chiller Subsystem

    International Nuclear Information System (INIS)

    Dana, C.M.

    1994-01-01

    The PWR Pool Chiller System is not longer required for support of the Shippingport Blanket Fuel Assemblies Storage. The Engineering Work Plan will provide the overall coordination of the documentation and physical changes to deactivate the unneeded subsystem. The physical removal of all energy sources for the Chiller equipment will be covered under a one time work plan. The documentation changes will be covered using approved Engineering Change Notices and Procedure Change Authorizations as needed

  17. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    Spanner, J.

    2015-01-01

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  18. First step of the project for implementation of two non-symmetric cooling loops modeled by the ALMOD3 code

    International Nuclear Information System (INIS)

    Dominguez, L.; Camargo, C.T.M.

    1984-09-01

    The first step of the project for implementation of two non-symmetric cooling loops modeled by the ALMOD3 computer code is presented. This step consists of the introduction of a simplified model for simulating the steam generator. This model is the GEVAP computer code, integrant part of LOOP code, which simulates the primary coolant circuit of PWR nuclear power plants during transients. The ALMOD3 computer code has a model for the steam generator, called UTSG, which is very detailed. This model has spatial dependence, correlations for 2-phase flow, distinguished correlations for different heat transfer process. The GEVAP model has thermal equilibrium between phases (gaseous and liquid homogeneous mixture), no spatial dependence and uses only one generalized correlation to treat several heat transfer processes. (Author) [pt

  19. PWR water chemistry controls: a perspective on industry initiatives and trends relative to operating experience and the EPRI PWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Choi, S.; Haas, C.; Pender, M.; Perkins, D.

    2010-01-01

    An effective PWR water chemistry control program must address the following goals: Minimize materials degradation (e.g., PWSCC, corrosion of fuel, corrosion damage of steam generator (SG) tubes); Maintain fuel integrity and good performance; Minimize corrosion product transport (e.g., transport and deposition on the fuel, transport into the SGs where it can foul tube surfaces and create crevice environments for the concentration of corrosive impurities); Minimize dose rates. Water chemistry control must be optimized to provide overall improvement considering the sometimes variant constraints of the goals listed above. New technologies are developed for continued mitigation of materials degradation, continued fuel integrity and good performance, continued reduction of corrosion product transport, and continued minimization of plant dose rates. The EPRI chemistry program, in coordination with other EPRI programs, strives to improve these areas through application of chemistry initiatives, focusing on these goals. This paper highlights the major initiatives and issues with respect to PWR primary and secondary system chemistry and outlines the recent, on-going, and proposed work to effectively address them. These initiatives are presented in light of recent operating experience, as derived from EPRI's PWR chemistry monitoring and assessment program, and EPRI's water chemistry guidelines. (author)

  20. Solubility of simulated PWR primary circuit corrosion products

    International Nuclear Information System (INIS)

    Kunig, R.H.; Sandler, Y.L.

    1986-08-01

    The solubility behavior of non-stoichiometric nickel ferrites, nickel-cobalt ferrites, and magnetite, as model substances for the corrosion products (''crud'') formed in nuclear pressurized water reactors, was studied in a flow system in aqueous solutions of lithium hydroxide, boric acid, and hydrogen with pH, temperature, and hydrogen concentrations as parameters. Below the temperature region of 300 to 330 0 C, at hydrogen concentrations of 25 to 40 cm 3 /kg H 2 O as used during reactor operation, the solubility of nickel-cobalt ferrite is the same as that of Ni and Co/sub x/Fe/sub 3-x/O 4 (x 3 /kg of hydrogen, the equilibrium iron and nickel solubilities increase congruently down to about 100 0 C, in a manner consistent with the solubility of Fe 3 O 4 , but sharply decline at lower temperatures, apparently due to formation of a borated layer. A cooldown experiment on a time scale of a typical Westinghouse reactor shutdown, as well as static experiments carried out on various ferrite samples at 60 0 C show that after addition of oxygen or peroxide evolution of nickel (and possibly cobalt) above the equilibrium solubility in hydrogen depends on the presence of dissociation products prior to oxidation. Thermodynamic calculations of various reduction and oxidative decomposition reactions for stoichiometric and non-stoichiometric nickel ferrite and cobalt ferrite are presented. Their significance to evolutions of nickel and cobalt on reactor shutdown is discussed. 30 refs., 38 figs., 34 tabs

  1. Cooling towers

    International Nuclear Information System (INIS)

    Korik, L.; Burger, R.

