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Sample records for chinese 300mwe pwr

  1. Core loading pattern optimization of a typical two-loop 300 MWe PWR using Simulated Annealing (SA), novel crossover Genetic Algorithms (GA) and hybrid GA(SA) schemes

    International Nuclear Information System (INIS)

    Highlights: • SA and GA based optimization for loading pattern has been carried out. • The LEOPARD and MCRAC codes for a typical PWR have been used. • At high annealing rates, the SA shows premature convergence. • Then novel crossover and mutation operators are proposed in this work. • Genetic Algorithms exhibit stagnation for small population sizes. - Abstract: A comparative study of the Simulated Annealing and Genetic Algorithms based optimization of loading pattern with power profile flattening as the goal, has been carried out using the LEOPARD and MCRAC neutronic codes, for a typical 300 MWe PWR. At high annealing rates, Simulated Annealing exhibited tendency towards premature convergence while at low annealing rates, it failed to converge to global minimum. The new ‘batch composition preserving’ Genetic Algorithms with novel crossover and mutation operators are proposed in this work which, consistent with the earlier findings (Yamamoto, 1997), for small population size, require comparable computational effort to Simulated Annealing with medium annealing rates. However, Genetic Algorithms exhibit stagnation for small population size. A hybrid Genetic Algorithms (Simulated Annealing) scheme is proposed that utilizes inner Simulated Annealing layer for further evolution of population at stagnation point. The hybrid scheme has been found to escape stagnation in bcp Genetic Algorithms and converge to the global minima with about 51% more computational effort for small population sizes

  2. NOx emission and thermal efficiency of a 300 MWe utility boiler retrofitted by air staging

    Energy Technology Data Exchange (ETDEWEB)

    Li, S.; Xu, T.M.; Hui, S.; Wei, X.L. [Chinese Academy of Sciences, Beijing (China). Inst. of Mechanics

    2009-09-15

    Full-scale experiments were performed on a 300 MWe utility boiler retrofitted with air staging. In order to improve boiler thermal efficiency and to reduce NOx emission, the influencing factors including the overall excessive air ratio, the secondary air distribution pattern, the damper openings of CCOFA and SOFA, and pulverized coal fineness were investigated. Through comprehensive combustion adjustment, NOx emission decreased 182 ppm (NOx reduction efficiency was 44%), and boiler heat efficiency merely decreased 0.21%. After combustion improvement, high efficiency and low NOx emission was achieved in the utility coal-fired boiler retrofitted with air staging, and the unburned carbon in ash can maintain at a desired level where the utilization of fly-ash as byproducts was not influenced.

  3. Probabilistic safety assessment of French 900 and 1,300 MWe nuclear plants

    International Nuclear Information System (INIS)

    Although reactor design is mainly governed by deterministic principles in France, the probabilistic approach has been considered an important aid to safety analysis since the early seventies. Various partial probabilistic studies have been performed by Electricite de France, by IPSN and by Framatome, for various types of reactor. In particular, these studies have made it possible to assess the reliability and availability of nuclear power plants safety systems as well as the probability of accident scenarios and have helped to define technical specifications (especially, allowed operating times in the event of a partial unavailability of safety systems). Simultaneously, evaluation methods and corresponding software have been widely developed. Besides. EDF has implemented the Systeme de Recueil de Donnees de Fiabilite - SRDF (Reliability Data Collection System) which allows follow-up of equipment behaviour on all the operating units, and has led to a particularly representative data base. In 1982 the decision was taken at IPSN to carry out a complete PSA for a standard reactor of the 900 MWe type, and in 1986 EDF launched an equivalent study on a 1,300 MWe reactor, taking Unit 3 Paluel as reference. These PSAs were terminated in the course of the first quarter of 1990

  4. NO{sub x} emission and thermal efficiency of a 300 MWe utility boiler retrofitted by air staging

    Energy Technology Data Exchange (ETDEWEB)

    Li, Sen; Wei, Xiaolin [Institute of Mechanics, Chinese Academy of Sciences, No.15 Beisihuanxi Road, Beijing 100080 (China); Xu, Tongmo; Hui, Shien [State Key Laboratory of Multiphase Flow in Power Engineering, Xi' an Jiaotong University, 28 Xian Ning Road, Xi' an 710049 (China)

    2009-09-15

    Full-scale experiments were performed on a 300 MWe utility boiler retrofitted with air staging. In order to improve boiler thermal efficiency and to reduce NO{sub x} emission, the influencing factors including the overall excessive air ratio, the secondary air distribution pattern, the damper openings of CCOFA and SOFA, and pulverized coal fineness were investigated. Through comprehensive combustion adjustment, NO{sub x} emission decreased 182 ppm (NO{sub x} reduction efficiency was 44%), and boiler heat efficiency merely decreased 0.21%. After combustion improvement, high efficiency and low NO{sub x} emission was achieved in the utility coal-fired boiler retrofitted with air staging, and the unburned carbon in ash can maintain at a desired level where the utilization of fly-ash as byproducts was not influenced. (author)

  5. CFD investigation on the flow and combustion in a 300 MWe tangentially fired pulverized-coal furnace

    Science.gov (United States)

    Khaldi, Nawel; Chouari, Yoldoss; Mhiri, Hatem; Bournot, Philippe

    2016-09-01

    The characteristics of the flow, combustion and temperature in a 300 MWe tangentially fired pulverized-coal furnace are numerically studied using computational fluid dynamics. The mathematical model is based on a Eulerian description for the continuum phase and a Lagrangian description for coal particles. The combustion reaction scheme was modeled using eddy dissipation concept. The application of a proper turbulence model is mandatory to generate accurate predictions of flow and heat transfer during combustion. The current work presents a comparative study to identify the suitable turbulence model for tangentially fired furnace problem. Three turbulence models including the standard k-ɛ model, the RNG k-ɛ model and the Reynolds Stress model, RSM are examined. The predictions are compared with the published experimental data of Zheng et al. (Proc Combust Inst 29: 811-818, 2002). The RNG k-ɛ model proves to be the most suitable turbulence model, offering a satisfactory prediction of the velocity, temperature and species fields. The detailed results presented in this paper may enhance the understanding of complex flow patterns and combustion processes in tangentially fired pulverized-coal furnaces.

  6. The development and verification of thermal-hydraulic code on passive residual heat removal system of Chinese advanced PWR

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The technology of passive safety is the current trend among safety systems in nuclear power plant. Passive residual heat removal system (PRHRS), a major part of passive safety systems of Chinese advanced PWR, is a novel design with three-fold natural circulation. On the basis of reasonable physics and mathematics models, MITAP-PRHRS code was developed to analyze steady and transient characteristics of the PRHRS. The calculation and analysis show that the code simulates steady characteristics of the PRHRS very well, and it is able to simulate transient characteristics of all startup modes of the PRHRS. However, the quantitative description is poor during the initial stages of the transition process when water hammer occurs.

  7. Study on severe accident mitigation measures for the development of PWR SAMG

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In the development of the Severe Accident Management Guidelines (SAMG), it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR), Station Blackout (SBO), and Anticipated Transients without Scram (ATWS) in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management (SAM) such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.

  8. Study and Economics Analysis for 18-month Refueling Management on Power Uprate of a 300 MWe NPP%30万千瓦核电厂提升功率下的18个月换料燃料管理研究及经济性分析

    Institute of Scientific and Technical Information of China (English)

    曹泓; 王丽华

    2013-01-01

    鉴于近年来提升功率在大量现役核电厂中的成功应用,以及长周期、高燃耗、低泄漏的堆芯燃料管理技术所带来的更高的燃料利用率,本文对现役30万千瓦核电厂基于提升功率下开展了长周期、高燃耗、低泄漏的堆芯燃料管理方案研究.本文初步设计的燃料管理策略在满足有关的设计准则和要求基础上,可满足堆芯额定热功率为1250 MW的18个月长周期的堆芯燃料管理目标.与现行的堆芯燃料管理方案相比,可使核电厂在提升功率的同时进一步提高燃料利用率,经济性得到显著提高.本研究作为一项技术支持和技术储备,对30万千瓦级核电厂的功率提升和燃料管理方案优化具有较强的指导意义.%In recent years,power uprate is successfully applied in many nuclear power plants.Moreover,a longer cycle,higher uprate burnup and lower leakage fuel management strategy could enhance the fuel utilization.Therefore,the purpose of this article is to study a longer cycle,uprate burnup and lower leakage fuel management for a 300MWe NPP after power uprate.The results show that the concluded fuel management scheme for a 300MWe NPP after power uprate achieves the projected 18-month refueling cycle design objectives with the nominal thermal power of 1 250 MW and meets the design criteria.As compared to the current fuel management strategy of a 300MWe NPP,the advanced strategy in present study gains a power uprate,enhances the fuel utilization and improves the operation economy.As a technical support and reserve,the study will provide significant instructions on power uprate of a 300MWe NPP and optimization of fuel management strategy.

  9. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  10. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    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

  11. PWR type reactor

    International Nuclear Information System (INIS)

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  12. PWR degraded core analysis

    International Nuclear Information System (INIS)

    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.)

  13. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    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.)

  14. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    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)

  15. Physics of hydride fueled PWR

    Science.gov (United States)

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  16. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    The authors believe the next generation nuclear power plant should be characterized by: (1) simplicity of design; (2) ease of operation and maintenance; (3) economic conformance with safety requirements; and (4) technologies easy to understand by the public. 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, the Japan Atomic Power Company (JAPC) supported by the other Japanese PWR Utilities, Electricite de France (EdF), Westinghouse (WH) and Mitsubishi Heavy Industry (MHI) have studied application of passive technologies at a power rating of about 1,000 MWe. The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Using the AP-600 reference design as a basis, the authors enlarged the plant size to 3 -loops and added engineering features to conform with Japanese practice and Utilities' preference. The Simplified PWR (SPWR) program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1,000 MWe and 3 loops

  17. Activity transport models for PWR primary circuits

    International Nuclear Information System (INIS)

    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.)

  18. Program of monitoring PWR fuel in Spain

    International Nuclear Information System (INIS)

    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)

  19. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    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)

  20. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    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.)

  1. Adding a much needed 300 MWe at South Africa's Arnot coal fired power plant

    Energy Technology Data Exchange (ETDEWEB)

    Rich, G. [Alstom, Rugby (United Kingdom)

    2008-12-15

    As power stations built in the last thirty years approach the end of their design life, and the cost of new capacity continues to increase, along with demands for improved efficiency and lower emissions, an integrated approach to retrofit looks increasingly compelling. The ambitious upgrade project currently underway at the Arnot coal fired plant in South Africa, which will result in an update from 6 x 350 MWe to 6 x 400 MWe and a life extension of 20 years, illustrates the benefits. 2 figs.

  2. Guangdong: China imports French and British technology for a twin unit PWR station

    International Nuclear Information System (INIS)

    A series of articles on the Guangdong twin unit PWR station at Daya Bay in China. The Guangdong Nuclear Power Joint Venture Company (GNPJVC) is the joint Chinese-Hong Kong grouping responsible for owning and operating the Guangdong station. Contracts have been signed with Framatome, EdF and GEC Turbine Generators for the supply of two 900MWe PWRs with associated turbine generators and ancillary equipment, and relevant training and technology transfer. The meeting of site specific needs is discussed and a description of the plant given. A cutaway diagram together with technical specifications of the Guangdong plant is presented. (U.K.)

  3. Thorium fuel cycle study for PWR applications

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Jae Yong; Kim, Myung Hyun [Kyung Hee Univ., Seoul (Korea, Republic of)

    1997-12-31

    A nuclear design feasibility of thorium fueled high converting PWR was investigated. Two kinds of fuel design option were tested for the comparison with conventional UO{sub 2} fuel. The first one was an application of MHTGR pyro-carbon coated particle fuels. The other design was an application of MOX fuels as a ThO{sub 2}-PuO{sub 2} ceramic pellet. In the case of carbon-coated particle fuels, there was no benefit in nuclear design aspect because enrichment of U-235 was required over 5 w/o in order to match with the K-infinite of Ulchin-3/4 fuels. However, the use of thorium based plutonium fuels in PWR gave favorable aspects in nuclear design such as flatter K-infinite curve, lower M. T. C. and lower F. T. C. than that of UO{sub 2} fuel. (author). 6 refs., 3 tabs., 6 figs.

  4. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    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

  5. The integrated PWR; Les REP integres

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs

    2002-07-01

    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.)

  6. Shielding design for PWR in France

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

  9. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  10. Industrywide survey of PWR organics. Final report

    International Nuclear Information System (INIS)

    Thirteen Pressurized Water reactor (PWR) secondary cycles were sampled for organic acids, total organic carbon, and inorganic anions. The distribution and removal of organics in a makeup water treatment system were investigted at an additional plant. TOC analyses were used for the analysis of makeup water systems; anion ion chromatography and ion exclusion chromatography were used for the analysis of secondary water systems. Additional information on plant operation and water chemistry was collected in a survey. The analytical and survey data were compared and correlations made

  11. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    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)

  12. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    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 keff 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 keff markedly. The PWR keff 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

  13. Basic information about development and construction of a PWR

    International Nuclear Information System (INIS)

    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.)

  14. Study of safety relief valve operation under ATWS conditions. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Nesmith, B.J.; Brukiewa, J.B.

    1979-06-25

    A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

  15. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

    2012-07-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  16. Zebra: An advanced PWR lattice code

    International Nuclear Information System (INIS)

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  17. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  18. Crevice chemistry control in PWR steam generators

    International Nuclear Information System (INIS)

    To establish a basis for predicting and eventually controlling crevice solution chemistry in PWR steam generators, hideout tests were performed at several units. Results indicated that impurity hideout rates varied with the species and with bulk water concentration. Field evaluations of crevice impurity inventory models based on the hideout rate data indicated that further model refinements were necessary, e.g., more frequent quantification of the relation of hideout rates and bulk water concentration. An alternate crevice inventory model based on a real-time mass balance approach also began to be pursued. Modeling results currently are being used at several PWRs to establish a chloride injection rate consistent with development of a near neutral crevice solution to minimize IGA/SCC. Hideout return data are being used to independently establish predictions of crevice chemistry and to substantiate the hideout rate and mass balance model predictions

  19. The underclad cracking in PWR reactor vessels

    International Nuclear Information System (INIS)

    The article describes the kind of cracking which can occur under the stainless steel cladding during the manufacturing process of PWR vessels: - cold cracking recently found in France on vessel nozzles-reheat cracking discovered some ten years ago in particular in Germany and in USA. Methods of examination for underclad cracking are put forward, together with results obtained on vessel nozzles of units currently being built in Belgium. Some nozzles are affected by the phenomenon of reheat cracking, whilst the hypothesis of cold cracking, which had been proposed because of the similar situation found in France should probably be abandoned. On the basis of the investigations and studies made, it is established that the cracking involved does not jeopardize the integrity of the vessels during their life time. (author)

  20. The material analysis for PWR primary equipment

    International Nuclear Information System (INIS)

    The primary equipment in pressurized water reactor includes reactor pressure vessel, reactor coolant piping, steam generator, pressurizer, and reactor coolant pump casing, etc., which form the pressure boundary of the primary loop. These primary equipment are all pressure vessels of QA Class 1, Safety-related Class 1, and Aseismatic Category 1. Under high temperature, high pressure and neutron irradiation, the requirements for the base material and welding properties of these pressure vessels are very high, so as to ensure the long-term stable operation of nuclear power plant. The base material and welding properties of these pressure vessels are analyzed and discussed according to ASME B and P Code, which can be as a reference for base material selection of PWR pressure vessels. (authors)

  1. Characterization of Factors affecting IASCC of PWR Core Internals

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-09-15

    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  2. The PWR cores management; La gestion des coeurs REP

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Rippert, D. [CEA Cadarache, Departement d' Etudes des Reacteurs, DER, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    2000-01-25

    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.)

  3. Hydraulic benchmark data for PWR mixing vane grid

    International Nuclear Information System (INIS)

    The purpose of the present study is to present new hydraulic benchmark data obtained for PWR rod bundles for the purpose of benchmarking Computational Fluid Dynamics (CFD) models of the rod bundle. The flow field in a PWR fuel assembly downstream of structural grids which have mixing vane grids attached is very complex due to the geometry of the subchannel and the high axial component of the velocity field relative to the secondary flows which are used to enhance the heat transfer performance of the rod bundle. Westinghouse has a CFD methodology to model PWR rod bundles that was developed with prior benchmark test data. As improvements in testing techniques have become available, further PWR rod bundle testing is being performed to obtain advanced data which has high spatial and temporal resolution. This paper presents the advanced testing and benchmark data that has been obtained by Westinghouse through collaboration with Texas A&M University. (author)

  4. Hot Operation of FTL for PWR Fuels Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sung Ho; Joung, Chang Yong; Lee, Jong Min; Park, Su Ki; Sim, Bong Sik; Ahn, Guk Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Fuel Test Loop (FTL) in HANARO is the test facility which can conduct a fuel irradiation test with commercial NPPs' operating conditions such as their pressure, temperature, flow and water chemistry. The FTL is used for the irradiation test of PWR type or CNNDU type fuels. In this paper, the hot operation of FTL for irradiation test of PWR fuels is introduced. The experimental results show the excellence of operation performance

  5. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  6. Seawater desalination using reusable type small PWR

    Energy Technology Data Exchange (ETDEWEB)

    Uchiyama, Y. [Institute of Engineering Mechanics and Systems, University of Tsukuba, Tsukuba, Ibaraki (Japan); Minato, A. [Planning Division, Central Research Institute of the Electric Power Industry, Komae-shi, Tokyo (Japan); Shimamura, K. [Nuclear Systems Engineering Department, Nuclear Energy Systems Engineering Center, Mitsubishi Heavy Industries, Ltd., Kanagawa (Japan)]. E-mail: shimamura@atom.hq.mhi.co.jp

    2003-07-01

    Demand for seawater desalination is increasing, especially in regions such as the Middle East and North Africa, where populations are growing at a high annual rate. If such demand is met by fossil fuel energy, the influence on the environment, such as global warming, cannot be disregarded. Since these regions are behind in their preparedness of social capital infrastructure, such as power transfer grids, small reactors are considered to be more suitable for introduction than the large reactors found commonly in developed countries. Therefore, a small reusable PWR with mid-range pressure and temperature services, which does not require on-site refuelling, was devised for seawater desalination. In a small reusable PWR, spent fuel is taken out together with the reactor vessel and refuelled on the exterior fuel exchange base prepared independently. Thus, the safeguards against nuclear proliferation increase at a plant site because the lid of the reactor vessel is never opened at the site, in principle. The reactor vessel will be transported from the plant site to a fuel exchange base under stipulated conditions within a transportation cask after a long (about six years) operation. Since fuel handling facilities at the site become unnecessary through centralisation at a fuel exchange base, initial plant construction costs are reduced. In addition, the reactor vessel is reused until its service life has expired. This examination was based on the marine reactor of the experimental nuclear ship, Mutsu, after it had been applied for land use: at a lowered, midrange pressure and temperature service, in theory. It is possible to produce fresh water through reverse osmosis (RO) membrane pressure-rising seawater by a steam turbine driven pump. Using the method of driving a desalination unit high-pressure pump directly by low-pressure steam generated from the heating reactor, fresh water can be produced efficiently. Furthermore, operating at reduced pressure makes it possible

  7. The Study of Nuclear Fuel Cycle Options Based On PWR and CANDU Reactors

    International Nuclear Information System (INIS)

    The study of nuclear fuel cycle options based on PWR and CANDU type reactors have been carried out. There are 5 cycle options based on PWR and CANDU reactors, i.e.: PWR-OT, PWR-OT, PWR-MOX, CANDU-OT, DUPIC, and PWR-CANDU-OT options. While parameters which assessed in this study are fuel requirement, generating waste and plutonium from each cycle options. From the study found that the amount of fuel in the DUPIC option needs relatively small compared the other options. From the view of total radioactive waste generated from the cycles, PWR-MOX generate the smallest amount of waste, but produce twice of high level waste than DUPIC option. For total plutonium generated from the cycle, PWR-MOX option generates smallest quantity, but for fissile plutonium, DUPIC options produce the smallest one. It means that the DUPIC option has some benefits in plutonium consumption aspects. (author)

  8. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  9. Maintenance technologies for SCC of PWR

    International Nuclear Information System (INIS)

    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.)

  10. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

  11. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  12. 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.

  13. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10−6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  14. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14

  15. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  16. Signal processing methods for PWR reactor noise diagnostic system

    International Nuclear Information System (INIS)

    A framework for a PWR reactor noise diagnostic system using various signal processing methods has been investigated. Supposing to treat not only reactor noise data in a stationary linear system but also those in a nonstationary or nonlinear system, the study covers a third-order-correlation of bispectrum, cepstrum analysis, Group Method of Data Handling (GMDH), chaotic quantity, neural network, and wavelet, in addition to Multivariate AutoRegressive analysis and Signal Transmission Path Diagram analysis (MAR/STPD). This paper describes consideration about the methods from viewpoints of theories and applications to PWR reactor noise diagnostic system. The point at the issue in the application system is how to extract many characteristics from the signals whatever states (linear or nonlinear, stationary or nonstationary) may happen in order to get more information and more exact diagnose to support human judgment. From this viewpoint, the paper discusses several signal processing techniques for the PWR diagnostic system. (J.P.N.)

  17. Industry-wide survey of organics in PWR's

    International Nuclear Information System (INIS)

    Interest in organic impurities found in Pressurized Water Reactors (PWR's) has stemmed from several sources. The most serious concern is that organic acids will increase cation conductivity, a parameter that is used to control power plant chemistry. This effect can complicate secondary water monitoring and control. Organics may foul or exhaust makeup demineralizers and condensate polishers, and thus result in increased operating costs or the in leakage of potentially corrosive agents into the steam generators. Some organics, however, such as mopholine and cyclohexylamine may reduce corrosion through oxygen scavenging or surface filming reactions, and may have a positive influence on the pH in areas of local corrosion. At the time this survey began, little information was available on the types or levels of organic impurities that are typically found in PWR's. this survey is intended to provide baseline data for future corrosion testing and to provide fundamental information that will be helpful in refining PWR chemistry guidelines and operating practices

  18. Load-following operation of PWR plants

    International Nuclear Information System (INIS)

    The load-following operation of nuclear power plants will become inevitable due to the increased nuclear share in the total electricity generation. As a groundwork for the load-following capability of the Korean next generation PWRs, the state-of-the-art has been reviewed. The core control principles and methods are the main subject in this review as well as the impact of load-following operations on the fuel performance and on the mechanical integrity of components. To begin with, it was described what the load-following operation is and in what view point the technology should be reviewed. Afterwards the load-following method, performance and problems in domestic 900 MWe class PWRs were discussed, and domestic R and D works were summarized. Foreign technologies were also reviewed. They include Mode G and Mode X of Foratom, D and L bank method of KWU, the method using PSCEA of ABB-CE, and MSHIM of Westinghouse. The load-following related special features of Foratom's N4 plant, KWU's plants, ABB-CE's Systems 80+, and Westinghouse's AP600 were described in each technology review. The review concluded that the capability of N4 plant with Mode X is the best and the methods in System, 80+ and AP600 would require verifications for the continued and usual load-following operation. It was recommended that the load-following operation experiences in domestic PWRs under operation be required to settle down the capability for the future. In addition, a more enhanced technology is required for the Korean next generation PWR regardless what the reference plant concept is. 30 figs., 19 tabs., 75 refs. (Author)

  19. Borssele PWR noise: measurements, analysis and interpretation

    International Nuclear Information System (INIS)

    In the Borssele reactor - a 450 MWe PWR - reactor noise measurements have been performed during four fuel cycles. Measurements were made with a set of ex-core neutron detectors, on one occasion an in-core displacement transducer, and with primary coolant pressure sensors. Digital analysis was applied, where the most powerful tool was the computer programme FAST, which computes auto and cross power spectra for all combinations from a set of many simultaneously recorded signals. Analyses of neutronic signals show a reactivity noise peak at 9.2 Hz, core barrel motion peaks at about 12 and 15 Hz, a damped oscillation at about 2 Hz. Results are given for begin and end of each fuel cycle. The r.m.s. value of the low frequency noise appears to depend linearly on the boron concentration over a wide range. Also some results of primary coolant pressure noise are presented, with coherent peaks below 15 Hz and incoherent peaks above. The second part of the paper describes an alternative way of analyzing and interpreting noise spectra, namely attempts to decompose the neutronic power spectra into physical components, using the information present in the CPSD's of all detector combinations. The components are characterised by their detector position dependency. Effects considered are: uncorrelated noise, global reactivity noise, core motion attenuation noise, and a possible coupling between reactivity and core motion. Results show excellent separation into reactivity and core motion components with possibilities to separate overlapping peaks. Weak peaks become more easily detectable. At low frequencies the decomposition of the spectra is not yet complete, however. (author)

  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. The traveller: a new look for PWR fresh fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    Bayley, B.; Stilwell, W.E.; Kent, N.A. [Westinghouse Electric Co., Columbia, SC (United States)

    2004-07-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain.

  2. The traveller: a new look for PWR fresh fuel packages

    International Nuclear Information System (INIS)

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain

  3. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  4. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary

  5. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  6. Studies of a small PWR for onsite industrial power

    International Nuclear Information System (INIS)

    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application

  7. Studies of a small PWR for onsite industrial power

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, O.H.; Smith, W.R.

    1977-04-19

    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application.

  8. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    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)

  9. Experience and reliability of Framatome ANP's PWR and BWR fuel

    International Nuclear Information System (INIS)

    Based on three decades of fuel supply to 169 PWR and BWR plants on four continents, Framatome ANP has a very large database from operating experience feedback. The performance of Framatome PWR and BWR fuel is discussed for the period 1992-2001 with special emphasis on fuel failures, countermeasures and their effectiveness. While PWR fuel performance in most reactors has been good, the performance in some years did suffer from special circumstances that caused grid-to-rod fretting failures in few PWRs. After solving this problem, fuel of all types showed high reliability again. Especially the current PWR fuel products AFA 3G, HTP, Mark B and Mark BW showed a very good operating performance. Fuel reliability of Framatome ANP BWR fuel has been excellent over the last decade with average annual fuel rod failure rates under 1x10-5 since 1991. More than 40% of all BWR fuel failures in the 1992-2001 decade were caused by debris fretting. The debris problem has been remedied with the FUELGUARDTM lower tie plate, and by reactor operators' efforts to control the sources of debris. PCI, the main failure mechanism in former periods, affected only 10 rods. All of these rods had non-liner cladding. (author)

  10. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  11. Make use of EDF orientations in PWR fuel management

    International Nuclear Information System (INIS)

    The EDF experience acquired permits to allow the PWR fuel performances and to make use of better management. In this domain low progress can be given considerable financial profits. The industrial and commercial structures, the time constant of the fuel cycle, has for consequence that the electric utilities can take advantage only progressively of the expected profits

  12. 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.

  13. Long term Integrity of PWR Spent Fuel in Dry Storage

    International Nuclear Information System (INIS)

    The newly established organization KRMC (Korea radioactive waste management corporation) which is responsible for all kinds of radioactive waste generated in the Republic of Korea launched the PWR spent fuel dry storage research project in June 2009. This project has objectives to develop a storage system and evaluate the integrity of PWR fuel in dry storage. The project consists of three steps. At first step, it would develop own degradation models by referring to pre-exist good models and develop the hot test scenarios. Second step, test facilities would be constructed and used for testing the degradation behaviour in each mechanisms and in total. As a final step, total evaluation code would be developed by integrating each degradation model produced in the first step and the test data produced in the second step. All the activities would be summarized into a report and applied to licensing work. The Republic of Korea PWR spent fuels have unique characteristics of various fuel types (array type, clad material) and high capacity factor (maximum usage of fuel which is bad for integrity). These facts could impact on the research ranges of experimental data needed for degradation evaluation. In this research, spent fuel performance data concerning long term dry storage will be analysed and the major degradation mechanisms like creep and hydride behaviour will be studied and proposed for Korean PWR spent fuels

  14. A comparative study of fuel management in PWR reactors

    International Nuclear Information System (INIS)

    A study about fuel management in PWR reactors, where not only the conventional uranium cycle is considered, but also the thorium cycle as an alternative is presented. The final results are presented in terms of U3O8 demand and SWU and the approximate costs of the principal stages of the fuel cycle, comparing with the stardand cycle without recycling. (E.G.)

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

    International Nuclear Information System (INIS)

    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. Nuclear Power Station Kalkar, 300 MWe Nuclear Prototype Power Station with Fast Sodium Cooled Reactor (SNR-300), Safety Report Reactor Core Mark-Ia

    International Nuclear Information System (INIS)

    The nuclear power station Kalkar is a prototype with a sodium cooled fast reactor (SNR-300) and a thermal power of 762 MW. The initial licensing procedure in 1972 was based on the so-called Mark-I core. During the following years, this core underwent some changes, for instance the thickness of the radial blanket was reduced to lower the electricity generation costs, the design of the absorber systems had been further optimized, and it became clear, that a full core with plutonium from MAGNOX-reactors could not be realized and that fuel from light-water reactors had also to be used. In this licensing document the modified reactor core Mark-Ia is described together with its assemblies and their loading procedure. The content of radioactive materials and the irradiation protection measures are discussed and those accidents are describe in an enveloping manner, from which an influence of the core modification cannot be excluded. Finally, both core versions (Mark-I and Mark-Ia) are compared with each other

  17. Commissioning of the THTR-300-MWe prototype power plant - A milestone for further application of this high-temperature reactor line

    International Nuclear Information System (INIS)

    With the completion of the THTR 300 and the development of the follow-on plant HTR 500, the BBC/HRB company group has taken the pebble bed high-temperature reactor to the threshold of the commercial stage. The HTR is an important innovation in the field of reactor technology which can play an important role in the intermediate and long-term supply of safe, environmental friendly and economic energy. The power level of 550 MW meets the requirements of the present energy market which shows a trend towards smaller power units as a result of grid size, investment effort, and the slower increase in electricity demand in industrial nations. The advantages of the high-temperature reactor, such as high thermal efficiency, low waste heat, low radiation exposure of operating and maintenance personnel, high inherent safety, simple mode of operation, flexible fuel cycle with the potential to extend fuel resources, high availability, are currently uncontested and will represent the future standards for the peaceful uses of nuclear energy. For special applications in industry (steam and electric power as a cogeneration product) and in case of special siting conditions (near industrial centers), BBC/HRB developed a small 100 MW HTR, which can also be constructed as a 200 MW twin plant at favorable cost conditions. For an economic use of domestic coal in a processed form, the HTR represents the optimum solution as to economic and environmental aspects as well as extension of resources, especially if combined with conventional gasification procedures and in direct application of nuclear process heat at high gas temperatures of about 950 deg. C. In this field the development of the heat-exchanging components remains to be completed, before commercial application will be possible. The HTR is particularly well suited for erection in developing countries and industrial threshold countries which turn to nuclear energy for the first time. On an international level the interest in the pebble bed high-temperature reactor has also increased recently. Thus the HTR is of great importance to electric power industry and industrial development

  18. Nuclear Power Station Kalkar, 300 MWe Nuclear Prototype Power Station with Fast Sodium Cooled Reactor (SNR-300), Short Description of the Reactor Core Mark-Ia

    International Nuclear Information System (INIS)

    The nuclear power station Kalkar is a prototype with a sodium cooled fast reactor (SNR-300) and a thermal power of 762 MW. The initial licensing procedure in 1972 was based on the so-called Mark-I core. During the following years, this core underwent some changes, for instance the thickness of the radial blanket was reduced to lower the electricity generation costs, the design of the absorber systems had been further optimized, and it became clear, that a full core with plutonium from MAGNOX-reactors could not be realized and that fuel from light-water reactors had also to be used. In this licensing document the modified reactor core Mark-Ia is described, and the radiological consequences of the core modification are quantified to be tolerable

  19. Report on the PWR-radiation protection/ALARA Committee

    International Nuclear Information System (INIS)

    In 1992, representatives from several utilities with operational Pressurized Water Reactors (PWR) formed the PWR-Radiation Protection/ALARA Committee. The mission of the Committee is to facilitate open communications between member utilities relative to radiation protection and ALARA issues such that cost effective dose reduction and radiation protection measures may be instituted. While industry deregulation appears inevitable and inter-utility competition is on the rise, Committee members are fully committed to sharing both positive and negative experiences for the benefit of the health and safety of the radiation worker. Committee meetings provide current operational experiences through members providing Plant status reports, and information relative to programmatic improvements through member presentations and topic specific workshops. The most recent Committee workshop was facilitated to provide members with defined experiences that provide cost effective ALARA performance

  20. Industrial assessment of nonbackfittable PWR design modifications. Final report

    International Nuclear Information System (INIS)

    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

  1. Investigation, experiment and analysis on PWR sump screen clogging issue

    International Nuclear Information System (INIS)

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Flow clogging characteristics were investigated based on data for the relation of pressure loss and flow velocity during flow clogging due to debris accumulation. Deposition of chemical precipitates on the fuel cladding using an electrically heated rod was investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis with a thermal-hydraulic code on the downstream effect has shown that the core could be cooled because the core inlet flow compensates a evaporation of coolant due to the decay-heat even if core inlet was 99% clogged just after the ECCS recirculation operation started during the cold-leg break LOCA in PWR plants. (author)

  2. Investigation, experiment and analysis on PWR sump screen clogging issue

    International Nuclear Information System (INIS)

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the chemical effect and the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Chemical effect tests show that corrosion of carbon steel and galvanized steal may come to be important in domestic plants, in addition to corrosion of aluminum and insulator which has been considered dominant in the chemical effect. With respect to the downstream effect, deposition of chemical precipitates on the fuel cladding using an electrically heated rod is investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis on the downstream effect has shown that even if core inlet was completely clogged just after the recirculation operation started during LOCA in PWR plants, although upper part of core may be uncovered temporary and cladding temperature increased, core could be cooled by coolant injection through the hot-leg. (author)

  3. A concept of PWR using plate and shell heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de, E-mail: luciano.ondir@gmail.com, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  4. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  5. A concept of PWR using plate and shell heat exchangers

    International Nuclear Information System (INIS)

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  6. On catholyte application for hydrogen water chemistry in PWR

    International Nuclear Information System (INIS)

    Considering liquid water as a chemical compound with a wide band gap shows that its Redox potential as Fermi level in the band gap is the measurable characteristic of a non-stoichiometric aqueous coolant in recirculation system of PWR. The hypo-stoichiometric state with the negative Redox potential is realized when Fermi level is shifted to the bottom of conduction band. This state can be fixed by the electro-reduced water (catholyte) of the alkaline solution. Then, the hydride anions (H3O-) as proton acceptors and the hydrox-onium radicals (H3O) as electron donors are emerged in the alkaline catholyte and form hydrated clusters (AH)n(H2O)m of alkaline hydride. These particles as very strong reducers have a molar portion more than the gaseous hydrogen in the aqueous coolant and are the effective remedy for holding the negative Redox potential as an effect of hydrogen water chemistry in PWR. (authors)

  7. A burnup credit calculation methodology for PWR spent fuel transportation

    International Nuclear Information System (INIS)

    A burnup credit calculation methodology for PWR spent fuel transportation has been developed and validated in CEA/Saclay. To perform the calculation, the spent fuel composition are first determined by the PEPIN-2 depletion analysis. Secondly the most important actinides and fission product poisons are automatically selected in PEPIN-2 according to the reactivity worth and the burnup for critically consideration. Then the 3D Monte Carlo critically code TRIMARAN-2 is used to examine the subcriticality. All the resonance self-shielded cross sections used in this calculation system are prepared with the APOLLO-2 lattice cell code. The burnup credit calculation methodology and related PWR spent fuel transportation benchmark results are reported and discussed. (authors)

  8. Three basic options for the management of PWR waste

    International Nuclear Information System (INIS)

    Relying on the national practices of France, Germany and Belgium, three reference management routes for PWR wastes were drawn up and subsequently evaluated in terms of costs and radiological impact. It was thus demonstrated that safety regulations and technical redundancies, especially for off-gas treatment, liquid waste processing and dry solid waste treatment, play an important part in the cost associated with each route. The analysis of the different treatment options for mixed solid low level waste highlighted the low cost effectiveness of incineration as compared to compaction. Whatever the scenario investigated, the disposal costs of PWR wastes proved to be quite marginal in the overall cost. The radiological impact associated with each route was assessed through individual doses resulting from liquid and gaseous effluents. This theoretical exercise included some sensitivity studies performed on a selection of important parameters

  9. PWR plant transient analyses using TRAC-PF1

    International Nuclear Information System (INIS)

    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

  10. Study on thermal-hydraulics during a PWR reflood phase

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  11. ORNL-PWR BDHT analysis procedure: an overview

    International Nuclear Information System (INIS)

    The key computer programs currently used by the analysis procedure of the ORNL-PWR Blowdown Heat Transfer Separate Effects Program are overviewed with particular emphasis placed on their interrelationships. The major modeling and calculational programs, COBRA, ORINC, ORTCAL, PINSIM, and various versions of RELAP4, are summarized and placed into the perspective of the procedure. The supportive programs, REDPLT, ORCPLT, BDHTPLOT, OXREPT, and OTOCI, and their uses are described

  12. Improvement on main control room for Japanese PWR plants

    International Nuclear Information System (INIS)

    The main control room which is the information center of nuclear power plant has been continuously improved utilizing the state of the art ergonomics, a high performance computer, computer graphic technologies, etc. For the latest Japanese Pressurized Water Reactor (PWR) plant, the CRT monitoring system is applied as the major information source for facilitating operators' plant monitoring tasks. For an operating plant, enhancement of monitoring and logging functions has been made adopting a high performance computer

  13. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

    The behavior of a PWR fuel rod when operating in normal power cycles, excluding in case of accidents, is analysed. A computer code, that makes the mechanical analysis of the cladding using the finite element method was developed. The ramps and power cycles were simulated suposing the existence of cracks in pellets when the cladding-pellet interaction are done. As a result, an operation procedure of the fuel rod in power cycle is recommended. (E.G.)

  14. Propagation of nuclear data Uncertainties for PWR core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cabellos, O.; Castro, E.; Ahnert, C.; Holgado, C. [Dept. of Nuclear Engineering, Universidad Politecnica de Madrid, Madrid (Spain)

    2014-06-15

    An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for {sup 235,238}U, {sup 239}Pu, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.

  15. In-pile test of Qinshan PWR fuel bundle

    International Nuclear Information System (INIS)

    In-pile test of Qinshan Nuclear Power Plant PWR fuel bundle has been conducted in HWRR HTHP Test loop at CIAE. The test fuel bundle was irradiated to an average burnup of 25000 Mwd/tU. The authors describe the structure of (3 x 3-2) test fuel bundle, structure of irradiation rig, fuel fabrication, irradiation conditions, power and fuel burnup. Some comments on the in-pile performance for fuel bundle, fuel rod and irradiation rig were made

  16. Safety Analysis Report for the PWR Spent Fuel Canister

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Choi, Jong Won; Cho, Dong Keun; Chun, Kwan Sik; Lee, Jong Youl; Kim, Seong Ki; Kim, Seong Soo; Lee, Yang

    2005-11-15

    This report outlined the results of the safety assessment of the canisters for the PWR spent fuels which will be used in the KRS. All safety analyses including criticality and radiation shielding analyses, mechanical analyses, thermal analyses, and containment analyses were performed. The reference PWR spent fuels were in the 17x17 and determined to have 45,000 MWD/MTU burnup. The canister consists of copper outer shell and nodular cast iron inner structure with diameter of 102 cm and height of 483 cm. Criticality safety was checked for normal and abnormal conditions. It was assumed that the integrity of engineered barriers is preserved and saturated with water of 1.0g/cc for normal condition. For the abnormal condition container and bentonite was assumed to disappear, which allows the spent fuel to be surrounded by water with the most reactive condition. In radiation shielding analysis it was investigated that the absorbed dose at the surface of the canister met the safety limit. The structural analysis was conducted considering three load conditions, normal, extreme, and rock movement condition. Thermal analysis was carried out for the case that the canister with four PWR assemblies was deposited in the repository 500 meter below the surface with 40 m tunnel spacing and 6 m deposition hole spacing. The results of the safety assessment showed that the proposed KDC-1 canister met all the safety limits.

  17. Validation of gadolinium burnout using PWR benchmark specification

    Energy Technology Data Exchange (ETDEWEB)

    Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2014-07-01

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  18. Validation of gadolinium burnout using PWR benchmark specification

    International Nuclear Information System (INIS)

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor Kinf, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157

  19. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  20. Actinides transmutation - a comparison of results for PWR benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Claro, Luiz H. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)], e-mail: luizhenu@ieav.cta.br

    2009-07-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO{sub 2} used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k{infinity} and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  1. QFLOOD-GT: a program for predicting PWR reflood

    International Nuclear Information System (INIS)

    A description is given of the present version of the QFLOOD-GT program for predicting the reflood stage of a large-break PWR loss-of-coolant accident. QFLOOD-GT has been developed from an earlier forced-reflood program which, using a conduction-controlled model for rewetting speed, gave good agreement with the FLECHT SEASET experiments. This earlier program has been incorporated into QFLOOD-GT as a subroutine called QFLOOD; in addition a downcomer model has been included in order to allow calculation of gravity reflood, and a computational scheme has been devised to simulate the chimney effect (the unequal distribution of inlet flow between hot and cool regions of the core). No quantitative comparisons between QFLOOD-GT predictions and integral-test data have yet been carried out, so the modelling decisions implemented in the program are at this stage unvalidated. Preliminary testing of the program has produced results which are for the most part qualitatively satisfactory. Calculations for indicative PWR conditions suggest that the chimney effect has a significant beneficial effect during PWR reflood, a conclusion in accordance with the findings of the Japanese 2D/3D experiments. (author)

  2. PWR experimental benchmark analysis using WIMSD and PRIDE codes

    International Nuclear Information System (INIS)

    Highlights: • PWR experimental benchmark calculations were performed using WIMSD and PRIDE codes. • Various models for lattice cell homogenization were used. • Multiplication factors, power distribution and reaction rates were studied. • The effect of cross section libraries on these parameters was analyzed. • The results were compared with experimental and reported results. - Abstract: The PWR experimental benchmark problem defined by ANS was analyzed using WIMSD and PRIDE codes. Different modeling methodologies were used to calculate the infinite and effective multiplication factors. Relative pin power distributions were calculated for infinite lattice and critical core configurations, while reaction ratios were calculated for infinite lattice only. The discrete ordinate method (DSN) and collision probability method (PERSEUS) were used in each calculation. Different WIMSD cross-section libraries based on ENDF/B-VI.8, ENDF/B-VII.0, IAEA, JEF-2.2, JEFF-3.1 and JENDL-3.2 nuclear data files were also employed in the analyses. Comparison was made with experimental data and other reported results in order to find a suitable strategy for PWR analysis

  3. PWR primary coolant sample lines - problems with measurement of corrosion products and experimental proposals for Ringhals PWR

    International Nuclear Information System (INIS)

    Coolant samples are drawn from PWR primary circuits through long narrow tubes. Concern that interaction with the sample line walls (by deposition and release) can result in inaccurate measurement of corrosion product concentrations has recently intensified after several observations of a dependence on sample line flow rate. Particularly significant instances of this have been observed at Ringhals PWR. A further problem is that measured concentrations show spurious transient increases after valving in the sample line. Sampling behaviour is complex since it involves particulate as well as soluble material, and deposition and release as well as localised phenomena associated with crud traps within the sample line. The present report has threefold function, firstly to review instances of anomalous sample line behaviour and secondly to present a basic theoretical background to aid interpretation of such behaviour. The third and most important function is to suggest plant measurements which might be made at Ringhals PWR to understand better the response of the sampling system by quantifying the effects due to corrosion product deposition on, and release from, sample line walls. (author)

  4. Research program of natural circulation steam generator design of national 1000 MWe PWR

    International Nuclear Information System (INIS)

    A concept design of natural circulation steam generator for the national 1000MWe PWR of the Chinese National Nuclear Program has been proposed and a relevant research program to validate the efficiency and/or effectiveness of some new design assemblies and/or components has been completed. There are three salient features in the steam generator design. Firstly, steam separation equipment was improved and carryover moisture was further reduced to below 0.1%. Secondly, the water level at the secondary side being elevated, secondary side water volume was expanded to satisfy the EPRI-URD requirements of LFB20. Finally, an inactive device, sludge collector was incorporated to enhance the secondary sludge control. The validation research program consists of two parts; cold state screening test, hot state validation test and corresponding computational analysis, and cold state test and corresponding computational analysis of sludge collector design. The validation tests were completed in 2001. Three sets of steam separator were selected from cold screening tests for hot validation. The hot test showed the most important parameters, outlet carryover, of all three sets were under 0.1%. The best case was only 0.0018%. The sludge collector showed a collection efficiency of over 50%

  5. EPRI PWR Safety and Relief Valve Test Program: test condition justification report

    Energy Technology Data Exchange (ETDEWEB)

    Hosler, J.

    1982-12-01

    In response to NUREG 0737, Item II.D.1.A requirements, several safety and relief valve designs were tested by EPRI under PWR utility sponsorship. Justification that the inlet fluid conditions under which these valve designs were tested are representative of those expected in participating domestic PWR units during FSAR, Extended High Pressure Injection, and Cold Overpressurization events is presented.