    1992-01-01

    What is the effect of 0.6C (1F) temperature rise across turbines, compressors, or evaporators? Enthalpy charts indicate for every 0.6C (1F) hotter water off the cooling tower will require an additional 2 1/2% more energy cost. Therefore, running 2.2C (4F) warmer due to substandard cooling towers could result in a 10% penalty for overcoming high heads and temperatures. If it costs $1,250,000.00 a year to operate the system, $125,000.00 is the energy penalty for hotter water. This paper investigates extra fuel costs involved in maintaining design electric production with cooling water 0.6C (1F) to 3C (5.5F) hotter than design. If design KWH cannot be maintained, paper will calculate dollar loss of saleable electricity. The presentation will conclude with examining the main causes of deficient cold water production. State-of-the-art upgrading and methodology available to retrofit existing cooling towers to optimize lower cooling water temperatures will be discussed

  2. PWR.2 - the unique transportation

    International Nuclear Information System (INIS)

    Howell, G.E.; Wills, L.I.

    1987-01-01

    Design studies of the prototype machinery and installation of same to be used for test and evaluation of a new design of nuclear power plant for submarines, showed that there were advantages if large units could be fitted out entirely at the manufacturer's base in Barrow-in-Furness. However, they had then to be shipped to the customer at the Vulcan Naval Reactor Test Establishment in Caithness, Scotland. The transportation of the loads involved is described. The main loads were the primary unit which weighed 1300 tonnes and the secondary unit which was transported as five separate assemblies, the largest two of which weighed 151 and 43 tonnes. Five basic transportation methods were used: skidding on PTFE pads, sea passage with barge on submersible barge, rolling on airbags, skating on water skates and lifting and rolling on multi-wheeled trailers. By careful planning the primary unit was moved in 16 days and the secondary unit in 19 days. The route and methods used are described and illustrated. (U.K.)

  3. Probabilistic analysis of 900 MWe PWR. Shutdown technical specifications

    International Nuclear Information System (INIS)

    Mattei, J.M.; Bars, G.

    1987-11-01

    During annual shutdown, preventive maintenance and modifications which are made on PWRs cause scheduled unavailabilities of equipment or systems which might harm the safety of the installation, in spite of the low level of decay heat during this period. The pumps in the auxiliary feedwater system, component cooling water system, service water system, the water injection arrays (LPIS, HPIS, CVCS), and the containment spray system may have scheduled unavailability, as well as the power supply of the electricity boards. The EDF utility is aware of the risks related to these situations for which accident procedures have been set up and hence has proposed limiting downtime for this equipment during the shutdown period, through technical specifications. The project defines the equipment required to ensure the functions important for safety during the various shutdown phases (criticality, water inventory, evacuation of decay heat, containment). In order to be able to judge the acceptability of these specifications, the IPSN, the technical support of the Service Central de Surete des Installations Nucleaires, has used probabilistic methodology to analyse the impact on the core melt probability of these specifications, for a French 900 MWe PWR

  4. The deformation of PWR fuel in a LOCA

    International Nuclear Information System (INIS)

    Mann, C.A.; Hindle, E.D.; Parsons, P.D.

    1982-04-01

    Available world-wide published data on the deformation of PWR fuel in a loss-of-coolant accident are reviewed. Adequate data exist for the oxidation of Zircaloy up to about 1500 0 C; data are increasingly sparse above this temperature and lacking above the melting point. The US NRC criteria for embrittlement are discussed and considered adequate for undeformed cladding, though they may be less so for deformed thinned material. Cladding deformation and the factors controlling it are considered in the light of data from the US, Germany, Japan and the UK. It is concluded that strains in the range 30% - 70% can be produced in experiments simulating LOCA conditions. The behaviour of cladding is strongly influenced by the spatial distribution of temperature, which is in turn dependent on heat transfer mechanisms at the surfaces of the cladding. No realistic experiment, i.e. one with a multirod array and simulated cooling, has produced deformations which would inhibit quenching. Such experiments have not, however, as yet covered the entire range of conditions which might obtain following a LOCA. (author)

  5. Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

    2015-07-01

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  6. Thermo-mechanical analysis of PWR bolts susceptible to IASCC

    International Nuclear Information System (INIS)

    Matteoli, C.; Hannink, M.H.C.; Blom, F.J.; Marck, S.C. van der; Charpin-Jacobs, F.

    2015-01-01

    Irradiation Assisted Stress Corrosion Cracking (IASCC) is considered a primary ageing issue for the Reactor Pressure Vessel (RPV) internals of Pressurized Water Reactors (PWR). In particular, this complex phenomenon which develops in an environment featuring thermal and mechanical stresses, interaction with corrosive compounds and irradiation, is affecting the bolts connecting the baffles and the formers in the Nuclear Power Plants' RPVs. The baffle-former assembly is the structure that borders the fuel assemblies region, contributing to keep them in position and separating in the radial direction, the core region from the downcomer region. An evaluation of the stresses and temperatures reached in the baffle-former bolts during normal operation was performed by means of a coupled thermo-mechanical study which uses reactor physics calculations to obtain the fluence in the reactor core and as a consequence the heat deposition in the RPV internals. The heat deposition data are coupled with a finite element model of the bolts and the RPV internals in order to perform a complete analysis taking in account thermal, mechanical and radiation loadings. The study is first carried out focusing on a section of the RPV internals, showing a single row of baffle-former bolts. Then the work is extended to the full core height. The model set up in this work, includes an in-depth study of the behavior of the core internals, in particular baffle-former bolts. The model has the capability of understanding the mechanical and thermal behavior of essential internal components in a PWR. (authors)

  7. Zircaloy-4 corrosion in PWR's

    International Nuclear Information System (INIS)

    Fyfitch, S.; Smalley, W.R.; Roberts, E.

    1985-01-01

    Zircaloy-4 waterside corrosion has been studied extensively in the nuclear industry for a number of years. Following the early crud-related corrosion failures in the Saxton test reactor, Westinghouse undertook numerous programs to minimize crud deposition on fuel rods in power reactors through primary coolant chemistry control. Modern plants today are operating with improved coolant chemistry guidelines, and crud deposition levels are very low in proportion to earlier experience. Zircaloy-4 corrosion under a variety of coolant chemistry, heat flux and exposure conditions has been studied extensively. Experience to date, even in relatively high coolant temperature plants, has indicated that -for both fuel cladding and structural components- Zircaloy-4 waterside corrosion performance has been excellent. Recognizing future industry trends, however, which will result in Zircaloy-4 being subjected to ever increasing corrosion duties, Westinghouse will continue accumulating Zircaloy-4 corrosion experience in large power plants. 13 refs.