  6. PWR safety and relief valve test program. Valve selection/juftification report. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    NUREG 0578 required that full-scale testing be performed on pressurizer safety valves and relief valves representative of those in use or planned for use in PWR plants. To obtain valve performance data for the entire population of PWR plant valves, nine safety valves and ten relief valves were selected as a fully representative set of test valves. Justification that the selected valves represent all PWR plant valves was provided by each safety and relief valve manufacturer. Both the valve selection and justification work was performed as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of the PWR utilities in response to the recommendations of NUREG 0578 and the requirements of the NRC. Results of the Safety and Relief Valve Selection and Justification effort is documented in this report.

  7. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.

    2015-07-01

    During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

  8. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  9. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

  10. Development of laser weld monitoring system for PWR space grid

    International Nuclear Information System (INIS)

    The laser welding monitoring system was developed to inspect PWR space grid welding for KNFC. The demands for this optical monitoring system were applied to Q.C. and process control in space grid welding. The thermal radiation signal from weld pool can be get the variation of weld pool size. The weld pool size and depth are verified by analyzed wavelength signals from weld pool. Applied this monitoring system in space grid weld, improved the weld productivity. (author). 4 refs., 5 tabs., 31 figs

  11. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author)

  12. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)

    1997-02-01

    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.

  13. Sizewell B - analysis of British application of US PWR technology

    International Nuclear Information System (INIS)

    This report provides information on the staff's evaluation of major design differences and issues developed by the British in their application (Sizewell B) of US PWR technology. One design change, the addition of steam-driven charging pumps, was assessed to have a relatively high value compared to the other changes. However, the assessment is based on a number of assumptions for which inadequate data exist to make an unqualified judgment. Other changes to the US design (as typified by the SNUPPS design) were found to have relatively low or moderate safety benefits for US application

  14. Conversion ratio in epithermal PWR, in thorium and uranium cycle

    International Nuclear Information System (INIS)

    Results obtained for the conversion ratio in PWR reactors with close lattices, operating in thorium and uranium cycles, are presented. The study of those reactors is done in an unitary fuel cell of the lattices with several ratios V sub(M)/V sub(F), considering only the equilibrium cycles and adopting a non-spatial depletion calculation model, aiming to simulate mass flux of reactor heavy elements in the reactor. The neutronic analysis and the cross sections generation are done with Hammer computer code, with one critical apreciation about the application of this code in epithermal systems and with modifications introduced in the library of basic data. (E.G.)

  15. Axial simulation of PWR core and study of actuators

    International Nuclear Information System (INIS)

    Development of an operation code allowing to simulate the behaviour of a PWR type reactor core. Load following is controled by bore and control rods, taking into account the temperature counter-reactions. The fine behaviour of the fuel element during transients is not simulated, on the other hand the central part of the reactor is completely simulated. The regulation equation are easily modifiable and thus it is possible to test in open loop any modification brought about to this regulation. Description of simulation tests on CAS-2B reactor: core control, static tests, dynamic tests

  16. Modelling of pellet-cladding interaction in PWR's

    International Nuclear Information System (INIS)

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

  17. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  18. Life management plants at nuclear power plants PWR; Planes de gestion de vida en centrales nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esteban, G.

    2014-10-01

    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  19. SCOR 1000: an economic and innovative conceptual design PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M.; Chenaud, M.S. [CEA Cadarache (DEN/DER/SESI), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Reacteurs; Tourniaire, B. [CEA Grenoble (DEN/DTN/SE2T/LPTM), 38 (France)

    2007-07-01

    Within the framework of innovative reactors studies, the Cea proposes the SCOR design (Simple COmpact Reactor) based on most of the advantages of innovative reactors. All main components are integrated in the vessel: the pressurizer, the canned pumps, the control rod mechanics of the driving system (CMD), and the dedicated heat exchangers of the passive heat removal system. The only steam generator is located above the vessel instead of the upper head. This design is featured by its compactness and by a large suppression or simplification of auxiliary systems. The first design with a 600 MWe shows its competitiveness with regard to the large loop-type PWR. To reduce the cost investment by the law sized effect, we examine the possibility of increasing the power of the reactor, while keeping the safety advantages of the medium sized SCOR. The electrical power of the new design is 1000 MWe. SCOR-1000 operates at much lower primary circuit pressure than standard PWRs (93 bars instead of the usual 155 bars), and the power density is lower (80 MW/m3 instead of 100 for the present PWRs). The reactivity is controlled by the CMD and by the burnable poison, without soluble boron. With the same safety advantages of the medium-sized SCOR, the cost reduction of the investment and of cost production could reach 18% with regard to the loop-type PWR. (authors)

  20. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  1. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces obtained by Navier-Stokes momentum equation about a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break. (5) The blowdown thrust force in the analysis greatly depends on the selection of the exit pressure. (author)

  2. Degradation of fastener in reactor internal of PWR

    International Nuclear Information System (INIS)

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  3. VERA Core Simulator Methodology for PWR Cycle Depletion

    Energy Technology Data Exchange (ETDEWEB)

    Kochunas, Brendan [University of Michigan; Collins, Benjamin S [ORNL; Jabaay, Daniel [University of Michigan; Kim, Kang Seog [ORNL; Graham, Aaron [University of Michigan; Stimpson, Shane [University of Michigan; Wieselquist, William A [ORNL; Clarno, Kevin T [ORNL; Palmtag, Scott [Core Physics, Inc.; Downar, Thomas [University of Michigan; Gehin, Jess C [ORNL

    2015-01-01

    This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclear reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.

  4. PWR reactor vessel in-service-inspection according to RSEM

    Energy Technology Data Exchange (ETDEWEB)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul [Intercontrole, 13, rue du Capricorne - SILIC 433, 94583 Rungis - Cedex (France)

    2006-07-01

    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

  5. Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

    International Nuclear Information System (INIS)

    This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

  6. Benchmark exercise on SBLOCA experiment of PWR PACTEL facility

    International Nuclear Information System (INIS)

    Highlights: • PWR PACTEL, the facility with EPR type steam generators, is introduced. • The focus of the benchmark was on the analyses of the SBLOCA test with PWR PACTEL. • System codes with several modeling approaches were utilized to analyze the test. • Proper consideration of heat and pressure losses improves simulation remarkably. - Abstract: The PWR PACTEL benchmark exercise was organized in Lappeenranta, Finland by Lappeenranta University of Technology. The benchmark consisted of two phases, i.e. a blind and an open calculation task. Seven organizations from the Czech Republic, Germany, Italy, Sweden and Finland participated in the benchmark exercise, and four system codes were utilized in the benchmark simulation tasks. Two workshops were organized for launching and concluding the benchmark, the latter of which involved presentations of the calculation results as well as discussions on the related modeling issues. The chosen experiment for the benchmark was a small break loss of coolant accident experiment which was performed to study the natural circulation behavior over a continuous range of primary side coolant inventories. For the blind calculation task, the detailed facility descriptions, the measured pressure and heat losses as well as the results of a short characterizing transient were provided. For the open calculation task part, the experiment results were released. According to the simulation results, the benchmark experiment was quite challenging to model. Several improvements were found and utilized especially for the open calculation case. The issues concerned model construction, heat and pressure losses impact, interpreting measured and calculated data, non-condensable gas effect, testing several condensation and CCFL correlations, sensitivity studies, as well as break modeling. There is a clear need for user guidelines or for a collection of best practices in modeling for every code. The benchmark offered a unique opportunity to test

  7. Frictional Behavior of Fe-based Cladding Candidates for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Hyung-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Byun, Thak Sang [Oak Ridge National Lab., Oak Ridge (United States)

    2014-10-15

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

  8. 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.

  9. Knowledge-based diagnosis of PWR secondary water chemistry

    International Nuclear Information System (INIS)

    A prototype knowledge-based diagnostic system has been developed for more effective processing of the in-line chemistry sensor data from the PWR secondary water-steam circuit with the SUN 3/80 workstation and the Nexpert Object shell program. The system consists of the data interface, the data interpreter, the CHEMISTRY-expert, the ACTION-expert, and the user interface. The knowledge base defines physical and conceptual models of the target domain in a class/object hierarchy, giving rise to a reduced number of rules with pattern matching. The rule base is broken down into separate rule groups for task control, classification, prioritization, and diagnosis to minimize the inference time. The system is scheduled for the Verification and Validation test to collect operational information feedback in one of the Korea nuclear power plants in the near future. (author)

  10. Modeling local chemistry in PWR steam generator crevices

    Energy Technology Data Exchange (ETDEWEB)

    Millett, P.J. [EPRI, Palo Alto, CA (United States)

    1997-02-01

    Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledge of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.

  11. PWR steam generator chemical cleaning, Phase I. Final report

    International Nuclear Information System (INIS)

    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI

  12. Failure probability of PWR reactor coolant loop piping

    International Nuclear Information System (INIS)

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria

  13. Design of an FPGA-based PWR ATWS mitigation system

    International Nuclear Information System (INIS)

    The present research is to explore the feasibility and conceptual design by using triple-redundant FPGA-based system for Anticipated-Transient-Without-Scram (ATWS) Mitigation System and Actuation Circuit (AMSAC) of a pressurized water reactor (PWR) type nuclear power plant (NPP). The Taipower's (Taiwan Power Company) Maanshan NPP was chosen for demonstration. An engineering simulated interface between AMSAC system and reactor/plant systems of Maanshan NPP was developed to provide an environment to validate the triple-redundant FPGA-based system. The software-free FPGA-based nuclear instrumentation and control (I and C) systems can easily be used for the modernization of the Taipower's nuclear power plant analog systems, thus may reduce the safety risk of undetectable software faults and common cause failures, and also minimize the regulatory licensing efforts and cost. (author)

  14. Development of high temperature adsorbent in PWR primary system

    International Nuclear Information System (INIS)

    Radiation exposure reduction in PWR is one of the most important problems to be solved. We have developed a high temperature Co adsorbent (HTA), which could be directly applied under primary reactor coolant conditions. This adsorbent was Fe-Ti-O system ceramics, and was fabricated to a suitable form for using in a packed column. Through those experiments of adsorption tests, compatibility tests, leaching tests and hot loop tests, it was found that HTA had superior adsorption capability to not only Co and Ni-ion but also many other transition metal ions. And it was also found that HTA was compatible with high temperature water, as well as advantageous for its waste solidification. Based on the experimental results, dose reduction effect was evaluated by a computer code. From this evaluation, it was found that more than 50 % dose reduction could be expected, when an advanced reactor coolant clean-up (RCC) system with HTA would be realized. (author)

  15. Sizewell 'B' PWR pre-construction safety report

    International Nuclear Information System (INIS)

    The Pre-Construction Safety Report (PCSR) for a PWR power station to be constructed as Sizewell 'B' is presented in 13 volumes containing 16 chapters. The PCSR has been submitted to the Nuclear Installations Inspectorate in support of the Central Electricity Generating Board's application for consent to the extension at Sizewell. It describes the design and provides the safety case for the proposed station, which comprises a 4-loop pressurized water reactor with associated generating plant and supporting auxiliary equipment. A general description of the station and its site is given. The strategy for ensuring nuclear safety is set out and the general design aspects of systems and plant outlined. The plant and systems, including their safety design bases and the fault analyses carried out for the design are described. Finally the way in which the plant will be decommissioned at the end of its useful life is outlined. (U.K.)

  16. Specification of water quality for the FRAMATOME PWR secondary circuit

    International Nuclear Information System (INIS)

    This paper describes the purpose, theory and scope of secondary system chemical specifications for FRAMATOME PWR nuclear power plants. All volatile treatment was chosen: controlling the feedwater pH by means of a volatile amine (ammonia, morpholine), and excluding oxygen by the addition of hydrazine. The pollutants are monitored at the steam generator drains by completely automatic measurements using simple and reliable techniques: pH measurement and a diagram of the cation conductivity versus sodium. An explanation is given of the monitoring techniques and to the effect of the various kinds of possible pollutant. A new concept is described, the annual quota expressed in day.microsiements.cm-1 which enables the amount of absorbed pollutants in the steam generator to be evaluated. The methods used for maintaining the desired chemical quality are dealt with

  17. Non linear identification applied to PWR steam generators

    International Nuclear Information System (INIS)

    For the precise industrial purpose of PWR nuclear power plant steam generator water level control, a natural method is developed where classical techniques seem not to be efficient enough. From this essentially non-linear practical problem, an input-output identification of dynamic systems is proposed. Through Homodynamic Systems, characterized by a regularity property which can be found in most industrial processes with balance set, state form realizations are built, which resolve the exact joining of local dynamic behaviors, in both discrete and continuous time cases, avoiding any load parameter. Specifically non-linear modelling analytical means, which have no influence on local joined behaviors, are also pointed out. Non-linear autoregressive realizations allow us to perform indirect adaptive control under constraint of an admissible given dynamic family

  18. PWR steam generator chemical cleaning, Phase I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Rothstein, S.

    1978-07-01

    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.

  19. Design of large steam turbines for PWR power stations

    International Nuclear Information System (INIS)

    The authors review the thermodynamic cycle requirements for use with pressurized-water reactors, outline the way thermal efficiency is maximized, and discuss the special nature of the wet-steam cycle associated with turbines for this type of reactor. Machine and cycle parameters are optimized to achieve high thermal efficiency, particular attention being given to arrangements for water separation and steam reheating and to provisions for feedwater heating. Principles and details of mechanical design are considered for a range both of full-speed turbines running at 3000 rev/min on 50 Hz systems and of half-speed turbines running at 1800 rev/min on 60 Hz systems. The importance of service experience with nuclear wet-stream turbines, and its relevance to the design of modern turbines for PWR applications, is discussed. (author)

  20. PWR pressurizer discharge piping system on-site testing

    International Nuclear Information System (INIS)

    Framatome PWR systems includes the installation of safety valves and relief valves wich permit the discharge of steam from the pressurizer to the pressurizer relief tank through discharge piping system. Water seal expulsion pluration then depends on valve stem lift dynamics which can vary according to water-stem interaction. In order to approaches the different phenomenons, it was decided to perform a test on a 900 MWe French plant, test wich objectives are: characterize the mechanical response of the discharge piping to validate a mechanical model; open one, two or several valves among the following: one safety valve and three pilot operated relief valves, at a time or sequentially and measure the discharge piping transient response, the support loads, the

  1. PWR-blowdown heat transfer separate effects program

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. CFD application to PWR subchannel void distribution benchmark

    International Nuclear Information System (INIS)

    A CFD study is performed to simulate the steady-state void distribution benchmark based on the NUPEC PWR Subchannel and Bundle Tests (PSBT). The CFD calculation predicted the void distributions in central typical and thimble subchannels, side subchannel and corner subchannel. The CFD prediction shows a higher void fraction near the heated wall and a migration of void in the subchannel gap region. A measured image of void distribution indicated a locally higher void fraction near the heated wall. The CFD predictions of void fraction and fluid density agree well with the measured ones for the low void test condition. However, the CFD calculations tend to underpredict the void fraction and overpredict the fluid density as the void fraction increases. (author)

  3. A system for trip analysis of PWR reactors using neural networks

    International Nuclear Information System (INIS)

    This work presents the basic concepts and the general description of a computational system developed for trip analysis in PWR nuclear power plants which is based on neural networks and artificial intelligence concepts. (author)

  4. N4 PWR makes full use of distributed processing and local networks

    Energy Technology Data Exchange (ETDEWEB)

    Aschenbrenner, J.F.; Tetreau, F.; Colling, J.M.

    1988-01-01

    The new instrumentation and control systems for the French N4 PWR power plant make extensive use of programmable controllers based on advanced microprocessor technology and distributed processing. Local networking techniques are widely used which simplify architecture and equipment design.

  5. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    International Nuclear Information System (INIS)

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  6. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Jang, Jung Nam; Hwang, Yong Hwa; Kwon, In Chan; Min, Duck Kee; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  7. Maintenance of Ni-based alloy at PWR plant

    International Nuclear Information System (INIS)

    Kansai Electric owns 11 PWR plants. At our PWR plants, we are taking various preventive maintenance measures on Ni-based alloy according to the prediction of possible trouble while past trouble occurred at overseas plants due to Primary Water Stress Corrosion Cracking (PWSCC) being considered. In addition, we are making an effort to put new maintenance techniques into practical use by conducting demonstration tests to confirm their applicability to actual plants. We have replaced reactor vessel heads at 7 plants with new ones. At the other 4 plants, we took, measures to reduce the temperature of reactor vessel head top to delay the timing of PWSCC occurrence. We are carrying out the constant load tests to predict the timing of PWSCC occurrence at these 4 plants. It is planned to conduct non-destructive inspections at an appropriate timing based on the result of the prediction. Based on the prediction of the timing of PWSCC occurrence at bottom-mounted instrumentation (BMI), we have developed water jet peening (WJP) technique to reduce residual stress and applied the technique to our plants successively. Meanwhile, a technique to cut and eliminate cracking has been developed. In addition, capping technique, which covers overall the concerned nozzle on the outer surface of the reactor vessel, has been also established. For alloy 132/82 weld metal for the connection, we are conducting ultrasonic inspection at our plants successively. In order to prepare against PWSCC occurrence, we have also established a technique to replace the entire section of concerned short piping with new one. (author)

  8. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    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 137Cs, 154Eu, and 134Cs 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.

  9. Westinghouse Passive Plants - AP600 and S PWR

    International Nuclear Information System (INIS)

    The original thought behind the AP600 passive design was that if the U. S. nuclear industry was to be revitalized, it would require a new, advanced technology with clearly proven benefits in safety. Response from the international arena indicates that, regardless of local domestic consideration, a revitalization of the U. S. industry is seen as very important, even essential, worldwide. And the potential for scale up of these passive safety features has been clearly established, allowing the benefits of the passive technology to be realized in countries that, for whatever reason, are interested in larger plant sizes only. Government projections indicate that U. S. energy demands in the 1990s will grow steadily, creating the need for approximately 117,000 to 322,000 MW of new generating capacity by the year 2010. Although this growth in electricity demand continues to be strong, orders for new nuclear power plants have not kept pace, in part due to licensing delays, prohibitive construction costs, and public uncertainty about safety. However, with the increased concerns about the environmental and economic security risks involved with an excessive dependence on fossil fuels, there is a growing realization that nuclear power must play a major role in our energy future. Looking to the future, Westinghouse is developing the AP600, a simplified two-loop PWR featuring passive safety systems. Drawing on the results of the AP600 development and testing programs, Westinghouse is also developing the larger S PWR, a passive, three-loop power plant with an output in the 900 to 1000 MW range

  10. First application of hollow fiber filter for PWR condensate polishing

    Energy Technology Data Exchange (ETDEWEB)

    Tsuda, S. [ORGANO Corp., Tokyo (Japan); Otoha, K.; Takiguchi, H. [Japan Atomic Power Co., Tokyo (Japan)

    2002-07-01

    In Tsuruga Unit-2 (PWR 1160 MWe commenced commercial operation in 1987), current procedure for secondary system clean-up before start-up had prolonged outage time and had consumed a huge amount of de-ionized (DI) water. In addition, iron oxide in condensate had accelerated the degradation of condensate demineralizer (CD) resin. The corrosion product of iron could also influence the secondary side corrosion of steam generator (SG) tubing if it intruded into SG through CD. To solve these problems, Japan Atomic Power Company (JAPC) decided to introduce hollow fiber filter (HFF) type condensate filter into Tsuruga-2, as the first application to PWR in the world. Because of retro-fitted HFF in Tsuruga Unit-2, limitations for installation space and flow resistance in condensate system and cost reduction required new design for compact and low differential pressure system and for long life filter module. JAPC and ORGANO assessed methodologies to achieve these goals. An advanced HFF system, including a newly developed compact HFF module design, was installed at Tsuruga Unit-2 in 1997 based on the assessment. During the 5 years since the installation, the HFF system has provided excellent crud removal that enables to shorten the outage period and to reduce DI water consumption drastically. Stable differential pressure (dP) trend of the HFF system indicates an expected module life of more than 7 years, with backwash cleaning required only 2 or 3 times per year. In addition to providing the expected operating cost reduction and improved SG tube integrity, numerous additional benefits have resulted from the retrofit. (authors)

  11. Identifying thermal cycling mechanisms in PWR branch line piping

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI, Charlotte, NC (United States); Keller, J.D.; Bilanin, A.J. [Continuum Dynamics, Inc., Ewing, NJ (United States)

    2002-07-01

    Predicting the onset and the characteristics of thermal cycling in pressurized water reactor (PWR) branch line piping systems is critical to formulation of thermal fatigue screening tools. The complex nature of the underlying thermal-hydraulic phenomena, however, significantly complicates prediction using analytical models or direct numerical simulations. Instead, it is necessary to perform scaled experiments to identify the physical mechanisms and to gather data for formulation of semi-empirical models for the thermal cycling phenomena. Through the EPRI Materials Reliability Program a test program is underway to identify and develop semi-empirical correlations for the physical thermalhydraulic mechanisms that cause thermal cycling in dead-ended PWR branch line piping systems. Three series of tests are being performed in this test program: configuration tests on a representative up-horizontal (UH) branch line piping geometry, configuration tests on a representative down-horizontal (DH) branch line piping geometry, and high Reynolds number tests to assess penetration of secondary flow structures into a dead-ended branch line. Results from UH and DH configuration tests indicate that random turbulence penetration is not sufficient for thermal cycling to occur. Rather a swirling flow structure, representative of a large, 'corkscrew' vortical structure, is required for thermal cycling. Scale tests on the UH configuration have simulated cycling phenomena observed in full-scale plant data and have been used to determine parametric sensitivities in formulating a predictive model for the thermal cycling. Data indicate that the mechanism for thermal cycling in UH configurations is stochastic but scales with the leak rate from the valve. The critical dependent variables are reduced to several non-dimensional scaling curves, resulting in a semiempirical predictive model. This paper discusses the test program and the results obtained to date. Application of these

  12. An Experience on RCS CRUD Sampling in European PWR Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Jae; Kim, Jong Bin [Sung Woo E and T Co., Seoul (Korea, Republic of); Kang, Duk Won [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    In most PWRs the normal method of corrosion product sampling is to collect a 'grab' sample from either the RCS hot or cold leg. This method is not ideal and the results are often dominated by soluble and particulate transients that can bias them high by factors of between ten and one hundred times. Nevertheless 'grab' samples can still give relatively satisfactory results from which qualitative trends of total soluble plus particulate corrosion product concentrations can be determined and, although 'grab' sampling may not be ideal, it is useful in detecting and following abnormal particulate releases from the core. It is possible to eliminate the worst of the transient effects by collecting a sample from continuously flowing RCS sample line, but the changes necessary to operate in this way are major, will be costly and may not be practicable for many existing plants. The evaluation of changes in corrosion product concentrations, particularly when the changes increase the particulate concentrations, can indicate that there is a risk that an Axial Offset Anomaly (AOA) may develop, or a risk of increased corrosion product releases when the plant shuts down for refueling. Recently, Diablo Canyon and Callaway in United States America, Ringhals in Sweden, Sizewell B in Great Britain, Vandellos in Spain and Doel in Belgium, these PWR plants have applied capillary sampling method to CRUD Analysis in parallel with grab sampling method under the recommendation of EPRI. In this thesis, it will show the practice based on actually tested method in European PWR plants.