  8. PWR type process heat reactor

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1974-01-01

    The nuclear reactor described is of the pressurized water type. It includes a prestressed concrete vessel, the upper part of which is shut by a closure, and a core surrounded by a core ring. The core fuel assemblies are supported by an initial set of vertical tubes integral with the bottom of the vessel, which serve to guide the rods of the control system. Over the core there is a second set of vertical tubes, able to receive the absorbing part of a control rod when this is raised above the core. An annular pressurizer around the core ring keeps the water in a liquid state. A pump is located above the second set of tubes and is integral with the closure. It circulates the water between the core and the intake of at least one primary heat exchanger, the exchanger (s) being placed between the wall of the vessel and the core ring [fr

  9. Cooling tower

    Energy Technology Data Exchange (ETDEWEB)

    Norbaeck, P; Heneby, H

    1976-01-22

    Cooling towers to be transported on road vehicles as a unit are not allowed to exceed certain dimensions. In order to improve the efficiency of such a cooling tower (of cross-flow design and box-type body) with given dimensions, it is proposed to arrange at least one of the scrubbing bodies displaceable within a module or box. Then it can be moved out of the casing into working position, thereby increasing the front surface available for the inlet of air (and with it the efficiency) by nearly a factor of two.

  10. Water chemistry control of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Hino, Yuichi; Makino, Ichiro; Yamauchi, Sumio; Fukuda, Fumihito.

    1992-01-01

    In PWR power plants, the primary system taking heat out of nuclear reactors and the secondary system generating steam and driving turbines are completely separated by steam generators, accordingly, by mutually independent water treatment, both systems are to be maintained in the optimal conditions. Namely, primary system is the closed water circulation circuit of simple liquid phase though under high temperature, high pressure condition, therefore, water shows the stable physical and chemical properties, and the minute water treatment for restraining the corrosion of structural materials and reducing radioactivity can be done. Secondary system is similar to the condensate and feedwater system of thermal power plants, and is the circuit for liquid-vapor two-phase transformation, but due to the local concentration of impurities by evaporation, the strict requirement is set for secondary water quality. However, secondary system can be treated in the state without radioactivity, and this is a great merit. The outline, basic concept and execution of primary water quality control, and the outline, concept, control criteria, facilities and execution of secondary water quality control are reported. (K.I.)

  11. Control in fabrication of PWR and BWR type reactor fuel elements

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1981-01-01

    Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

  12. Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3

    International Nuclear Information System (INIS)

    Preece, R.J.; Putney, J.M.

    1993-07-01

    A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser

  13. Stochastic cooling

    International Nuclear Information System (INIS)

    Bisognano, J.; Leemann, C.

    1982-03-01

    Stochastic cooling is the damping of betatron oscillations and momentum spread of a particle beam by a feedback system. In its simplest form, a pickup electrode detects the transverse positions or momenta of particles in a storage ring, and the signal produced is amplified and applied downstream to a kicker. The time delay of the cable and electronics is designed to match the transit time of particles along the arc of the storage ring between the pickup and kicker so that an individual particle receives the amplified version of the signal it produced at the pick-up. If there were only a single particle in the ring, it is obvious that betatron oscillations and momentum offset could be damped. However, in addition to its own signal, a particle receives signals from other beam particles. In the limit of an infinite number of particles, no damping could be achieved; we have Liouville's theorem with constant density of the phase space fluid. For a finite, albeit large number of particles, there remains a residue of the single particle damping which is of practical use in accumulating low phase space density beams of particles such as antiprotons. It was the realization of this fact that led to the invention of stochastic cooling by S. van der Meer in 1968. Since its conception, stochastic cooling has been the subject of much theoretical and experimental work. The earliest experiments were performed at the ISR in 1974, with the subsequent ICE studies firmly establishing the stochastic cooling technique. This work directly led to the design and construction of the Antiproton Accumulator at CERN and the beginnings of p anti p colliding beam physics at the SPS. Experiments in stochastic cooling have been performed at Fermilab in collaboration with LBL, and a design is currently under development for a anti p accumulator for the Tevatron

  14. The effects of cold rolling orientation and water chemistry on stress corrosion cracking behavior of 316L stainless steel in simulated PWR water environments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Junjie [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@t.shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Xiao, Qian; Ru, Xiangkun; Han, Guangdong; Chen, Zhen [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); Zhou, Bangxin [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Shoji, Tetsuo [New Industry Creation Hatchery Center, Tohoku University, Sendai 980-8579 (Japan)