  13. Estimating PWR fuel rod failures throughout a cycle

    International Nuclear Information System (INIS)

    A fuel performance engineer requires good prediction models for fuel conditions to help assure that any fuel repair operation he may recommend for the next refueling outage will have a minimal impact on nuclear plant operation. For nearly two decades, simple equilibrium equations have been used to provide estimates of the number of failed fuel rods in a pressurized water reactor (PWR) core. The unknown parameter is the isotopic escape rate (upsilon), which is often assumed to be --1 X 10/sup -8//s for the release of /sup 131/I from a 3- to 4-m-long PWR rod. The use of this escape rate value will generally produce end-of-cycle (EOC) predictions that are accurate within a factor of --3. When applied at the time when fuel rods initially fail, such as early in a reactor cycle, however, the prediction obtained may overestimate the number of failed rods present by a factor of 10 or more. While a goal of Combustion Engineering's (C-E's) efforts on failed fuel prediction (FFP) models over the past decade has been to increase the accuracy of the EOC estimate, recent efforts have emphasized improving prediction capability for failed rods present early in a reactor cycle. The C-E approach to modeling iodine release from failed fuel rods is based on dynamic escape rate theory that is incorporated in the C-E IODYNE (for iodine dynamic evaluation) code. This theory has been empirically modified to account for specific observed time dependencies of the release rates for /sup 131/I and /sup 133/I from a failed rod. In a current version of IODYNE, four such factors have been included in the FFP model, as described in this paper

  14. Development of the Combination Method of PWR Spent Fuel for DUPIC Fuel Preparation

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Ju Ho; Kim, S. K.; Jung, T. C.; June, T. H.; Lee, J. M.; Kim, I. S.; Park, C. S.; Kim, M. J.; An, J. I.; Park, S. H. [Kyung Hee University, Seoul (Korea, Republic of)

    1997-07-15

    Optimum finding method of PWR spent fuel was developed in application of DUPIC fuel composition to nuclear fuel production. In order to make the database of the PWR spent fuel for the optimum composition, composition data of the PWR spent fuels from Youngkwang unit 1 and 2, Kori unit 3 and 4 and Uljin unit 1 and 2 were collected, analyzed and stored. Artificial intelligent access was attempted in optimizing the composition, and the combination algorithm for PWR spent fuel was developed. In this work database of the composition data of the PWR spent fuels from Youngkwang unit 1 and 2, Kori unit 3 and 4 and Uljin unit 1 and 2 as well as their combination algorithm for PWR spent fuel were developed. The combination algorithm is to find the combination of the spent fuel assembly which is quite close to the requirement per unit mass of DUPIC fuel. The required data are total weight of the fuel, tolerance of the errors, importance of the elements and the discharge data. This combination algorithm enables to find the optimum PWR spent fuel assembly for DUPIC fuel with the database of the spent fuels according to the DUPIC fuel standards. The combination algorithm developed in this work can afford the technical support to fuel supply in preparing the DUPIC fuel, and make contribution in DUPIC fuel cycle technology. It can be directly used in DUPIC fuel cycle technology, and can be also used in the management of the spent fuels with respect to their compositions and ingredients as well as the nuclear safeguards. Composition of the PWR spent fuel in each assembly depends on the initial concentration, degree of combustion, specific power, and its location in the reactor core. They may be affected by the kind of fuel rod and its axial length. Therefore, analysis procedures in these regards should be established for the effective application of the results of this work. 6 refs., 12 tabs., 46 figs. (author)

  15. Bias identification in PWR pressurizer instrumentation using the generalized liklihood-ratio technique

    International Nuclear Information System (INIS)

    A method for detecting and identifying biases in the pressure and level sensors of a pressurized water reactor (PWR) pressurizer is described. The generalized likelihood ratio (GLR) technique performs statistical tests on the innovations sequence of a Kalman filter state estimator and is capable of determining when a bias appears, in what sensor the bias exists, and estimating the bias magnitude. Simulation results using a second-order linear, discrete PWR pressurizer model demonstrate the capabilities of the GLR method

  16. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    Energy Technology Data Exchange (ETDEWEB)

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  17. Identification and evaluation of PWR in-vessel severe accident management strategies

    International Nuclear Information System (INIS)

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

  18. Things Chinese.

    Science.gov (United States)

    Law, Yip Wang

    Presented in this booklet are brief descriptions of items and activities that are symbolic of Chinese culture. Some of the items and activities described include a traditional Chinese child's outfit, dolls, sandalwood fans, writing and printing materials and techniques, toys and crafts, a Chinese abacus, and eating utensils. Several recipes for…

  19. From Daya Bay to Ling Ao. The benefits of a duplication policy

    International Nuclear Information System (INIS)

    Over the past 15 years, the People's Republic of China has experienced very rapid economic growth of annual average 8%, which must be supported by fast expanding energy production, notably of electricity. China has the considerable amount of coal resources, but most of these resources are located in the north of the country, and the vast hydroelectric potential in Southwestern China is difficult to develop. Therefore, in the coastal provinces of Southeast China, where economic expansion is greatest, nuclear power has been chosen to meet the need. The Qinshan No. 1 PWR with 300 MWe output is the first Chinese nuclear power facility, and started the operation in 1992. Two 985 MWe PWRs have been operated since 1994 at Daya Bay. The construction of Qinshan No. 2 and 3 PWRs of 600 MWe each are in progress, and are expected to start the operation in 2001. These plants were designed by China based on the Framatome technology. Two more 985 MWe plants will be constructed on Ling Ao site, and operated in 2002 and 2003. The main milestones of Framatome collaboration with Chinese partners are explained. The Daya Bay nuclear power station and the Ling Ao project are reported. The benefits of duplicating the Daya Bay nuclear power station at Ling Ao are summarized. The M310 PWR plants of 985 MWe are the modern, proven type backed by more than reactor-years of operating experience. (K.I.)

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

    International Nuclear Information System (INIS)

    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)

  1. The chemical decontamination of the Callisto PWR loop

    International Nuclear Information System (INIS)

    The CALLISTO (Capability for Light water Irradiation in Steady state and Transient Operation) is a PWR experimental facility for scientific in-pile studies installed into the BR2 Material Test Reactor. Three experimental rigs, called In-Pile Sections (IPS), are installed in three reactor channels. They are connected to a common pressurized loop, which operates with representative PWR water chemistry (typically 400 ppm boron, 3,5 ppm lithium and 30 ccSTP/kg dissolved hydrogen). The IPSs can be provided with adequate instrumentation and be modified to perform valid irradiation studies in a high neutron flux and in a relevant thermos-hydraulic environment. During more than 15 years of operation, activation products have accumulated into the loop leading to a continuous increase of the dose rates at the work area. Consequently periodic maintenance and inspection operations have become more and more expensive in terms of collective dose uptake. In consultation with the internal and external safety authorities the decision has been made to proceed to the chemical closed-loop decontamination of the most important components of CALLISTO (heater, pressurizer, main and bleed flow coolers). The objective of reducing the dose rates without compromising the integrity of the operational loop has led to the combined use of known soft chemical decontamination products as KMnO4 and H2C2O4. About 10 GBq of Co-60 activity and 250 g of corrosion products were removed from the stainless steel CALLISTO loop. The systems involved had a total volume of 0,5 m3 and a surface area of 18 m2. All released activity and corrosion products were removed by ion exchange resins, leading to the generation of 2x150 liters of radioactive waste. The dose rate reduction factors in contact with the treated components varied between 2 and 12. The collective dose uptake of the entire operation (preparation - decontamination - clean-up) was about 5,5 man.mSv, and thereby in line with the ALARA estimations

  2. Issues and remedies for secondary system of PWR/VVER

    International Nuclear Information System (INIS)

    Secondary side degradation of steam generators (SG) and Flow Accelerated Corrosion (FAC) in the secondary system have been for a long time important issues in PWR and VVER types of Nuclear Power Plants. With the evolution of the design, the most important issues are progressively moving from secondary side corrosion of Alloy 600 SG tubing, which is being replaced, to a larger variety of risks associated with potential inadequate chemistries. As far as FAC of carbon steel is concerned, the evolution of treatment selection for minimizing corrosion products transport toward the SG, as well as progressive replacement of components in the feedwater train, decreases the risk of dramatic failures which have occurred in the past. After having briefly explained the reason for the past problems encountered in the secondary system of PWR and VVER, this paper evaluates the risk associated with various impurities or contaminants that may be present in the secondary system and how to mitigate them in the most appropriate, efficient, economical and environmental friendly way. The covered species are sodium, calcium, magnesium, chloride, sulfate and sulfur compounds, fluorides, organic compounds, silica, oxygen, lead, ion exchange resins. This paper also proposes the best remedies for mitigating the new issues that may be encountered in operating plants or units under construction. These are mainly: - Selecting a steam water treatment able to minimize the quantity of corrosion products transported toward the SG; - Mitigating the risk of Flow Induced Vibration by a proper control of deposits in sensitive areas; - Minimizing the risk of concentration of impurities in local areas where they may induce corrosion; - Avoiding the presence of abnormal quantities of some species in SG, such as the detrimental presence of lead and ion exchange resin debris or the controversial presence of organic compounds; - Optimizing costs of maintenance activities (SG mechanical, chemical cleaning

  3. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  4. Reassessment of PWR pressure-vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    A continuing analysis of the PTS problem associated with PWR postuated OCA's indicates that the previously accepted degree of conservatism in the fracture-mechanics model needs to be more closely evaluated, and if excessive, reducted. One feature that was believed to be conservative was the use of two-dimensional as opposed to finite-length (three-dimensional) flaws. A flaw of particular interest is one that is located in an axial weld of a plate-type vessel. For those vessels that suffer relatively high radiation damage in the welds, the length of the flaw will be no greater than the length of the weld, and recent calculations indicate that a deep flaw of that length (approx. 2 m) is not effectively infinitely long, contrary to previous thinking. The benefit to be derived from consideration of the 2-m flaw and also a semielliptical flaw with a length-to-depth ratio of 6/1 was investigated by analyzing several postulated transients. In doing so the sensitivity of the benefit to a specified maximum crack arrest toughness and to the duration of the transient was investigated. Results of the analysis indicate that for some conditions the benefit in using the 2-m flaw is substantial, but it decreases with increasing pressure, and above a certain pressure there may be no benefit, depending on the duration of the transient and the limit on crack arrest toughness

  5. Applicability of oxygenated water chemistry for PWR secondary systems

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P. [Studsvik Nuclear AB, Nykoeping (Sweden); Takiguchi, H.; Otoha, K. [Japan Atomic Power Co., Tokyo (Japan)

    2002-07-01

    Introduction of oxygenated water chemistry (OWC) in PWR secondary side is considered as a means to reduce the transportation of corrosion products into the steam generator and thus also minimizing crevice deposits and subsequent materials problems. One main concern, however, is the risk of inter-granular attack (IGA) in crevices. In order to study effects on crevice tube IGA by OWC, a series of experiments were performed in a steam generator (SG) simulating loop. This comprised a SG tube and a tube support plate (TSP) together forming the crevice. The over-all objective of the work accounted here was to demonstrate that it is possible to operate the steam generator secondary side with OWC without causing intolerable IGA or other types of attack on the tube in the crevice area. Tubes of sensitized Alloy 600 were exposed during a total of nine experiments in an autoclave using a TSP/tube arrangement with an asymmetric crevice design. Experiments were performed at high and low pH and potential under open and packed crevice conditions. The aggressiveness of the crevice environment was also further increased by addition of carbonate and chloride. Furthermore the tube was pressurized. Experimental parameters were monitored on the primary side as well as in the secondary bulk phase and in the crevice. (authors)

  6. Enhancing heat transfer and crud mitigation in PWR fuel

    International Nuclear Information System (INIS)

    This paper discusses three methods for increasing single phase heat transfer in PWR fuel. The primary effect of increasing heat transfer is a reduction in the steaming rate from the fuel rods, which in turn reduces the likelihood of crud formation on the fuel rods and the potential for adsorption of boron into the crud. The advantage of lowering boron mass on the fuel is reduced risk of Axial Offset Anomaly (AOA). Another benefit of reduced crud formation is a lower risk of localized corrosion, a known contributor to rod cladding failures. Thinner crud leads to locally lower rod operating temperatures (lower corrosion rate) since crud acts as a thermal insulator between the rod and the coolant. The first method of increasing heat transfer involves addition of more than one Intermediate Flow Mixing vane grid (IFM) in the span between two neighboring structural spacing grids. The second method includes optimization of the mixing vane according to axial position. The third method involves variation of the IFMs axial position to optimize axial distribution of rod heat transfer. (authors)

  7. Boron mixing transient in a PWR vessel. Physical studies

    International Nuclear Information System (INIS)

    EDF has conducted a R and D action, aiming at gaining more knowledge on vessel thermal-hydraulics; it consists of two complementary approaches based on mock-up experiments and numerical simulations. Maintenance scenarios studies began in 1995. They have been performed solely with the FEM CFD code N3S. The FEM model take into account the U pipe, the primary pump and the cold leg. This mesh can be connected to the vessel mesh used in the study of previous configurations. The first case in progress concerns the influence of the start-up of a boron unsaturated demineralizer. The study concerns the plug formation in the U pipe involved by the clear and cold seal injection water entering the primary circuit. At the end of the diluted water injection the primary pump is started up and the U pipe fluid is sent in the reactor vessel. This paper presents first the CPY 900 MW PWR vessel taken into account in these physical studies, with a special focus on the geometric peculiarities. Then the 1/5. scale BORA-BORA mock-up and the 3D FEM Thermal Hydraulic code N3S are described. The results obtained until now are presented. The degree of achievement of the studies on the three priority cases (start-up, hot shut-down normal operation, cold shut-down normal operation)

  8. PWR loading pattern optimization using Harmony Search algorithm

    International Nuclear Information System (INIS)

    Highlights: ► Numerical results reveal that the HS method is reliable. ► The great advantage of HS is significant gain in computational cost. ► On the average, the final band width of search fitness values is narrow. ► Our experiments show that the search approaches the optimal value fast. - Abstract: In this paper a core reloading technique using Harmony Search, HS, is presented in the context of finding an optimal configuration of fuel assemblies, FA, in pressurized water reactors. To implement and evaluate the proposed technique a Harmony Search along Nodal Expansion Code for 2-D geometry, HSNEC2D, is developed to obtain nearly optimal arrangement of fuel assemblies in PWR cores. This code consists of two sections including Harmony Search algorithm and Nodal Expansion modules using fourth degree flux expansion which solves two dimensional-multi group diffusion equations with one node per fuel assembly. Two optimization test problems are investigated to demonstrate the HS algorithm capability in converging to near optimal loading pattern in the fuel management field and other subjects. Results, convergence rate and reliability of the method are quite promising and show the HS algorithm performs very well and is comparable to other competitive algorithms such as Genetic Algorithm and Particle Swarm Intelligence. Furthermore, implementation of nodal expansion technique along HS causes considerable reduction of computational time to process and analysis optimization in the core fuel management problems

  9. Robots in P.W.R. nuclear powerplants

    International Nuclear Information System (INIS)

    The satisfactory operation of 37 900-MWe PWR powerplants in France, Belgium and South-Africa and the start-up of 1300 MWe powerplants allowed the development of a wide range of automatic units and robots for the periodic maintenance of nuclear plants, reducing the risk of ionizing radiation for the personnel. A large number of automated tools have been built. Among them: - inspection and maintenance systems for the tube bundle of steam generators, - robotized arms ROTETA and ROMEO for the heavy maintenance and delicate operations such as tube extraction or shot peening of tubes to improve their resistance to corrosion; - the versatile manipulator T.A.M. with electrically controlled articulations. The development of functionally versatile tools and robots and the integration of new technologies such as 3-D vision allowed the construction of the self-guided vehicle FRASTAR capable of moving within a nuclear building and in a cluttered environment. This vehicle includes means for avoiding isolated obstacles and can move on stairs

  10. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Directory of Open Access Journals (Sweden)

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  11. Integrated training support system for PWR operator training simulator

    International Nuclear Information System (INIS)

    The importance of operator training using operator training simulator has been recognized intensively. Since 1986, we have been developing and providing many PWR simulators in Japan. We also have developed some training support systems connected with the simulator and the integrated training support system to improve training effect and to reduce instructor's workload. This paper describes the concept and the effect of the integrated training support system and of the following sub-systems. We have PES (Performance Enhancement System) that evaluates training performance automatically by analyzing many plant parameters and operation data. It can reduce the deviation of training performance evaluation between instructors. PEL (Parameter and Event data Logging system), that is the subset of PES, has some data-logging functions. And we also have TPES (Team Performance Enhancement System) that is used aiming to improve trainees' ability for communication between operators. Trainee can have conversation with virtual trainees that TPES plays automatically. After that, TPES automatically display some advice to be improved. RVD (Reactor coolant system Visual Display) displays the distributed hydraulic-thermal condition of the reactor coolant system in real-time graphically. It can make trainees understand the inside plant condition in more detail. These sub-systems have been used in a training center and have contributed the improvement of operator training and have gained in popularity. (author)

  12. Dynamic modelling of PWR fuel assembly for seismic behaviour

    International Nuclear Information System (INIS)

    Vibration and snap back tests have shown that the behaviour of PWR fuel assemblies was non linear : the fuel assembly eigenfrequencies decrease with the excitation level or with the motion amplitude, which was supposed to be due to the slippage of the fuel rods through the grids. Up to now the fuel assembly models were linear and composed by one beam alone representing both the guide thimbles and the fuel rods or by two beams (one for the guide thimbles and one for the fuel rods). The stiffness of such models' were adjusted to fit with the measured eigenfrequency corresponding to a given amplitude. The aim of this paper is to identify the influence of the slippage between grids and fuel rods on the dynamic behaviour of the fuel assembly. For that purpose a non linear fuel assembly model is proposed representing explicitly the slippage phenomenon and is applied to the reduced scale fuel assemblies which have been tested in the framework of a collaboration between FRAMATOME and CEA-DMT. Comparisons between calculations and experiments will be presented and the limitation of this model will be also discussed

  13. Break location effects on PWR small break LOCA phenomena

    International Nuclear Information System (INIS)

    The report presents experimental results of a small lower plenum break test of SB-PV-01 conducted at the large-Scale Test Facility (LSTF) of the Rig-of-Safety Assessment (ROSA)-IV program. This test simulates a loss-of-coolant accident (LOCA) caused by instrument tubes break (break area corresponds to 0.5% of the cold leg flow area) in a Westinghouse-type pressurized water reactor (PWR) assuming both manual actuation for all of the high pressure injection (HPI) systems and failure of the auxiliary feedwater systems. The report clarifies long-term system responses, especially the core cooling conditions related to the primary mass inventory. Also it clarifies break location effects on small break LOCA phenomena by comparing other five similar LOCA tests with break locations at cold leg, hot leg, upper head, pressurizer top (TMI-type) and SG U-tubes. It is coucluded that the lower plenum break is the severest on core heatup due to the highest break flow rate and the least primary mass recovery after the ECCS among the six tests. (author)

  14. Automatic defect identification on PWR nuclear power station fuel pellets

    International Nuclear Information System (INIS)

    This article presents a new automatic identification technique of structural failures in nuclear green fuel pellet. This technique was developed to identify failures occurred during the fabrication process. It is based on a smart image analysis technique for automatic identification of the failures on uranium oxide pellets used as fuel in PWR nuclear power stations. In order to achieve this goal, an artificial neural network (ANN) has been trained and validated from image histograms of pellets containing examples not only from normal pellets (flawless), but from defective pellets as well (with the main flaws normally found during the manufacturing process). Based on this technique, a new automatic identification system of flaws on nuclear fuel element pellets, composed by the association of image pre-processing and intelligent, will be developed and implemented on the Brazilian nuclear fuel production industry. Based on the theoretical performance of the technology proposed and presented in this article, it is believed that this new system, NuFAS (Nuclear Fuel Pellets Failures Automatic Identification Neural System) will be able to identify structural failures in nuclear fuel pellets with virtually zero error margins. After implemented, the NuFAS will add value to control quality process of the national production of the nuclear fuel.

  15. Manufacture of nuclear fuel elements for commercial PWR in China

    International Nuclear Information System (INIS)

    Yibin Nuclear Fuel Element Plant (YFP) under the leadership of China National Nuclear Corporation is sole manufacturer in China to specialize in the production of fuel assemblies and associated core components for commercial PWR nuclear power plant. At the early of 1980's, it began to manufacture fuel assemblies and associated core components for the first core of QINSHAN 300 MW nuclear power plant designed and built by China itself. With the development of nuclear power industry in China and the demand for localization of nuclear fuel elements in the early 1990's, YFP cooperated with FRAMATOME France in technology transfer for design and manufacturing of AFA 2G fuel assembly and successfully supplied the qualified fuel assemblies for the reloads of two units of GUANGDONG Da Ya Bay 900 MW nuclear power plant (Da Ya Bay NPP), and has achieved the localization of fuel assemblies and nuclear power plants. Meanwhile, it supplied fuel assemblies and associated core components for the first core and further reloads of Pakistan CHASHMA 300 MW nuclear power plant which was designed and built by China, and now it is manufacturing AFA 2G fuel assemblies and associated core components for the first core of two units of NPQJVC 600 MW nuclear power plant. From 2001 on, YFP will be able to supply Da Ya Bay NPP with the third generation of fuel assembly-AFA 3G which is to realize a strategy to develop the fuel assembly being of long cycle reload and high burn-up

  16. Effect of coolant chemistry on PWR radiation transport processes

    International Nuclear Information System (INIS)

    The effect of various PWR-type coolant chemistry regimes on the behavior of corrosion products has been studied in the DIDO Water Loop at Harwell. While the extent of in-core spinel deposition is influenced by pH in a manner to be expected from the temperature coefficient of solubility of nickel-iron spinel, there is evidence that boric acid plays a role apart from its influence on pH. Out-of-core deposition of active cobalt on stainless steel takes place largely in the chromium-rich inner oxide layer, and there is also significant uptake of corrosion products into the film on Zircaloy. Deposition depends on flow characteristics in different ways for different elements. The evidence suggests that in DWL soluble species are dominant in out-of-core deposition processes for corrosion products. The adsorption of cobalt in zirconium oxide provides a route for deposition on fuel elements which may in some circumstances be more significant than spinel deposition. A most important factor governing deposition behavior is surface condition; the influence of weld regions and the effect of varying pretreatment conditions have both been demonstrated. Some alternative chemistry regimes have been explored, but do not appear to offer any advantages with respect to activity transport control over the more conventional regime based on lithium hydroxide and hydrogen dosing. 8 refs., 26 figs., 28 tabs

  17. PWR-440 water chemistry optimization to reduce AOA effect

    International Nuclear Information System (INIS)

    The pressure drop increase in PWR-440 is mainly caused by the fact that the coolant contains numerous corrosion products, which are generated after decontamination and deposited in the top part of the fuel assembly as well as by coolant nucleate boiling that under standard water chemistry conditions leads to acceleration of corrosion products deposition and coolant radioactivity growth respectively. The modeling of the pressure drop changes were based on standard data of water chemistry, reactor operating characteristics and fundamental thermodynamic parameters to predict the pressure drop growth. The results of the performed research and modeling of the corrosion products mass transfer processes allowed to qualify relative contribution of thermohydraulic and chemical parameters in the processes and to fulfill the activities as follows: To perform power units operation at water chemistry with maximum permissible alkali metals content. To increase the coolant flow rate through the core; to do so, throttling orifices were replaced and canister-shields were removed. To reduce the number of steam generators to be decontaminated to 2 per year in a single power unit. As a result deposits accumulation in fuel assemblies has been minimized and there is no leakage in the fuel element; reactor thermal output limitation has been eliminated. (author)

  18. Development of a dry storage cask for PWR spent fuel

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power Co., Ltd.(KHNP), which operates all the nuclear power plants in Korea, is developing a new dry storage cask to store twenty four spent fuel assemblies generated from pressurized water reactors for at-reactor or away-from-reactor interim storage facility in Korea. The dry storage cask is designed and evaluated according to the requirements of the IAEA, the US NRC and the Korean regulations for the dry spent fuel storage system. It provides confinement, radiation shielding, structural integrity, subcritical control and passive heat removal for normal and accident conditions. The dry storage cask consists of a dual purpose canister providing a confinement boundary for the PWR spent fuel, and a storage overpack providing a structural and radiological boundary for long-term storage of the canister placed inside it. The overpack is constructed by a combination of steel and concrete, and is equipped with penetrating ducts near its lower and upper extremities to permit natural circulation of air to provide for the passive cooling of the canister and the contained spent fuel assemblies. This paper describes development status, description, design criteria, evaluation and demonstration tests of the dry storage cask. (authors)

  19. Barium silicate glass/Inconel X-750 interaction. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kelsey, Jr., P. V.; Siegel, W. T.; Miley, D. V.

    1980-01-01

    Water reactor safety programs at the Idaho National Engineering Laboratory have required the development of specialized instrumentation. An example is the electrical conductivity-sensitive liquid level transducer developed for use in pressurized-water reactors (PWRs) in which the operation of the sensing probe relies upon the passage of current through the water between the center pin of the electrode and its shell such that when water is present the resulting voltage is low, and conversely, when water is absent the voltage is high. The transducer's ceramic seal is a hot-pressed glass ceramic; its metal housing is Inconel X-750. The ceramic material provides an essential dielectric barrier between the center pin and the outer housing. The operation of the probe as well as the integrity of the PWR environment requires a hermetically-bonded seal between the ceramic and the metal. However, during testing, an increasing number of probe assemblies failed owing to poor glass-to-metal seals as well as void formation within the ceramic. Therefore, a program was initiated to characterize the metallic surface with respect to pre-oxidation treatment and determine optimum conditions for wetting and bonding of the metal by the glass to obtain baseline data relevant to production of acceptable transducer seals.

  20. PWR safety/relief valve blowdown analysis experience

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.Z.; Chou, L.Y.; Yang, S.H. (Gilbert/Commonwealth Engineers and Consultants, Reading, PA (USA). Speciality Engineering Dept.)

    1982-10-01

    The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively. General approaches for generating forcing functions from thermal fluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the 'acceleration or wave force' method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawbacks are discussed. Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.

  1. Advanced methods for the study of PWR cores

    International Nuclear Information System (INIS)

    This document gathers the transparencies presented at the 6. technical session of the French nuclear energy society (SFEN) in October 2003. The transparencies of the annual meeting are presented in the introductive part: 1 - status of the French nuclear park: nuclear energy results, management of an exceptional climatic situation: the heat wave of summer 2003 and the power generation (J.C. Barral); 2 - status of the research on controlled thermonuclear fusion (J. Johner). Then follows the technical session about the advanced methods for the study of PWR reactor cores: 1 - the evolution approach of study methodologies (M. Lambert, J. Pelet); 2 - the point of view of the nuclear safety authority (D. Brenot); 3 - the improved decoupled methodology for the steam pipe rupture (S. Salvatores, J.Y. Pouliquen); 4 - the MIR method for the pellet-clad interaction (renovated IPG methodology) (E. Baud, C. Royere); 5 - the improved fuel management (IFM) studies for Koeberg (C. Cohen); 6 - principle of the methods of accident study implemented for the European pressurized reactor (EPR) (F. Foret, A. Ferrier); 7 - accident studies with the EPR, steam pipe rupture (N. Nicaise, S. Salvatores); 8 - the co-development platform, a new generation of software tools for the new methodologies (C. Chauliac). (J.S.)

  2. Effect of ethanolamine injection on wall thinning rate of PWR carbon steel components

    International Nuclear Information System (INIS)

    For pH control of PWR secondary system water chemistry, some plants have changed to ethanolamine injection. The purpose of this work was to understand the effect of changing water chemistry on wall thinning rate of the PWR secondary system due to flow accelerated corrosion. For that purpose, evaluations of water chemistry were carried out by a mass balance calculation, wall thinning rate measurement by a rotary disk test and wall thinning rate evaluation based on calculated magnetite solubility. As a result, it was found to be effective to inhibit the wall thinning rate of the PWR secondary system due to flow accelerated corrosion by ethanolamine injection, but it was not sufficiently effect to neglect wall thinning rate due to flow accelerated corrosion. The effect of the wall thinning rate inhibition also varied greatly for each component of the PWR secondary system. It was found that maintenance of the carbon steel used for the PWR secondary system was still required under ethanolamine injection condition. (author)

  3. Survey of experiments and code development for the passive residual heat removal system of PWR in China

    Institute of Scientific and Technical Information of China (English)

    HUANG Yan-Ping; ZHUO Wen-Bing; YANG Zu-Mao; XIAO Ze-Jun; CHEN Bing-De; JIA Dou-Nan

    2004-01-01

    Three different kinds of experiments and their typical results are surveyed for the passive residual heat removal system (PRHRS) of PWR performed in Nuclear Power Institute of China (NPIC) recent ten years. The typical results of MISAP. a special code for PWR passive residual heat removal system developed and assessed by NPIC,are also described briefly in this paper.

  4. Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Putney, J.M.; Preece, R.J. [National Power, Leatherhead (GB). Technology and Environment Centre

    1993-06-01

    An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3.

  5. Development of a lead extrusion damper for PWR reactor coolant loop system

    International Nuclear Information System (INIS)

    Conventional seismic design for PWR reactor coolant loop system is conducted under a philosophy of rigid design and large site of rigid supports and many snubbers are used as seismic supports. But recently various type of alternative supports to snubbers have been proposed. A lead extrusion damper (LED) is one of the devices being considered. This paper is devoted to experimental and analytical work on the development of the LED for PWR reactor coolant loop system. In the study, the fundamental mechanism of the damper and the damping effect on the response of a steam generator supported by the LED were studied. From experimental and analytical approaches, the feasibility of application of the LED to PWR reactor coolant loop system was confirmed

  6. Chinese Cooking.

    Science.gov (United States)

    Kane, Tony

    This unit, intended for secondary level students, is a general introduction to Chinese cooking. It is meant to inform students about the origins of Chinese cooking styles in their various regional manifestations, and it can be used to discuss how and why different cultures develop different styles of cooking. The first part of the unit, adapted…

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

    International Nuclear Information System (INIS)

    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. Supplementary, it is recommended to consider amendments to existing water chemistry

  8. Analyses of PWR boron dilution consequences with the Arrotta code

    Energy Technology Data Exchange (ETDEWEB)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced.

  9. Effect of water chemistry on deposition for PWR plant operation

    International Nuclear Information System (INIS)

    For Pressurized Water Reactor (PWR) operation, water chemistry guidelines, specifications and associated surveillance programs are key to avoid deposition of oxides. Deposition of oxides can be detrimental by disrupting results of flow measurements, decreasing the thermal exchange capacity, or even by impairing safety. This paper describes the most important cases of deposition, their consequences for operation, and the implemented improvements to avoid their reoccurrence. Deposition that led to a Crud Induced Power Shift (CIPS) is also described. In the primary and in the secondary sides, orifice plates are typically used for measuring feedwater flow rate in nuclear power plants. Feedwater flow rates are used for control purposes and are important safety parameters as they are used to determine the plant's operating power level. Fouling of orifice plates in the primary side has been found during surveillance testing. For reactor coolant pumps, the formation of deposits on the seal No.1 can cause abnormally high or low leak rates through the seal. The leak rate through this seal must be carefully maintained within a prescribed range during plant operation. In the secondary side, orifice plate fouling has been the cause of feedwater flow/reference thermal power drift. For the steam generators (SG), magnetite deposition has led to fouling of the tube bundle, clogging of the quadri-foiled support plate holes and hard sludge formation on the base plate. For the generators, copper hollow conductors are widely used. Buildup of copper oxides on the interior walls of copper conductors has caused insufficient heat transfer. All these deposition cases have received adequate attention, understanding and response via improvement of our surveillance programs. (authors)

  10. Evaluation of PWR steam generator water hammer. Final technical report, June 1, 1976--December 31, 1976

    International Nuclear Information System (INIS)

    An investigation of waterhammer in the main feedwater piping of PWR steam generators due to water slugs formed in the steam generator feedring is reported. The relevant evidence from PWR operation and testing is compiled and summarized. The state-of-the-art of analysis of related phenomena is reviewed. Original exploratory modeling experiments at 1/10 and 1/4 scale are reported. Bounding analyses of the behavior are performed and several key phenomena have been identified for the first time. Recommendations to the Nuclear Regulatory Commission are made

  11. Application of diffusion theory methods to PWR [pressurized water reactors] analysis

    International Nuclear Information System (INIS)

    In-core physics analysis of pressurized light water reactors (PWRs) requires accurate predictions of three-dimensional pin-by-pin power distributions. The PWR analyses must rely on diffusion theory approximation because no practical methods exist for performing routine three-dimensional pin-by-pin transport calculations. Pin-by-pin diffusion calculations are also prohibitively expensive in three-dimensional geometry, and PWR analyses utilize either two-dimensional pin-by-pin models or three-dimensional advanced nodal models. The purpose of this paper is to detail and contrast approximations required by pin-by-pin and nodal diffusion methods

  12. Contribution to the study of the conversion PWR type reactors to the thorium cycle

    International Nuclear Information System (INIS)

    The use of the thorium cycle in PWR reactors is discussed. The fuel has been calculated in the equilibrium condition for a economic comparison with the uranium cycle (in the same condition). First of all, a code named EQUILIBRIO has been developed for the fuel equilibrium calculation. The results gotten by this code, were introduced in the LEOPARD code for the fuel depletion calculation (in the equilibrium cycle). Same important physics details of fuel depletion are studied, for instance: the neutron balance, power sharing, fuel burnup, etc. The calculations have been done taking as reference the Angra-1 PWR reactor. (Author)

  13. Assessment of options for the treatment of Sizewell PWR liquid effluent

    International Nuclear Information System (INIS)

    This report describes the origins of PWR liquid waste streams, their composition and rates of arising. Data has been collected from operational PWRs and estimates obtained for Sizewell B PWR liquid waste streams. Current liquid waste treatment practices are reviewed and assessments made of established and novel treatment techniques which could be applicable to Sizewell B. A short list of treatment options is given and recommendations are made relating to established treatment technologies suitable for Sizewell B and also to development work on more novel treatments which could lead to a reduction in waste disposal volumes. (author)

  14. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  15. Behavior of a PWR-containment under rising internal pressure load

    International Nuclear Information System (INIS)

    Reactor safety containments are dimensioned so that in a postulated design accident (fracture of the primary duct), the coolant flowing out can be reliably accommodated. The internal pressure in German PWR's is about 5 bar. Radioactive contamination of the environment is largely avoided in this way. The experiments were done for the safety containment of the PWR at Philippsburg. It is a freestanding spherical shell 56 metres in diameter and 38 mm thick. The tensioning of the concrete foundations is 400 below the equator. The spherical shell is welded from about 500 curved sheets made of 15 MnNi 63 material. (orig./GL)

  16. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Xiangjie Liu

    2014-01-01

    Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

  17. PWR Secondary Water Chemistry Control Status: A Summary of Industry Initiatives, Experience and Trends Relative to the EPRI PWR Secondary Water Chemistry Guidelines

    International Nuclear Information System (INIS)

    The latest revision of the EPRI Pressurized Water Reactor (PWR) Secondary Water Chemistry Guidelines was issued in February 2009. The Guidelines continue to focus on minimizing stress corrosion cracking (SCC) of steam generator tubes, as well as minimizing degradation of other major components / subsystems of the secondary system. The Guidelines provide a technically-based framework for a plant-specific and effective PWR secondary water chemistry program. With the issuance of Revision 7 of the Guidelines in 2009, many plants have implemented changes that allow greater flexibility on startup. For example, the previous Guidelines (Revision 6) contained a possible low power hold at 5% power and a possible mid power hold at approximately 30% power based on chemistry constraints. Revision 7 has established a range over which a plant-specific value can be chosen for the possible low power hold (between 5% and 15%) and mid power hold (between 30% and 50%). This has provided plants the ability to establish significant plant evolutions prior to reaching the possible power hold; such as establishing seal steam to the condenser, placing feed pumps in service, or initiating forward flow of heater drains. The application of this flexibility in the industry will be explored. This paper also highlights the major initiatives and industry trends with respect to PWR secondary chemistry; and outlines the recent work to effectively address them. These will be presented in light of recent operating experience, as derived from EPRI's PWR Chemistry Monitoring and Assessment (CMA) program (which contains more than 400 cycles of operating chemistry data). (authors)

  18. Chinese Confucianism

    OpenAIRE

    Macfarlane, Alan; Cullen, Christopher

    2004-01-01

    Confucianism has deeply influenced Chinese civilization. Christopher Cullen describes its effect on education, social structure and knowledge over the past centuries, against the backdrop of a Confucian building in Beijing.

  19. Steam Generator Chemical Cleaning Application: Korean Experience in PWR NPP

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power (KHNP) performed an EPRI/SGOG chemical cleaning of the secondary side of the steam generators at Ulchin Unit 3 (UCN3) in March 2011 and at Ulchin Unit 4 (UCN4) in September 2011. The steam generator chemical cleaning (SGCC) was performed with venting at the top-of-tube sheet (TTS) and at tube support plates (TSPs) 4, 5, 6, 7, 8, 9, and 10. A primary objective of this SGCC was to address outer diameter stress corrosion cracking (ODSCC), which has been observed at the TTS and TSPs in the UCN3 SGs. The EPRI/SGOG process has been shown to effectively reduce prevailing ODSCC rates at the TTS and TSPs, particularly when applied with periodic venting in this application. This was the first full-length SGCC campaign with venting performed in Korea. Ulchin Unit 3 commenced commercial operation in August 1998 and Ulchin Unit 4 commenced commercial operation in December 1999. UCN3 and UCN4 are a two-loop pressurized water reactor (PWR) of the Korea Standard Nuclear Plant (KSNP) design. The SGs contain high-temperature mill annealed (HTMA) Alloy 600 tubing and are similar in design to the Combustion Engineering CE-80. The KSNP SGs have been susceptible to outer diameter stress corrosion cracking (ODSCC), which is consistent with operating experience for other SGs containing Alloy 600HTMA tubing material. The UCN3/4 SGs have recently begun to experience ODSCC. Hankook Jungsoo Industries Co., Ltd (HaJI) was selected as the cleaning vendor by KHNP. To date, HaJI has completed five Advanced Scale Conditioning Agent (ASCA) cleaning applications and two EPRI/SGOG Steam Generator Chemical Cleaning (SGCC) campaigns for KHNP. The goal of total deposit removal of the applications were successfully achieved and the amounts are 3,579 kg at UCN3 and 3,786 kg at UCN4 which values were estimated before each cleaning by analysing ECT signal and liquid samples from the SGs. The deposits from the SGs were primarily composed of magnetite. There were no chemical

  20. An investigation into the efficiency of ion-exchange membranes in simulated PWR coolants

    International Nuclear Information System (INIS)

    This report describes an investigation of the retention efficiency of cation-exchange membranes for magnesium, calcium and nickel ions in PWR-coolant type solutions containing 2 ppm lithium (as lithium hydroxide) and 1000 ppm boron (as boric acid). By analysis of the membranes themselves or of the effluent, the retention characteristics of the membranes in various experimental conditions have been examined. (author)

  1. Radiation embrittlement of Sizewell 'B' PWR pressure vessel during a 40 year lifetime

    International Nuclear Information System (INIS)

    Irradiation embrittlement data produced from PWR surveillance and materials test reactor accelerated experiments on low alloy Mn Mo Ni pressure vessel steels, which satisfy the Sizewell 'B' forging materials specification on copper (<= 0.09% wt.%) and nickel (<= 0.85 wt.%), have been examined. (author)

  2. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  3. Response of pressurized water reactor (PWR) to network power generation demands

    International Nuclear Information System (INIS)

    The flexibility of the PWR type reactor in terms of response to the variations of the network power demands, is demonstrated. The factors that affect the transitory flexibility and some design prospects that allow the reactor fits the requirements of the network power demands, are also discussed. (M.J.A.)

  4. PWR nozzle 'crotch corner' inspection: an effective additional ultrasonic technique for radial cracks

    International Nuclear Information System (INIS)

    An ultrasonic non-destructive technique for testing the integrity of the nozzle crotch corner of a PWR pressure vessel is described which uses two angled probes to detect the specular reflection from one probe to the other via a crack lying in the important radial plane of the nozzle. (U.K.)

  5. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    International Nuclear Information System (INIS)

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO2 fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  6. Neutronic Analysis of Advanced SFR Burner Cores using Deep-Burn PWR Spent Fuel TRU Feed

    International Nuclear Information System (INIS)

    In this work, an advanced sodium-cooled fast TRU (Transuranics) burner core using deep-burn TRU feed composition discharged from small LWR cores was neutronically analyzed to show the effects of deeply burned TRU feed composition on the performances of sodium-cooled fast burner core. We consider a nuclear park that is comprised of the commercial PWRs, small PWRs of 100MWe for TRU deep burning using FCM (Fully Ceramic Micro-encapsulated) fuels and advanced sodium-cooled fast burners for their synergistic combination for effective TRU burning. In the small PWR core having long cycle length of 4.0 EFPYs, deep burning of TRU up to 35% is achieved with FCM fuel pins whose TRISO particle fuels contain TRUs in their central kernel. In this paper, we analyzed the performances of the advanced SFR burner cores using TRU feeds discharged from the small long cycle PWR deep-burn cores. Also, we analyzed the effect of cooling time for the TRU feeds on the SFR burner core. The results showed that the TRU feed composition from FCM fuel pins of the small long cycle PWR core can be effectively used into the advanced SFR burner core by significantly reducing the burnup reactivity swing which reduces smaller number of control rod assemblies to satisfy all the conditions for the self controllability than the TRU feed composition discharged from the typical PWR cores

  7. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  8. Application of the BEACON-TSM system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    BEACON-TSM is an advanced system of the operation support of PWR reactors that combines the capabilities of an advanced nodal neutronic model and the measures of the instrumentation available in plant to determine, accurately and continuously, the distribution of power in the core and the available margins to the limits of the beak factors.

  9. SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

    International Nuclear Information System (INIS)

    Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.)

  10. ''The place of the fatigue risks in the PWR maintenance programs''

    International Nuclear Information System (INIS)

    The parts of components submitted to fatigue risk are more particularly controlled in operation. Three main cases are identified: the mechanical oligo-cyclic fatigue, the vibrating fatigue and the thermal fatigue. These cases are presented in this paper. As a precaution a complementary investigation program is implementing during the Number two decennial inspections of the 900 MW PWR. (A.L.B.)

  11. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    International Nuclear Information System (INIS)

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables

  12. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  13. Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

    International Nuclear Information System (INIS)

    As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

  14. Assessment of void swelling in austenitic stainless steel PWR core internals.

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. M.; Energy Technology

    2006-01-31

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling

  15. Analysis of difficulties accounting and evaluating nuclear material of PWR fuel plant

    International Nuclear Information System (INIS)

    Background: Nuclear materials accountancy must be developed for nuclear facilities, which is required by regulatory in China. Currently, there are some unresolved problems for nuclear materials accountancy of bulk nuclear facilities. Purpose: The retention values and measurement errors are analyzed in nuclear materials accountancy of Power Water Reactor (PWR) fuel plant to meet the regulatory requirements. Methods: On the basis of nuclear material accounting and evaluation data of PWR fuel plant, a deep analysis research including ratio among random error variance, long-term systematic error variance, short-term systematic error variance and total error involving Material Unaccounted For (MUF) evaluation is developed by the retention value measure in equipment and pipeline. Results: In the equipment pipeline, the holdup estimation error and its total proportion are not more than 5% and 1.5%, respectively. And the holdup estimation can be regraded as a constant in the PWR nuclear material accountancy. Random error variance, long-term systematic error variance, short-term systematic error variance of overall measurement, and analytical and sampling methods are also obtained. A valuable reference is provided for nuclear material accountancy. Conclusion: In nuclear material accountancy, the retention value can be considered as a constant. The long-term systematic error is a main factor in all errors, especially in overall measurement error and sampling error: The long-term systematic errors of overall measurement and sampling are considered important in the PWR nuclear material accountancy. The proposals and measures are applied to the nuclear materials accountancy of PWR fuel plant, and the capacity of nuclear materials accountancy is improved. (authors)

  16. Crack growth rates of nickel alloy welds in a PWR environment.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.; Energy Technology

    2006-05-31

    In light water reactors (LWRs), vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This report presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320 C. Crack growth tests were conducted on 1-T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The CGRs of Alloy 182 in the PWR environment are a factor of {approx}5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking for the Alloy 182 weld is close to the average behavior of Alloy 600 in the PWR environment. The weld orientation was found to have a profound effect on the magnitude of crack growth: cracking was found to propagate faster along the dendrites than across them. The existing CGR data for Ni-alloy weld metals have been compiled and evaluated to establish the effects of key material, loading, and environmental parameters on CGRs in PWR environments. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking.