    2016-04-15

    Stress corrosion cracking behaviors of one-directionally cold rolled 316L stainless steel specimens in T–L and L–T orientations were investigated in hydrogenated and deaerated PWR primary water environments at 310 °C. Transgranular cracking was observed during the in situ pre-cracking procedure and the crack growth rate was almost not affected by the specimen orientation. Locally intergranular stress corrosion cracks were found on the fracture surfaces of specimens in the hydrogenated PWR water. Extensive intergranular stress corrosion cracks were found on the fracture surfaces of specimens in deaerated PWR water. More extensive cracks were found in specimen T–L orientation with a higher crack growth rate than that in the specimen L–T orientation with a lower crack growth rate. Crack branching phenomenon found in specimen L–T orientation in deaerated PWR water was synergistically affected by the applied stress direction as well as the preferential oxidation path along the elongated grain boundaries, and the latter was dominant. - Highlights: • Transgranular fatigue crack growth rate was not affected by the cold rolling orientation. • Locally intergranular SCC was found in the hydrogenated PWR water. • Extensive intergranular SCC cracks were found in deaerated PWR water. • T–L specimen showed more extensive SCC cracks and a higher crack growth rate. • Crack branching related to the applied stress and the preferential oxidation path.

  15. Study of anticipated transient without scram for PWR

    International Nuclear Information System (INIS)

    Pu Jilong.

    1985-01-01

    Anticipated Transient Without Scram (ATWS) of PWR, the one of the 'Unresolved Safety Issue' with NRC, has been investigated for many years. The latest analysis done by the author considers the PWR's inherent stability and long-term performence under the condition of ATWS combined with SBLOCA and studies the sensitivity of several assumptions, which shows positive results

  16. Pushing back the boundaries of PWR fuel performance

    International Nuclear Information System (INIS)

    Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

    1985-01-01

    In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

  17. Highlights of the French program on PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pages, J P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1997-12-01

    The presentation reviews the French programme on PWR fuel including the overall results of the year 1996 for nuclear operation; fuel management and economy; French nuclear electricity generation sites; production of nuclear generated electricity; energy availability of the 900 and 1,300 Mw PWR units; average radioactive liquid releases excluding tritium per unit; plutonium recycling experience.

  18. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

  19. Chemical decontamination solutions: Effects on PWR equipment

    International Nuclear Information System (INIS)

    Pezze, C.M.; Colvin, E.R.; Aspden, R.G.

    1992-01-01

    A critical objective for the nuclear industry is the reduction of personnel exposure to radiation. Reductions have been achieved through industry's radiation management programs including training and radiation awareness concepts. Increased plant maintenance and higher radiation fields at many sites continue to raise concerns. To alleviate the radiation exposure problem, the sources of radiation which contribute to personnel exposure must be removed from the plant. A feasible was of significantly reducing these sources from a Pressurized Water Reactor (PWR) is to chemically decontaminate the entire reactor coolant system (RCS). A program was conducted to determine the technical acceptability of using certain dilute chemical solvent processes for full RCS chemical decontamination. The two processes evaluated were CAN-DEREM and LOMI. The purpose of the program was to define and complete a systematic evaluation of the major issues that need to be addressed for the successful decontamination of the entire RCS and affected portions of the auxiliary systems of a four-loop PWR system. A test program was designed to evaluate the corrosion effects of the two decontamination processes under expected plant conditions. Materials and sample configurations dictated by generic PWR components were evaluated. The testing also included many standard corrosion coupons. The test data were then used to assess the impact of chemical decontamination on the physical condition and operability of the components, equipment and mechanical systems that make up the RCS. An overview of the test program, sample configurations, data and engineering evaluations is presented. The data demonstrate that through detailed engineering evaluations of corrosion data and equipment function, the impact of full RCS chemical decontamination on plant equipment is established

  20. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    Sotoma, H.

    1973-01-01

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)