  17. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  18. Design Development and Verification of a System Integrated Modular PWR

    International Nuclear Information System (INIS)

    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

  19. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

  20. Development and application of methods and computer codes of fuel management and nuclear design of reload cycles in PWR

    International Nuclear Information System (INIS)

    Description of methods and computer codes for Fuel Management and Nuclear Design of Reload Cycles in PWR, developed at JEN by adaptation of previous codes (LEOPARD, NUTRIX, CITATION, FUELCOST) and implementation of original codes (TEMP, SOTHIS, CICLON, NUDO, MELON, ROLLO, LIBRA, PENELOPE) and their application to the project of Management and Design of Reload Cycles of a 510 Mwt PWR, including comparison with results of experimental operation and other calculations for validation of methods. (author)

  1. Techniques and devices developed by the CEA for hot cell and in-situ examinations of PWR components and PWR fuel assembliess after irradiation

    International Nuclear Information System (INIS)

    Within the framework of the electro-nuclear development of the PWR system, the CEA has provided itself with facilities for developing techniques for analyzing assemblies, pins and fuels. These are examinations and tests on irradiated heads and assemblies with the aid of the Fuel Examination Module (FEM), of machining of assemblies and examinations in the Celimene hot laboratory or detailed examinations and analyses on fuel elements using eddy currents, the electronic microprobe and the Fisher ''permeascope'' which enables the outline of the oxide coat present on the cladding to be followed

  2. Management of radioactive waste nuclear power plants

    International Nuclear Information System (INIS)

    The authors give a survey of the sources, types and amounts of radioactive waste in LWR nuclear power stations (1,300 MWe). The amount of solid waste produced by a Novovorenezh-type PWR reactor (2 x 400 resp. 1 x 1,000 MWe) is given in a table. Treatment, solidification and final storage of radioactive waste are shortly discussed with special reference to the problems of final storage in the CSR. (HR)

  3. Waste management and disposal I

    International Nuclear Information System (INIS)

    The author gives a survey of the nuclear fuel cycle and of the type and amount of the radioactive wastes as developing within the fuel cycle. The input/output data and the yearly waste production of a 1,300 MWe BWR reactor and PWR reactor are shown in tabular form. The possible dangers for man caused by the radioactive waste are also mentioned. (HR)

  4. Chinese Dream

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    The general managers of South Korean auto giants Hyundai and Kia have high hopes for the growing Chinese auto market. Both companies went through a painstaking period as the financial crisis first roared across the globe. Jin Shan-fa, General Manager of Hyundai Motor Group

  5. An Empirical Approach to Bounding the Axial Reactivity Effects of PWR Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    P. M. O' Leary; J. M. Scaglione

    2001-04-04

    One of the significant issues yet to be resolved for using burnup credit (BUC) for spent nuclear fuel (SNF) is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters (such as local power, fuel temperature, moderator temperature, burnable poison rod history, and soluble boron concentration) affect the isotopic inventory of fuel that is depleted in a pressurized water reactor (PWR). However, obtaining the detailed operating histories needed to model all PWR fuel assemblies to which BUC would be applied is an onerous and costly task. Simplifications therefore have been suggested that could lead to using ''bounding'' depletion parameters that could be broadly applied to different fuel assemblies. This paper presents a method for determining a set of bounding depletion parameters for use in criticality analyses for SNF.

  6. RANS modeling for flow in nuclear fuel bundle in pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    This paper presents use of Reynolds-Averaged Navier-Stokes (RANS) based turbulence model for single-phase CFD analysis of flow in Pressurized Water Reactor (PWR) Assemblies. An open source code called OpenFoam was used for computational fluid dynamics (CFD) study using computational meshes generated using Shari Harpoon. The PWR assembly design used in this analysis represents a 5 x 5 pin design including structural grid equipped with mixing vanes. The design specifications used in this study were obtained from the experimental setup at Texas A&M University and the results obtained are used to validate the CFD software, algorithm, and the turbulence model used in this analysis. (author)

  7. Improving electron beam weldability of heavy steel plate for PWR-steam generator

    International Nuclear Information System (INIS)

    There are plans to install or replace many PWR-steam generators. Electron beam welding (EBW) can greatly reduce the welding period compared to conventional welding methods (narrow-gap GMAW and SAW). The problems in applying EBW are to prevent weld defects and to improve the toughness of the weld metal. Successful defect-free welding procedures were established for thick steel plates. As for weld metal toughness, we investigated the factors that deteriorate weld metal toughness of EBW and made clear the manufacturing process which utilizes a new secondary refining process and a high-torque mill in actual mass-production. As a result, application of EBW to PWR-steam generators has become possible and large amounts of ASTM A533B C12 (JIS SQV2B) steel plates for EBW have come to be produced. We evaluated EBW base metal and weld joints including fracture toughness. (author). 12 refs., 8 figs., 6 tabs

  8. Optimization of small long-life PWR based on thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Subkhi, Moh Nurul, E-mail: nsubkhi@students.itb.ac.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology. Jalan Ganesha 10, Bandung (Indonesia); Physics Dept., Faculty of Science and Technology, State Islamic University of Sunan Gunung Djati Bandung Jalan A.H Nasution 105 Bandung (Indonesia); Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Waris, Abdul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology. Jalan Ganesha 10, Bandung (Indonesia)

    2015-09-30

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% {sup 233}U & 2.8% {sup 231}Pa, 6% {sup 233}U & 2.8% {sup 231}Pa and 7% {sup 233}U & 6% {sup 231}Pa give low excess reactivity.

  9. Conceptual design study of small long-life PWR based on thorium cycle fuel

    Energy Technology Data Exchange (ETDEWEB)

    Subkhi, M. Nurul [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) and Physics Dept., Faculty of Science and Technology, State Islamic University of Sunan Gunung (Indonesia); Su' ud, Zaki; Waris, Abdul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung) (Indonesia)

    2014-09-30

    A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of {sup 233}U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation.

  10. Review of the high conversion-type core study. Review about PWR

    International Nuclear Information System (INIS)

    The study of the high conversion light water reactor was proposed by Edlund in the United States of America in 1975. This theme was aggressively studied in the 1980s. As the reason, the increase of Pu produced from the reprocessing and the delay of practical use of FBR is given. A high converter core can realize comparatively easily by changing core of existing PWR to the tight and short core. In this report, the high converter core study about PWR were reviewed. In the United States of America, the study has already ended but an aggressive study is carried forward in Germany, Japan and so on. In addition to the reactor physics computation, the wide range study of such as critical experiment, conduct experiment of the heat transfer, the fracture behavior of fuel and the reactor-type strategy are carried forward. To investigate these studies is extremely useful in examining a future. (author)

  11. Crud formation on low duty PWR fuel in the Halden reactor

    International Nuclear Information System (INIS)

    A previous paper summarised observations on the effects of water chemistry and thermal-hydraulic conditions on crud formation on PWR fuel in the Halden reactor. These observations led to the conclusion that a critical degree of fuel duty (which can be expressed as degree of coolant sub-cooled boiling, void fraction or mass evaporation rate) was required for the formation of tenacious crud deposits. Recent measurements of the oxide layers on low duty PWR fuel have revealed the formation of tenacious crud deposits. This paper describes the operating history of the fuel rods, including water chemistry and thermal-hydraulic conditions, and suggests reasons for the sudden appearance of the crud deposits. (author)

  12. Estimation of the hydrodynamic effects of a LOCA in A 4-loop PWR

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. [CEA Saclay, SEMT, 91 - Gif sur Yvette (France); Potapov, S. [Electricite de France (EDF-SEP / AMV), 92 - Clamart (France); Tephany, F. [Electricite de France (EDF SEPTEN), 69 - Villeurbanne (France)

    2001-07-01

    The PWR safety studies involve an analysis of the consequences of a hypothetical rupture of a primary pipe. From the opening tune, the blowdown at the break causes the propagation of an acoustic wave through the whole primary circuit, as well as pipe whipping. The local pressure gaps due to the depressurization wave propagation may induce component recoils and internal structure movements. In parallel with the acoustic wave propagation, the circuit empties progressively first with a monophasic regime and later with a diphasic one. This paper presents a hydrodynamic simulation of the flows in the primary circuit of 4-loop PWR during a LOCA. The results concern the propagation of the depressurization acoustic wave along the circuit, coupled with the transient fluid flows. (authors)

  13. EPRI PWR Safety and Relief Value Test Program: safety and relief valve test report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    A safety and relief valve test program was conducted by EPRI for a group of participating PWR utilities to respond to the USNRC recommendations documented in NUREG 0578 Section 2.1.2, and as clarified in NUREG 0737 Item II.D.1.A. Seventeen safety and relief valves representative of those utilized in or planned for use in participating domestic PWR's were tested under the full range of selected test conditions. This report contains a listing of the selected test valves and the corresponding as tested test matrices, valve performance data and principal observations for the tested safety and relief valves. The information contained in this report may be used by the participating utilities in developing their response to the above mentioned USNRC recommendations.

  14. Dosimetry experiment 'Dompac'. Neutronic simulation of the thickness of a PWR pressure vessel. Irradiation damage characterization

    International Nuclear Information System (INIS)

    For suitable extrapolation of irradiated PWR ferritic steel results, proper irradiation of the pressure vessel has been 'simulated' in test reactor. For this purpose, a huge steel block (20 cm in depth) was loaded with Saclay's graphite (GAMIN) and tungsten damage detectors. Core-block water gap was optimized through spectrum indexes method, by ANISN and SABINE codes so that spectrum in 1/4 thickness matches with ANISN computations for PWR Fessenheim 1. A good experimental agreement is found with calculated dpa damage gradient. 3D Monte Carlo computation (TRIPOLI) was performed on the DOMPAC device, and spectrum indexes evolution was found consistent with experimental results. Surveillance rigs behind a 'thermal shield' were also simulated, including damage and activation monitors. Dosimetry results give an order of magnitude of accuracies involved in projecting steel sample embrittlement to the pressure vessel

  15. Aerosol removal by emergency spray in PWR containment: synthesis of the TOSQAN aerosol tests

    International Nuclear Information System (INIS)

    During the course of a severe accident in a nuclear Pressurized Water Reactor (PWR), containment reactor is pressurized by steam and hydrogen released from a primary circuit breach and distributed into the containment according to convective flows and steam wall condensation. In addition, core degradation leads to fission product release into the containment. Water spraying is used in the containment as mitigation means in order to reduce pressure, to remove fission products and to enhance the gas mixing in case of presence of hydrogen. This paper presents the synthesis of the results of the TOSQAN aerosol program undertaken by the Institut de Radioprotection et de Surete Nucleaire (IRSN) devoted to study the aerosol removal by a spray, for typical accidental thermal hydraulic conditions in PWR containment. (author)

  16. EDF/CIDEN - ONECTRA: PWR decontamination; EDF/CIDEN - ONECTRA: assainissement REP

    Energy Technology Data Exchange (ETDEWEB)

    Fayolle, P. [EDFICIDEN, 35-37, rue Louis Guerin - B.P. 21212, 69611 Villeurbanne Cedex (France); Orcel, H. [ONECTRA, ZA les Tomples BP45, 26701 Pierrelatte Cedex (France); Wertz, L. [ONECTRA, Le Britannia, Allee C, 20 Bd Eugene Deruelle, 69432 Lyon Cedex 03 (France)

    2010-07-01

    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

  17. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design

    International Nuclear Information System (INIS)

    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

  18. Problems of control of WWER-type pressurized water reactors (PWR's)

    International Nuclear Information System (INIS)

    The problems are dealt with of nuclear power reactor control. Special attention is paid to the reactor of the WWER type, which will play the most important part in the Czechoslovak power system in the near future. The subsystems are described which comprise the systems of reactor control and protection. The possibilities are outlined of using Czechoslovak instrumentation for the control and safety system of the WWER-type PWR. (author)

  19. Gamma-ray measurements of spent PWR fuel and determination of residual power

    OpenAIRE

    Jansson, Peter; Håkansson, Ane; Bäcklin, Anders

    1997-01-01

    The method for determining residual thermal power in spent BWR fuel described in ISV-4/97 have been used in an extended study where spent PWR fuel assemblies have been considered. The experimental work has been carried out at the interim storage CLAB. By using the 137Cs radiation it is shown in the present study that it is possible to experimentally determine the residual thermal power within 3%.

  20. Chemical and radiochemical specifications - PWR power plants; Specifications chimiques et radiochimiques - Centrales REP

    Energy Technology Data Exchange (ETDEWEB)

    Stutzmann, A. [Electricite de France (EDF), 93 - Saint-Denis (France)

    1997-07-01

    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.)

  1. Anti -corrosion Effect of ETA on Materials in Secondary Loop of PWR

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    In the world, over sixty percent of nuclear power plant have used advanced amunes ETA(Ethanolamine) as pH control agent in secondary loop of PWR. There are eighty percent of nuclear powerplants using ETA in USA. The corrosion of materials in steam generator (SG) tube and secondary looppower water reactor have been inhibited, the life of SG and the economics of the plant are increasedbecause of using ETA.

  2. Methods and computer programs for PWR's fuel management: Programs Sothis and Ciclon

    International Nuclear Information System (INIS)

    Methos and computer programs developed at JEN for fuel management in PWR are discussed, including scope of model, procedures for sistematic selection of alternatives to be evaluated, basis of model for neutronic calculation, methods for fuel costs calculation, procedures for equilibrium and trans[tion cycles calculation with Soth[s and Ciclon codes and validation of methods by comparison of results with others of reference (author) '

  3. STUDY OF THE THERMAL STRATIFICATION IN PWR REACTORS AND THE PTS (PRESSURIZED THERMAL SHOCK) PHENOMENON

    OpenAIRE

    ROMERO HAMERS, ADOLFO

    2014-01-01

    In the event of hypothetical accident scenarios in PWR, emergency strategies have to be mapped out, in order to guarantee the reliable removal of decay heat from the reactor core, also in case of component breakdown. One essential passive heat removal mechanism is the reflux condensation cooling mode. This mode can appear for instance during a small break loss-of-coolant-accident (LOCA) or because of loss of residual heat removal (RHR) system during mid loop operation at plant outage after th...

  4. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  5. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    Science.gov (United States)

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

    2014-06-01

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  6. PWR ENDF/B-VII Cross-Section Libraries for ORIGEN-ARP

    International Nuclear Information System (INIS)

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.

  7. RNL NDT studies related to PWR pressure vessel inlet nozzle inspection

    International Nuclear Information System (INIS)

    Non-destructive examinations of the Reactor Pressure Vessel (RPV) of a Pressurized Water Reactor (PWR) play an important role in assuring vessel integrity throughout its operational life. Automated ultrasonic techniques for the detection and sizing of flaws in thick-section seam welds and near-surface regions in a PWR RPV have been under development at RNL for some time. Techniques for the inspection of complex geometry welds and other regions of the vessel are now being assessed and further developed as part of the UK NDT development programme in support of the Sizewell PWR. One objective of this programme is to demonstrate that the range of ultrasonic techniques already shown to be effective for the inspection of seam welds and inlet nozzle corner regions, through exercises such as the Defect Detection Trials, can also be effective for inspection of these other vessel regions. The nozzle-to-vessel welds and nozzle crotch corners associated with the RPV water inlet and outlet nozzles are two such regions being examined in this programme. In this paper, a review is given of the work performed at RNL in the development of a laboratory-based inspection system for inlet nozzle inspection. The main features of the system in its current stage of development are explained. (author)

  8. Improving electron beam weldability of heavy steel plates for PWR-steam generator

    International Nuclear Information System (INIS)

    Installation and replacement of many PWR-steam generators are planned inside and outside Japan. The steel plates for steam generators are heavy in thickness, and increase the number of welding passes and prolong the welding time. Electron beam welding (EBW) can greatly reduce the welding period compared with conventional welding methods (narrow-gap gas metal arc welding (GMAW) and submerged arc welding (SAW)). The problems in applying EBW are to prevent weld defects and to improve the toughness of the weld metal. Defect-free welding procedures were successfully established even in thick steel plates. The factors that deteriorate weld-metal (WM) toughness of EBW were investigated. The manufacturing process, which utilizes a new secondary refining process at steelmaking and a high-torque mill at plate mill in actual mass-production, were established. EBW base metal and WM have better properties including fracture toughness than those of conventional welding processes. As a result, an application of EBW to the fabrication of PWR-steam generators has become possible. Large amounts of ASTM A533 Gr B Cl 2 (JIS SQV2B) steel plates in actual PWR-steam generators have come to be produced (more than 1,500 ton) by applying EBW. (author)

  9. Development of mechanical test techniques for structural components of irradiated PWR fuel assembly

    International Nuclear Information System (INIS)

    An increase of fuel burnup and duration of fuel life remains one of the main methods for a nuclear power engineering enhancement. Properties of structural materials providing corrosion resistance, mechanical strength, and dimensional instability of the components of a fuel assembly (FA) are of great importance for fuel operational reliability in such fuel life cycles. Generally, PWR fuel assemblies consist of a top nozzle, spacer grid, bottom nozzle, and guide/instrumentation tubes. The top and bottom nozzle are fixed to the guide tubes using a screw or bulge method. The spacer grid fixed to the guide/instrumentation tubes using a spot weld or bulge method. To understand the in-reactor performance of PWR FA, several devices and test techniques have been developed for mechanical property tests. Among the structural components of PWR FA, a spacer grid, a hold down spring of a top nozzle and a connecting part of FA were considered. Experimental works were carried out for the unirradiated and irradiated components of advanced nuclear fuel assemblies for KSNPs and Westinghouse type PWRs at IMEF (Irradiated Materials Examination Facility) at KAERI. The developed techniques were verified through a hot cell tests. (author)

  10. Total evaluation of in bundle void fraction measurement test of PWR fuel assembly

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation is performing the various proof or verification tests on safety and reliability of nuclear power plants under the sponsorship of the Ministry of International Trade and Industry. As one program of these Japanese national projects, an in bundle void fraction measurement test of a pressurized water reactor (PWR) fuel assembly was started in 1987 and finished at the end of 1994. The experiments were performed using the 5 x 5 square array rod bundle test sections. The rod bundle test section simulates the partial section and full length of a 17 x 17 type Japanese PWR fuel assembly. A distribution of subchannel averaged void fraction in a rod bundle test section was measured by the gamma-ray attenuation method using the stationary multi beam systems. The additional single channel test was performed to obtain the required information for the calibration of the rod bundle test data and the assessment of the void prediction method. Three test rod bundles were prepared to analyze an axial power distribution effect, an unheated rod effect, and a grid spacer effect. And, the obtained data were used for the assessment of the void prediction method relevant to the subchannel averaged void fraction of PWR fuel assemblies. This paper describes the outline of the experiments, the evaluation of the experimental data and the assessment of void prediction method

  11. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Energy Technology Data Exchange (ETDEWEB)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

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

    International Nuclear Information System (INIS)

    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)

  13. Secure and effective valve stem sealing in PWR power generating plants

    International Nuclear Information System (INIS)

    The PWR power generating plant combines severe operating conditions with the highest safety requirements, making it one of the most demanding environments for seals. An analysis of the conditions inherent in its operation reveals: an aggressive and radioactive fluid at high temperature and pressure; frequent thermal shocks; and hazards for maintenance personnel in the containment area unless the reactor is shut down. The achievement of today's quality and safety standards owes much to the experience, research and testing carried out by the Electricite de France during its graduation from its first nuclear unit to become the world's most important manager of PWR plants with over 45 now under its control. The number of valves involved in the French nuclear program is in excess of 1,300,000. Knowing what the affect of a leak can be, especially if it necessitates a shutdown of the power station, the need to insure the quality of valve sealing can be appreciated. At the beginning of their nuclear building program, the EdF was finding that valves, representing only 2 percent of the investment in a PWR plant, caused 20% of the unwanted outages and cost 60% of the total of plant maintenance. In this report, the author endeavors to show how this problem was solved by team work and concerted action by the EdF, the valve constructors and seal manufacturer, not forgetting the importance of informing and training the maintenance and repair teams within the power stations themselves

  14. A Novel Burnable Absorber Concept for PWR: BigT (Burnable Absorber-Integrated Guide Thimble)

    Energy Technology Data Exchange (ETDEWEB)

    Yahya, Mohdsyukri; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Chung, Chang Kyu [KEPCO Engineering and Construction Company, Daejeon (Korea, Republic of)

    2014-05-15

    This paper presents the essential BigT design concepts and its lattice neutronic characteristics. Neutronic performance of a newly-proposed BA concept for PWR named BigT is investigated in this study. Preliminary lattice analyses of the BigT absorber-loaded WH 17x17 fuel assembly show a high potential of the concept as it performs relatively well in comparison with commercial burnable absorber technologies, especially in managing reactivity depletion and peaking factor. A sufficiently high control rod worth can still be obtained with the BigT absorbers in place. It is expected that with such performance and design flexibilities, any loading pattern and core management objective, including a soluble boron-free PWR, can potentially be fulfilled with the BigT absorbers. Future study involving full 3D reactor core simulations with the BigT absorbers shall hopefully verify this hypothesis. A new burnable absorber design for Pressurized Water Reactor (PWR) named 'Burnable absorber-Integrated control rod Guide Thimble' (BigT) was recently proposed. Unlike conventional burnable absorber (BA) technologies, the BigT integrates BA materials directly into the guide thimble but still allows insertion of control rod (CR). In addition, the BigT offers a variety of design flexibilities such that any loading pattern and core management objective can potentially be fulfilled.

  15. Application of a burnup verification meter to actinide-only burnup credit for spent PWR fuel

    International Nuclear Information System (INIS)

    A measurement system to verify reactor records for burnup of spent fuel at pressurized water reactors (PWR) has been developed by Sandia National Laboratories and tested at US nuclear utility sites. The system makes use of the Fork detector designed at Los Alamos National Laboratory for the safeguards program of the International Atomic Energy Agency. A single-point measurement of the neutrons and gamma- rays emitted from a PWR assembly is made at the center plane of the assembly while it is partially raised from its rack in the spent fuel pool. The objective of the measurements is to determine the variation in burnup assignments among a group of assemblies, and to identify anomalous assemblies that might adversely affect nuclear criticality safety. The measurements also provide an internal consistency check for reactor records of cooling time and initial enrichment. The burnup verification system has been proposed for qualifying spent fuel assemblies for loading into containers designed using burnup credit techniques. The system is incorporated in the US Department of Energy's.''Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages'' [DOE/RW 19951

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

    International Nuclear Information System (INIS)

    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)

  17. Chinese Geography through Chinese Cuisine

    Science.gov (United States)

    Lipman, Jonathan

    2010-01-01

    China has the world's largest population, now over 1.3 billion, but its land area (much of it high mountains or desert) is about the same as that of the United States, which has less than one-fourth as many people. So Chinese farmers have learned to use every inch of their fertile land intensively. Pressure on the land has required extremely…

  18. Numerical simulation using CFD software of countercurrent gas-liquid flow in a PWR hot leg under reflux condensation

    Energy Technology Data Exchange (ETDEWEB)

    Utanohara, Yoichi, E-mail: utanohara@inss.co.j [Institute of Nuclear Safety System, Inc., 64 Sata, Mihama-cho, Mikata-gun, Fukui 919-1205 (Japan); Kinoshita, Ikuo, E-mail: kinoshita@inss.co.j [Institute of Nuclear Safety System, Inc., 64 Sata, Mihama-cho, Mikata-gun, Fukui 919-1205 (Japan); Murase, Michio, E-mail: murase@inss.co.j [Institute of Nuclear Safety System, Inc., 64 Sata, Mihama-cho, Mikata-gun, Fukui 919-1205 (Japan); Minami, Noritoshi, E-mail: minami.noritoshi@c4.kepco.co.j [Kansai Electric Power Company, Inc., Mihama-cho, Mikata-gun, Fukui 919-1141 (Japan); Nariai, Toshifumi [Graduate School of Engineering, Kobe University, Nada, Kobe 657-8501 (Japan); Tomiyama, Akio, E-mail: tomiyama@people.kobe-u.ac.j [Graduate School of Engineering, Kobe University, Nada, Kobe 657-8501 (Japan)

    2011-05-15

    Research highlights: The wave height measurements indicated the interfacial drag force mainly consisted of form drag. The resolution of the computational cell affected the CCFL characteristics. The interfacial drag correlations employed in this study can be applied to PWR scale simulations. - Abstract: In order to improve the countercurrent flow model of a transient analysis code, countercurrent air-water tests were previously conducted using a 1/15 scale model of the PWR hot leg and numerical simulations of the tests were carried out using the two-fluid model implemented in the CFD software FLUENT 6.3.26. The predicted flow patterns and CCFL characteristics agreed well with the experimental data. However, the validation of the interfacial drag correlation used in the two-fluid model was still insufficient, especially regarding the applicability to actual PWR conditions. In this study, we measured water levels and wave heights in the 1/15 scale setup to understand the characteristics of the interfacial drag, and we considered a relationship between the wave height and the interfacial drag coefficient. Numerical simulations to examine the effects of cell size and interfacial drag correlations on numerical predictions were conducted under PWR plant conditions. Wave heights strongly related with the water level and interfacial drag coefficient, which indicates that the interfacial drag force mainly consists of form drag. The cell size affected the gas velocity at the onset of flooding in the process of increasing gas flow rate. The gas volumetric fluxes at CCFL predicted using fine cells were higher than those using normal cells. On the other hand, the cell size did not have a significant influence on the process of decreasing gas flow rate. The predictions for the PWR condition using a reference set of interfacial drag correlations agreed well with the Upper Plenum Test Facility data of the PWR scale experiment in the region of medium gas volumetric fluxes. The

  19. Chinese Weddings

    Institute of Scientific and Technical Information of China (English)

    1997-01-01

    ACCORDING to the Marriage Law of the People’s Republic of China, marital kinship is established and protected by law when a couple registers at tile local marriage registration office. The newly-weds usually hold a wedding feast in celebration at home or in a restaurant. The big red Chinese character, "Double Happiness," would be pasted on walls at the ceremonial hall to

  20. Secondary water chemistry control practices and results of the Japanese PWR plants

    International Nuclear Information System (INIS)

    In Japan, since the start of the operation of the first PWR plant, Mihama Unit-1 in 1970, 24 PWR plants have been built by 2010, and all of them are in operation. Due to the plant-specific needs of management, and by flexibly incorporating the state-of-the-art insights into the design, the system configurations of the plants vary so many as 15 types. Meanwhile, the geographical feature of Japan makes all the Japanese PWR plants to have condensers cooled by sea water, and all the plants have a common system with a full-flow Condensate Polisher System (CPS). To prevent corrosion, continued improvements of the secondary water chemistry management has been performed like other countries, and one of the major features of the Japanese PWR plants is an enhanced provision for the condenser leakage. The water quality of SG (Steam Generator) has been significantly improved by the provision for the sea water leakage, in combination with other improvements in water chemistry management. Also in Japan, almost all of the treatments of the spent polisher resin and the wastewater are performed within the power plant sites. To facilitate the treatment of the waste water and the regeneration of the spent resins, either ammonia or ETA (Ethanol Amine) is selected as the pH adjustment agent for the secondary system water. Also at the ammonia treatment, high pH accomplishes the inhibition of the piping wall thinning and the lower iron transportation into SGs. In addition, the iron transported into the SG is removed by the chemical conditioning treatment called ASCA (Advanced Scale Conditioning Agent). This provides the effective recovery of the SG heat-transfer performance, and the improved SG support plate BEC (Broached Egg Crate) hole blockage rates. Basically in Japan, the secondary water chemistry management has been improved based on a single basic specification, for the variety of the plant configurations, with the plant-specific investigations and analyses. This paper summarizes

  1. Chinese Culture and Leadership.

    Science.gov (United States)

    Wong, Kam-Cheung

    2001-01-01

    Describes essential characteristics of Chinese philosophical tradition; Discusses Western perspectives on value leadership in education, particularly moral leadership. Discuses moral leadership from a Chinese philosophical perspective, especially Confucianism. Draws implications for using Chinese cultural and philosophical traditions to develop…

  2. Chinese restaurant syndrome

    Science.gov (United States)

    Chinese restaurant syndrome is a set of symptoms that some people have after eating Chinese food. A food additive ... Chinese restaurant syndrome is most often diagnosed based on the symptoms. The health care provider may ask the following ...

  3. CHINESE JOURNAL OF CHEMISTRY

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    @@Chinese Journal of Chemistry is an international journal published in English by the Chinese Chemical Society with its editorial office hosted by Shanghai Institute of Organic Chemistry, Chinese Academy of Sciences.

  4. Absolutely Chinese

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    <正>Ink painting animation - an exquisite art form that ends up in a museum Little Tadpole Looking for Mummy Little Tadpole Looking for Mummy, China’s first ink painting animation, was produced in 1961. With innovation in painting, photography and production technology, it was the first effort at ’animating’ ink paintings. Fishes, shrimps, frogs and crabs in the film resemble those in Qi Baishi’s works. By any shot, it’s an animated painting of fishes and insects, one that is suffused with a taste of Chinese ink-and-wash painting.

  5. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Martinez, R. M.; Fernandez, K.; Morato, D. J.; Bocanegra Melian, R.; Mena, L.; Queral, C.

    2014-07-01

    During the development of a severe accident in a PWR reactor can be generated, large quantities of hydrogen by the oxidation of metals present in the nucleus, mainly the zirconium fuel pods. This hydrogen, along with steam and other gases, can be released to the atmosphere of contention by a leak or break in the primary circuit and achieving conditions in which combustion may occur. Combustion causes thermal and pressure loads that can damage the security systems and the integrity of the containment building, last barrier of confinement of radioactive materials. The main condition that defines the characteristics of the combustion is the concentration of species, detailed knowledge of the distribution of hydrogen is very important to correctly predict the possible damage to the containment in the event that there is combustion. (Author)

  6. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  7. Optimization of thermal efficiency of nuclear central power like as PWR; Otimizacao da eficiencia termica de uma usina nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lapa, Nelbia da Silva

    2005-10-15

    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

  8. Chinese Calendar and Chinese Telegraphic Code.

    Science.gov (United States)

    Defense Language Inst., Monterey, CA.

    This manual contains: (1) Chinese calendars for the hundred years from 1881 to 1980; and (2) the Chinese telegraphic code. Each page in Part One presents the calendar for each year in both Chinese and English. There are 97 charts in Part Two representing the telegraphic code. (AMH)

  9. In service inspection of fatigue cracks in a 1/R th scale PWR vessel with pressure cycles

    International Nuclear Information System (INIS)

    THE JOINT RESEARCH CENTER (JRC) at ISPRA has undertaken for several years a study on the propagation of fatigue cracks in PWR vessels. The goal is to establish a relation between the size and the location of the defects on the one hand, and the residual life of the structure on the other hand/1/. In order to verify the validity of theoretical models which were developed, a mock-up at a 1/5th scale of a PWR vessel was built. It was submitted to pressurization cycles representative of the operating conditions in a PWR. Therefore, it was necessary to have Non-Destructive methods which allows a stable and reliable track on a long period of time. This paper makes a review of results obtained by CEA during 5 successive In-Service Inspections on a 4 years period

  10. Validation of SWAT for burnup credit problems by analysis of post irradiation examination of 17*17 PWR fuel assembly

    International Nuclear Information System (INIS)

    For adopting burnup credit in transport or storage of spent fuel (SF), development of a reliable burnup calculation code is crucial. For this purpose, data of Post Irradiation Examination (PIE) have been extensively analyzed to evaluate accuracy of burnup calculation codes for a 14*14 or 15*15 PWR fuel assembly. This study shows results of analysis of this latest PIE with SWAT and ORIGEN2.1. SWAT is an integrated burnup code system for a 17*17 PWR fuel assembly that has been developed by Tohoku University and JAERI. The results show that SWAT can more precisely predict nuclide composition of latest PWR assembly than ORIGEN2.1. (O.M.)

  11. Valve inlet fluid conditions for pressurizer safety and relief valves in Westinghouse-designed plants. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Meliksetian, A.; Sklencar, A.M.

    1982-12-01

    The overpressure transients for Westinghouse-designed NSSSs are reviewed to determine the fluid conditions at the inlet to the PORV and safety valves. The transients considered are: licensing (FSAR) transients; extended operation of high pressure safety injection system; and cold overpressurization. The results of this review, presented in the form of tables and graphs, define the range of fluid conditions expected at the inlet to pressurized safety and power-operated relief valves utilized in Westinghouse-designed PWR units. These results will provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI/PWR Safety and Relief Valve Test Program indeed envelop those expected in their units.

  12. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  13. PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF

    Directory of Open Access Journals (Sweden)

    Benfarah Moez

    2016-01-01

    Full Text Available Normal operation of PWR generates corrosion and wear products in the primary circuit which are activated in the core and constitute the major source of the radiation field. In addition, cases of fuel failure and alpha emitter dissemination in the coolant system could represent a significant radiological risk. Radiation field and alpha risks are the main constraints to carry out maintenance and to handle effluents. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products and actinides and to carry out the appropriate measurements in PWR circuits and loop experiments. As a matter of fact, it is more than necessary to develop and use a reactor contamination assessment code in order to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has actually been developed and used for this aim. It is presented in this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination. Then, the use of OSCAR is illustrated with two examples from our engineering studies. The first example of OSCAR engineering studies is linked to the behavior of the activated corrosion products. The selected example carefully explores the impact of the restart conditions following a reactor mid-cycle shutdown on circuit contamination. The second example of OSCAR use concerns fission products and disseminated fissile material behavior in the primary coolant. This example is a parametric study of the correlation between the quantity of disseminated fuel and the variation of Iodine 134 in the primary coolant.

  14. Current status of research and development of nuclear fuel elements for PWR in Indonesia

    International Nuclear Information System (INIS)

    Full text: The energy need of Indonesia is increasing due to the population growth and for the economic progress. The government of Indonesia intends to apply an optimum energy mix comprising all viable prospective energy sources. The Government Regulation No. 5 year 2006 indicates the target of energy mix until 2025 and the share of nuclear energy is about 2% of primary energy or 4% of electricity (4000 MWe). The first two units of NPP is expected to be operated before 2020 as stated in Act No. 17 year 2007 on National Long Term Development Planning 2005-2025. The first NPP to be operated in Indonesia is PWR type with capacity of 1000 MWe/unit. One of the strategies to strength and increase national capacity in the program for NPP introduction is domestication industry for nuclear fuel. To reach this purpose, the activities of research and development is focus on nuclear fuel production technology for PWR. Currently research and development activity in Indonesia is to produce prototype of nuclear fuel element for PWR in the form of test fuel pin or mini pin. In this paper we will presenting the pelletization and fabrication technology development. The existing facility was designed for PHWR fuel element of CIRENE type. Development of pelletization technology is carried out by modifying the compacting machine. Parameters of compacting and sintering are determined based on both compressibility and compactibility of the pellet as indicated by density and mechanical strength of the UO2 green pellets. The sintering parameters to be determined are temperature, heating rate, and soaking time. Currently, the fabrication process is under experiment. All of the data resulted from the experiment that will be presented in this meeting. (author)

  15. Performance of PWR study in the technology supplier countries: south korea and japan case

    International Nuclear Information System (INIS)

    Electricity is needed as an infrastructure to support the national economic growth. For economic development sustainability, energy alternatives should be provided. Nuclear Power Plant (NPP) become the alternative of electricity generation for optimum energy mix in Indonesia and planned to operate in the 2016. Several studies have already done to prepare the NPP construction. This study focused on NPP performance especially PWR type in Asia, namely Japan and South Korea. Methodology used in this is literature tracing and a small calculation. The energy availability per unit per year is used as a parameter for evaluating the NPP performance. This conclusion are 1) the amount of NPP - PWR type in Japan is 22 units with total operational experiences 526 reactor-years and the average energy availability factor about 70.7% per unit per year. Meanwhile for the same type South Korea has 16 unit with total operational experience 222 reactor-years and average availability factor per unit per year is about 86.9%. 2) the 1000 class of PWR type both South Korea and Japan have 14 units. The operational experiences for thi class is 170 reactor-year for South Korean and 307 reactor-year for Japan. Meanwhile the average availability factor per unit per year is about 87.0% for South Korea and 69.6% for Japan. 3) the average availability factor is closed to capacity factor, so is important for real figure in assuming the techno-economic parameters, because it will influence the result o economic calculation. (author)

  16. PWR physics, operation and safety - Management of accidental situations of the reactor system

    International Nuclear Information System (INIS)

    This document contains a brief presentation and the table of contents of a book in which the author first presents the main types of accidents which are taken into account in safety demonstration. He presents the risk concerning the three safety barriers, and the various accidents affecting the three safety functions: reactivity control, power evacuation, confinement by the third barrier. Then the author describes approaches to the management of accidents affecting these three safety functions: reactivity insertion accidents due to absorber withdrawal (presentation, absorber cluster extraction transients, primary fluid dilution transient), steam pipe failure accidents or reactivity insertion by primary cooling (presentation, description of a transient of steam-pipe failure, sensitivity study of main parameters), loss-of-coolant accidents (presentation, intermediate breach, the big breach, peculiar case of breaches in stopped status), total loss of support systems such as in Fukushima (loss of electric supplies, of the cold source), loss of steam generator tubes. In the next part, the author addresses the Three Mile Island (TMI) accident and the lessons learned in terms of post-accidental management: presentation of the reactor and description of the accident. The author presents the 'status approach' of the post-accidental management, addresses the core post-fusion situations and their consequences as far as containment is concerned. He finally proposes ways to manage accidental situations for the PWR system. Appendices propose some additional aspects of system thermal-hydraulics, a presentation of safety deterministic and probabilistic approaches, comments on the Chernobyl and Fukushima accidents, comments on human and organizational factors regarding nuclear safety, some specific design aspects of the PWR reactor regarding safety, a presentation of assessment equations and data for the 1300 MWe PWR model

  17. Criticality analysis of PWR spent fuel storage facilities inside nuclear power plants

    International Nuclear Information System (INIS)

    This paper describes some of the main features of the actinide plus fission product burnup credit methodology used by Siemens for criticality safety design analysis of wet PWR storage pools with soluble boron in the pool water. Application of burnup credit requires knowledge of the isotopic inventory of the irradiated fuel for which burnup credit is taken. This knowledge is gained by using depletion codes. The results of the depletion analysis are a necessary input to the criticality analysis. Siemens performs depletion calculations for PWR fuel burnup credit applications with the aid of the Siemens standard design procedure SAV90. The quality of this procedure relies on statistics on the differences between calculation and measurement extracted from in-core measurement data and chemical assay data. Siemens performs criticality safety calculations with the aid of the criticality calculation modules of the SCALE code package. These modules are verified many times with the aid of various kinds of critical experiments and configurations: Application of these modules to spent LWR fuel assembly storage pools was verified by analyzing critical experiments simulating such storage pools. Actinide plus fission product burnup credit applications of these modules were verified by analyzing PWR reactor critical configurations. The result of performing a burnup credit analysis is the determination of a burnup, credit loading curve for the spent fuel storage racks designed for burnup credit. This curve specifies the loading criterion by indicating the minimum burnup necessary for the fuel assembly with a specific initial enrichment to be placed in the storage racks designed for burnup credit. The loading of the spent fuel storage racks designed for burnup credit requires the implementation of controls to ensure that the loading curve is met. The controls include the determination of fuel assembly burnup based on reactor records. (author)

  18. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    International Nuclear Information System (INIS)

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO2 fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  19. Cold leg condensation model for analyzing loss-of-coolant accident in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Liao, Jun, E-mail: liaoj@westinghouse.com; Frepoli, Cesare; Ohkawa, Katsuhiro

    2015-04-15

    Highlights: • Direct contact cold leg condensation model for full spectrum LOCA evaluation model. • The cold leg condensation model addresses both large break LOCA and small break LOCA. • The model is assessed against both large break and small break LOCA experiments. • Scalability of the cold leg condensation model to full scale PWR is discussed. - Abstract: Direct contact condensation in the cold leg of pressurized water reactor is an important phenomenon during a postulated loss-of-coolant accident. The amount of condensation in the cold legs impacts the thermal hydraulic behavior of the reactor coolant system and eventually the integration of reactor nuclear core. A cold leg condensation model was developed for the WCOBRA/TRAC-TF2 safety analysis code. The model correlated the COSI test data and addressed the scaling issues with respect to geometry, pressure, and steam and water flow rates expected during a typical PWR LOCA. The correlation was found to be in good agreement with separate effects and integral effects experimental data and implemented in the WCOBRA/TRAC-TF2 safety analysis code. The cold leg condensation model was assessed against various small break and large break LOCA separate effects tests such as COSI experiments, ROSA experiments and UPTF experiments. Those experiments cover a wide range of cold leg dimensions, system pressures, mass flow rates, and fluid properties. All the predicted condensation results match reasonably well with the experimental data. Scalability discussions on the diameter, flow area, length, superficial velocity, Reynolds number of both cold leg and SI line, and Froude number of SI line in the Westinghouse COSI test facility were provided. The distortion of the SI jet Reynolds number is moderate. The scaling analysis together with the validation matrix covering a wide range of cold leg diameter, SI flow rate and SI Reynolds number support the scalability of the developed cold leg condensation model to the full

  20. In- and ex-vessel coupled analysis of IVR-ERVC phenomenon for large scale PWR

    International Nuclear Information System (INIS)

    Highlights: • MELCOR models are built for large scale PWR with thermal power reaching 5000 MWt. • In- and ex-vessel coupled transient analysis of IVR-ERVC phenomenon is performed. • Results show that the IVR-ERVC strategy is an effective way to maintain RPV integrity during a severe accident. - Abstract: As a key severe accident management strategy for light water reactors (LWRs), in-vessel retention (IVR) through external reactor vessel cooling (ERVC) has been the focus of relevant studies for decades. However, previous studies only investigated the molten pool configurations considered to be in a final steady state mainly for reactors of such as AP600 and AP1000. Furthermore, most of studies performed in the past dealt with analysis for an isolated IVR-ERVC process, without considering the strong coupling between the internal and external reactor pressure vessel (RPV) conditions. This paper addresses the IVR-ERVC issues from a transient perspective using the severe accident code MELCOR for a large advanced passive power plant: a three-loop, 5000 MWt scale pressurized water reactor with passive safety features. The analysis is mainly focused on the severe accident transients including core degradation and relocation, molten pool formation and growth, and heat transfer within a molten pool. Furthermore, internal and external RPV conditions are combined together in the IVR-ERVC analysis. MELCOR calculations for lower head heat flux are then compared with critical heat flux (CHF) to assess the effectiveness of IVR-ERVC. The results suggest that lower head heat flux is below the CHF value. Therefore, the IVR-ERVC strategy for this large PWR is considered to be feasible. It was also found that as the reactor power is raised to large scale PWR, new accident sequences may occur during the severe accident evolution, thus leading to a proposal of a completely new molten pool configuration for future studies

  1. PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Blakeman, Edward D [ORNL; Peplow, Douglas E. [ORNL; Wagner, John C [ORNL; Murphy, Brian D [ORNL; Mueller, Don [ORNL

    2007-09-01

    The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.

  2. Chinese letterkunde. Een inleiding

    NARCIS (Netherlands)

    Idema, Wilt; Haft, Lloyd

    2005-01-01

    De Chinese cultuur mag zich verheugen in een groeiende belangstelling. Chinese films bereiken in Nederland een steeds omvangrijker publiek en ook de moderne Chinese literatuur, die sinds de jaren tachtig een grote bloei doormaakt, wordt door veel liefhebbers op de voet gevolgd. Chinese Letterkunde b

  3. The continued development of the MFM suite and its practical application on a PWR system

    DEFF Research Database (Denmark)

    Thunem, Harald P-J; Zhang, Xinxin

    2015-01-01

    This paper reports on the results from the practical application of the Shape Shifter framework on the continued development of a graphical editing suite, the MFM Suite, for MFM and process model design and analysis. The primary use of the MFM Suite is diagnosis and prognosis of anomalies in phys...... physical processes. One of the Halden Reactor Project’s advanced NPP simulators based on a PWR is used to demonstrate the applicability of the suite in realistic situations. The paper presents a summary and suggests some plans for future research and development....

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

    International Nuclear Information System (INIS)

    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.)

  5. The three-dimensional PWR transient code ANTI; rod ejection test calculation

    International Nuclear Information System (INIS)

    ANTI is a computer program being developed for three-dimensional coupled neutronics and thermal-hydraulics description of a PWR core under transient conditions. In this report a test example calculated by the program is described. The test example is a simulation of a control rod ejection from a very small reactor core (to save somputing time). In order to show the influence of cross flow between adjacent fuel elements the same calculation was performed both with the cross flow option and with closed hydraulic channels. (author)

  6. Simulation of a low-pressure severe accident scenario in a PWR with ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, Mathias; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2013-07-01

    The plant behavior of a Pressurized Water Reactor (PWR) during a severe accident scenario is analyzed with system code ATHLET-CD Mod. 2.2C in order to assess the code capabilities in terms of the late-phase of the core degradation. For this purpose a severe accident sequence caused by a Station Black-out and a large break in the primary cooling system is simulated both without any accident management measures and with a delayed reflooding of the substantially degraded core. Selected code results are presented in this paper. (orig.)

  7. Application of LBB to high energy piping systems in operating PWR

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  8. Simulation model for the dynamic behavior of the hydraUlic circuito of PWR reactors

    International Nuclear Information System (INIS)

    The present work consist of the development of a computer code for the simulations of hydraulic transients caused by stoppages of the primary coolant pumps of nuclear reactors and it applied to the hydraulic circuits typical of PWR reactor. The code calculates the time-histories of the mass flux, rotation speed, electric and hydraulic torque and dynamic head of the pumps. It can be used for any combination of active and inactive pumps. Several transients were analysed and the results were compared with comparared with data from the Angra-I nuclear power plant. The results were considered satisfactory. (author)

  9. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  10. Environment-insensitive equivalent diffusion theory group constants for PWR [pressurized water reactor] radial reflector regions

    International Nuclear Information System (INIS)

    A novel approach combining the nodal equivalence theory (NET) and the response matrix homogenization methods into a single procedure, has been developed for generating equivalent nodal parameters for the radial reflector of a PWR [pressurized water reactor]. This procedure yields equivalent parameters which are per definition less sensitive to local environment than those obtained via a straightforward NET homogenization method. Unlike other response matrix approaches which may achieve this same goal, this method requires only a single multi-group spectral calculation to obtain all the necessary information for generating the equivalent parameters. Numerical results are presented which demonstrate the potential and the superiority of this reflector model over other modern models

  11. A systematic approach for development of a PWR cladding corrosion model

    International Nuclear Information System (INIS)

    A new model for the in-reactor corrosion of Improved (low-tin) Zircaloy-4 cladding irradiated in commercial pressurized water reactors (PWRs) is described. The model is based on an extensive database of PWR fuel cladding corrosion data from fuel irradiated in commercial reactors, with a range of fuel duty and coolant chemistry control strategies which bracket current PWR fuel management practices. The fuel thermal duty with these current fuel management practices is characterized by a significant amount of sub-cooled nucleate boiling (SNB) during the fuel's residence in-core, and the cladding corrosion model is very sensitive to the coolant heat transfer models used to calculate the coolant temperature at the oxide surface. The systematic approach to developing the new corrosion model therefore began with a review and evaluation of several alternative models for the forced convection and SNB coolant heat transfer. The heat transfer literature is not sufficient to determine which of these heat transfer models is most appropriate for PWR fuel rod operating conditions, and the selection of the coolant heat transfer model used in the new cladding corrosion model has been coupled with a statistical analysis of the in-reactor corrosion enhancement factors and their impact on obtaining the best fit to the cladding corrosion data. The in-reactor corrosion enhancement factors considered in this statistical analysis are based on a review of the current literature for PWR cladding corrosion phenomenology and models. Fuel operating condition factors which this literature review indicated could have a significant effect on the cladding corrosion performance were also evaluated in detail in developing the corrosion model. An iterative least squares fitting procedure was used to obtain the model coefficients and select the coolant heat transfer models and in-reactor corrosion enhancement factors. This statistical procedure was completed with an exhaustive analysis of the model

  12. The continued development of the MFM suite and its practical application on a PWR system

    DEFF Research Database (Denmark)

    Thunem, Harald P-J; Zhang, Xinxin

    2015-01-01

    This paper reports on the results from the practical application of the Shape Shifter framework on the continued development of a graphical editing suite, the MFM Suite, for MFM and process model design and analysis. The primary use of the MFM Suite is diagnosis and prognosis of anomalies...... in physical processes. One of the Halden Reactor Project’s advanced NPP simulators based on a PWR is used to demonstrate the applicability of the suite in realistic situations. The paper presents a summary and suggests some plans for future research and development....

  13. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  14. Application of fine absolute rated filters to PWR CVCS and radwaste filter duties

    International Nuclear Information System (INIS)

    Filters for PWR systems have generally been specified in terms of nominal removal ratings. Products supplied by different manufacturers against these specifications vary considerably in removal efficiency and service life. The paper describes the method used by Pall Corporation for reliable, reproducible assessment of filter performance and determination of absolute filtration ratings. The old nominal specifications are shown to be too coarse for effective removal of particle borne activity. The introduction into service of new absolute rated fine filter media with positive zeta potential is described. (author)

  15. LOBI-BL-40: an approach to the SGTR phenomenology in a one loop PWR

    International Nuclear Information System (INIS)

    The aim of the ''LOBI-BL-40'' project is to study the phenomena and the recovery techniques involved in Steam Generator Tube Rupture (SGTR) events in a commercial single-loop PWR plant. Experimental support for the project will be provided by the BL-40 experiment, to be carried out at the LOBI installation (JRC, Ispra-Varese, Italy). This paper describes the project, its objectives, the activities involved and the progress made, with special attention to specification of the BL-40 experiments. The work is being undertaken by the Spanish utility UNION-FENOSA. (author)

  16. Application of burnup credit with partial boron credit to PWR spent fuel storage pools

    International Nuclear Information System (INIS)

    The outcome of performing a burnup credit criticality safety analysis of a PWR spent fuel storage pool is the determination of burnup credit loading curves BLC=BLC(e) for the spent fuel storage racks designed for burnup credit, cp. Reference. A burnup credit loading curve BLC=BLC(e) specifies the loading criterion by indicating the minimum burnup BLC(e) necessary for the fuel assembly with a specific initial enrichment e to be placed in storage racks designed for burnup credit. (orig.)

  17. Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT

    Directory of Open Access Journals (Sweden)

    Uwe Imke

    2012-01-01

    Full Text Available SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT. The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy.

  18. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. 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) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  19. Behaviour of a PWR safety containment with rising internal pressure load

    International Nuclear Information System (INIS)

    For the PWR Philippsburg (KKP II) a slowly increasing containment pressure typical for a core melt accident is assumed. The failure pressure and failure mode of the containment are determined using the ROTMEM code. It turns out that the sealing box covering the bolted connection at the equipment hatch will fail at a containment overpressure between 12.9 and 13.7 bar. Then the leakage through the bolted connection is sufficient to prevent further pressure increase. However, if the sealing box failed at a somewhat higher pressure, a global containment failure with extreme mechanical damage would have to be expected. (orig.)

  20. Logic flowgraph model for disturbance analysis of a PWR pressurizer system

    Energy Technology Data Exchange (ETDEWEB)

    Guarro, S.; Okrent, D.

    1984-01-01

    The Logic Flowgraph Methodology (LFM) has been developed as a synthetic simulation language for process reliability or disturbance analysis applications. A Disturbance Analysis System (DAS) using the LFM models can store the necessary information concerning a given process in an efficient way, and automatically construct in real time the diagnostic tree(s) showing the root cause(s) of occurring disturbances. A comprehensive LFM model for a PWR pressurizer system is presented and discussed, and the latest version of the LFM tree synthesis routine, optimized to achieve reduction of computer memory usage, is used to show the LFM diagnoses of selected hypothetic disturbances.

  1. 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).

  2. Validation of seismic soil structure interaction (SSI) methodology for a UK PWR nuclear power station

    International Nuclear Information System (INIS)

    The seismic loading information for use in the seismic design of equipment and minor structures within a nuclear power plant is determined from a dynamic response analysis of the building in which they are located. This dynamic response analysis needs to capture the global response of both the building structure and adjacent soil and is commonly referred to as a soil structure interaction (SSI) analysis. NNC have developed a simple and cost effective methodology for the seismic SSI analysis of buildings in a PWR nuclear power station at a UK soft site. This paper outlines the NNC methodology and describes the approach adopted for its validation

  3. The effects of fission gas release on PWR fuel rod design and performance

    International Nuclear Information System (INIS)

    The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50 000 to 60 000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

  4. The effects of fission gas release on PWR fuel rod design and performance

    International Nuclear Information System (INIS)

    The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50,000 to 60,000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

  5. The analysis of dose reduction practices in the PWR units of Electricite de France

    International Nuclear Information System (INIS)

    Electricite de France is currently operating twenty-one 900 MWe standard PWR units. Along with the increasing size of the french nuclear program, experience has been gathered on the problem of occupational exposures and actions have been undertaken in order to keep doses of workers as low as reasonably achievable. This paper first presents the main results concerning collective exposures at units already in operation. Then a set of protective actions already taken have been assessed, taking into account their costs and effectiveness and allowing for the possible interdependence of protection and energy production objectives. The limits of such an approach are discussed in conclusion. (author)

  6. Evolution of reactor monitoring and protection systems for PWR; Evolution des systemes de surveillance et de protection des REP

    Energy Technology Data Exchange (ETDEWEB)

    Chaloin, B. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Mourlevat, J.L. [FRAMATOME ANP, 92 - Paris-La-Defence (France)

    2004-07-01

    This paper presents the evolution of the reactor protection systems and of the reactor monitoring systems for PWR since the initial design in the Fessenheim plant to the latest development for the EPR (European pressurized reactor). The features of both systems for the different kinds of PWR operating in France: 900 MWe, 1300 MWe and N4, are reviewed. The expected development of powerful micro-processors for computation, for data analysis and data storage will make possible in a near future the monitoring on a 3-dimensional basis and on a continuous manner, of the nuclear power released in the core. (A.C.)

  7. Assessment of management alternatives for LWR wastes. Volume 3. Description of German scenarios for PWR and BWR wastes

    International Nuclear Information System (INIS)

    This report deals with the description of a management route for PWR waste relying to a certain extent on German practices in this particular area. This description, which aims at providing input data for subsequent cost evaluation, includes all management steps which are usually implemented for solid, liquid and gaseous wastes from their production up to the interim storage of the final waste products. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for PWR and BWR wastes based on economical and radiological criteria

  8. Management of Chinese restaurant

    OpenAIRE

    Cui , Longbo

    2009-01-01

    With Chinese economy developing rapidly, the Chinese restaurant is under the spotlight, but the management of Chinese restaurant is weak at the moment, especially on the service management, which is an important part of service management in the Chinese restaurant. On the other hand, the managers of Chinese restaurant should pay more attention on the service management for instance brand, service innovation. Service management is core and essential concept for every service company recently, ...

  9. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  10. Estimate of the speed of the refrigerant on a PWR: three way based on the analysis of noise; Estimacion de la volecidad del refrigerante en un PWR: tres vias basadas en el analisis de ruido

    Energy Technology Data Exchange (ETDEWEB)

    Montalvo, C.; Ruiz, M.; Garcia Berrocal, A.

    2014-07-01

    The speed of the refrigerant is a key parameter in the monitoring of the operation a PWR. He know this value and be able to track on-site It allows an understanding of the State of the kernel with valuable information about the refrigerant, and thus behavior on heat exchange which takes place in the reactor. (Author)

  11. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)

    2015-08-15

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

  12. PWR-FBR with closed fuel cycle for a sustainable nuclear energy supply in China

    Institute of Scientific and Technical Information of China (English)

    XU Mi

    2007-01-01

    From the thermal reactor to the fast reactor and then to the fusion reactor; this is the three-step strategy that has been decided for a sustainable nuclear energy supply in China. As the main thermal reactor type, the commercialized development phase of the pressurized water reactor (PWR) has been stepped up. The development of the fast reactor (FBR) is still in the early stage, marked by China experimental fast reactor (CEFR), which is currently under construction. According to the strategy study on the fast reactor development in China, its engineering development will be divided into three steps: the CEFR with a power of 65 MWt 20 Mwe; the China prototype fast reactor (CPFR) with a power of 1 500 MWt/600 Mwe; and the China demonstration fast reactor (CDFR) with a power of 2 500-3 750 MWt 1 000-1 500 Mwe. With regards to the fuel cycle, a 100 ta PWR spent fuel reprocessing pilot plant and a 500 kg/a MOX fabrication plant are under construction. A project involving the construction of an industrial reprocessing plant and an MOX fabrication plant are also under application phase.

  13. Computer code validation study of PWR core design system, CASMO-3/MASTER-{alpha}

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Kim, M. H. [Kyounghee Univ., Taejon (Korea, Republic of); Woo, S. W. [KINS, Taejon (Korea, Republic of)

    1999-05-01

    In this paper, the feasibility of CASMO-3/MASTER-{alpha} nuclear design system was investigated for commercial PWR core. Validation calculation was performed as follows. Firstly, the accuracy of cross section generation from table set using linear feedback model was estimated. Secondly, the results of CASMO-3/MASTER-{alpha} was compared with CASMO-3/NESTLE 5.02 for a few benchmark problems. Microscopic cross sections computed from table set were almost the same with those from CASMO-3. There were small differences between calculated results of two code systems. Thirdly, the repetition of CASMO-3/MASTER-{alpha} calculation for Younggwang Unit-3, Cycle-1 core was done and their results were compared with nuclear design report(NDR) and uncertainty analysis results of KAERI. It was found that uncertainty analysis results were reliable enough because results were agreed each other. It was concluded that the use of nuclear design system CASMO-3/MASTER-{alpha} was validated for commercial PWR core.

  14. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  15. Phenomenology and Course of Severe Accidents in PWR-Plants - Training by Teaching and Demonstration

    International Nuclear Information System (INIS)

    A special one day training course on 'Phenomenology and Course of Severe Accidents in PWR-Plants' was developed at GRS initiated by the interest of German utilities. The work was done in the frame of projects sponsored by the German Ministries for Environment, Nature Conservation and Nuclear Safety (BMU) and for Education, Science, Research and Technology (BMBF). In the paper the intention and the subject of this training course will be discussed and selected parts of the training course will be presented. Demonstrations are made within this training course with the GRS simulator system ATLAS to achieve a broader understanding of the phenomena discussed and the propagation of severe accidents on a plant specific basis. The GRS simulator system ATLAS is linked in this case to the integral code MELCOR and pre-calculated plant specific severe accident calculations are used for the demonstration together with special graphics showing plant specific details. Several training courses have been held since the first one in November, 1996 especially to operators, shift personal and the management board of a German PWR. In the meantime the training course was updated and suggestions for improvements from the participants were included. In the future this type of the training course will be made available for members of crisis teams, instructors of commercial training centres and researchers of different institutions too. (authors)

  16. The role of phosphorus in the irradiation embrittlement of PWR pressure vessel steels

    International Nuclear Information System (INIS)

    An analysis has been performed of the influence of phosphorus on post-irradiation materials properties and microstructures determined on a variety of PWR steels and variants following exposure to MTR or reactor surveillance irradiations to doses not exceeding 7 x 1019n.cm-2 (E>1.0MeV) at 250-2900C. The irradiation-induced shifts in impact transition temperature, matrix hardening and the relative small angle neutron scattering response were found to rise most rapidly with increasing phosphorus when the copper content of the steel was 0.03 w/o. The sensitivity of the changes in mechanical properties to phosphorus content decreased as the copper content was increased. At copper levels typical of modern PWR steel manufacture (Cu3P) produced by the irradiation induced segregation of phosphorus to defect sinks and the depletion of phosphorus in solid solution as detected by high sensitivity electron microscopy and other analytical techniques. At higher levels of copper (approx. 0.3 w/o) the effect of phosphorus on properties was reduced by a factor of three due to the observed incorporation of phosphorus into the small copper precipitates formed during irradiation. Grain boundary embrittlement by phosphorus under irradiation is not thought to be important but further evidence concerning the post-irradiation fracture mode and the development of the deleterious influence of phosphorus with irradiation dose is required for a comprehensive understanding of its action. Some suggestions for future work are made. (author)

  17. Optimal burnable poison-loading in a PWR with carbon coated particle fuel

    International Nuclear Information System (INIS)

    An innovative PWR concept that uses carbon-coated particle fuels moderated by graphite as that of HTGR but cooled by pressurized light water has been studied. The aim of this concept is to take both the best advantages of fuel integrity against fission products release and the reliability PWR technology based on the long operational experience. The purpose of the study is to optimize loading pattern of burnable poison in the proposed core in order to suppress excess reactivity during a cycle. Although there are many parameters to be determined for optimization of the usage of burnable poison, the emphasis is put here on loading patterns of Gadolinia in an assembly and in the core. We investigated the burnup characteristics of the core varying the concentration of burnable poison in a fuel rod, the number of burnable poison-rods in an assembly, and the number of burnable poison-assemblies in the core. The result suggested that Gadolinia was more suitable for this reactor than boron as burnable poison, and it was possible to make the reactivity swing negligible by combining at least three kinds of burnable poison-assemblies in which the amount of Gadolinia was different. Therefore the requirement for the number of control rods was reduced and it meant that Control Rod Programming would become easier. (author)

  18. Validation of STAR-CCM+ for bouyancy driven mixing in a PWR reactor pressure vessel

    International Nuclear Information System (INIS)

    Within the OECD/NEA PKL-II project, experiments have been carried out aimed at investigating the flow mixing in the downcomer and lower plenum of a pressurized water reactor (PWR) in the buoyancy driven mixing regimes. The experiments have been performed at the ROCOM test facility, a 1:5 scaled representation of a KONVOI type pressurized water reactor (PWR). The facility is equipped with advanced instrumentation (i.e. wiremesh sensors) allowing a detailed measurement of flow mixing in the downcomer annulus and at the core inlet. A computational fluid dynamic (CFD) model has been developed at the Paul Scherrer Institute within the STARS project, employing the STAR-CCM+ code. The CFD model has been validated against the ROCOM experimental results. It has been shown that the developed model provided a good agreement with experiment. In order to evaluate the difference between momentum driven and density driven mixing regimes, calculations were performed assuming no density difference, and with 12% higher density in one of the loops respectively. (author)

  19. ALIBABA, an assistance system for the detection of confinement leaks in a PWR reactor

    International Nuclear Information System (INIS)

    The objective of the Crisis Technical Center (CTC) of the French Institute for Nuclear Protection and Safety (IPSN) is to estimates the consequences of a given nuclear accident on the populations and the environment. ALIBABA is a data processing tool available at the CTC and devoted to the detection of confinement leaks in 900 MWe PWR reactors using the activity values measured by the captors of the installation. The heart of this expert system is a structural and functional representation of the different components directly involved in the leak detection (isolating valves, ventilation systems, electric boards etc..). This tool can manage the availability of each component to make qualitative and quantitative balance-sheets. This paper presents the ALIBABA software, an industrial prototype realized with the SPIRAL knowledge base systems generator at the CEA Reactor Studies and Applied Mathematics Service (SERMA) and commercialized by CRIL-Ingenierie Society. It describes the techniques used for the modeling of PWR systems and for the visualization of the survey. The functionality of the man-machine interface is discussed and the means used for the validation of the software are summarized. (J.S.). 6 refs

  20. Design features of Tomari unit 3. Latest 900MW-class PWR design in Japan

    International Nuclear Information System (INIS)

    Tomari Unit 3 is the latest 912 MW PWR in Japan. Its major design features are (1) improvement of thermal efficiency, (2) application of the advanced main control board, (3) system design according to PSA, (4) fatigue analysis in consideration of environmental effect and high-cycle thermal fatigue and (5) concurrent engineering and 3D-CAD. The improvement of thermal efficiency is achieved by increasing generating capacity. It is the result of increase of reactor coolant temperature, improvement of steam turbine efficiency and higher vacuum of condenser. The improvement of steam turbine efficiency is the result of 3D hydraulic design and 54-inch last blade of LP turbine. 54-inch last blade is the first application in Japan. The advanced main control board is the first application to PWR in Japan and reduces operator's mental workload and human error probability. System design according to PSA, fatigue analysis in consideration of environmental effect and high-cycle thermal fatigue and 3D-CAD enhance safety, reliability and integrity of Tomari Unit 3. (author)

  1. Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-06-01

    Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations.

  2. Nuclear criticality safety analysis for the traveller PWR fuel shipping package

    Energy Technology Data Exchange (ETDEWEB)

    Vescovi, P.J.; Kent, N.A.; Casado, C.A. [Westinghouse Electric Co., LLC, Columbia, SC (United States)]|[ENUSA Industrias Avanzadas SA, Madrid (Spain)

    2004-07-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse.

  3. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

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

    International Nuclear Information System (INIS)

    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)

  5. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir, E-mail: luciano.ondir@gmail.com; Andrade, Delvonei Alves de, E-mail: delvonei@ipen.br

    2014-12-15

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs.

  6. Nuclear criticality safety analysis for the traveller PWR fuel shipping package

    International Nuclear Information System (INIS)

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse

  7. Evaluation of new pH control agents for PWR secondary water systems. Final report

    International Nuclear Information System (INIS)

    This report presents an account of work evaluating new organic pH control agents for PWR station secondary systems. A discussion of corrosion control with pH controlling amines is given together with some information on current utility practice with volatile amines. The report describes opportunities for improved corrosion control with new pH controlling amines and the selection of candidates for laboratory evaluation. Progress on EPRI program RP 1571-7, ''Spectroscopic Characterization Methods for New pH Control Agents'', is presented. Also presented is an outline of the second phase of the RP 1571-7 program which was directed toward ''Evaluation of pH Control Agents in Simulated Steam Generator Tubesheet Crevice Environments.'' Quinuclidine and diaminopropane were evaluated using laser Raman spectroscopy. Hydrothermal stability and pH properties at temperature were determined. Quinuclidine was superior to diaminopropane and presents advantages over ammonia, hydrazine and morpholine for plant use as a pH control agent. It was further evaluated in tests with a scaled PWR crevice simulator for corrosion compatibility as well as pH control. 17 refs., 22 figs., 9 tabs

  8. Development and preliminary verification of the PWR on-line core monitoring software system. SOPHORA

    International Nuclear Information System (INIS)

    This paper presents an introduction to the development and preliminary verification of a new on-line core monitoring software system (CMSS), named SOPHORA, for fixed in-core detector (FID) system of PWR. Developed at China General Nuclear Power Corporation (CGN), SOPHORA integrates CGN’s advanced PWR core simulator COCO and thermal-hydraulic sub-channel code LINDEN to manage the real-time core calculation and analysis. Currents measured by the FID are re-evaluated and used as bases to reconstruct the 3-D core power distribution. The key parameters such as peak local power margin and minimum DNBR margin are obtained by comparing with operation limits. Pseudo FID signals generated by data from movable in-core detector (MID) are used to verify the SOPHORA system. Comparison between predicted power peak and the responding MID in-core flux map results shows that the SOPHORA results are reasonable and satisfying. Further verification and validation of SOPHORA is undergoing and will be reported later. (author)

  9. An extension of the validation of SCALE (SAS2H) isotopic predictions for PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Hermann, O.W.

    1996-09-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms. Unlike fresh fuel assumptions typically used in criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict spent fuel composition; these isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the depletion codes and data, experiment is compared with predictions; such comparisons have been done in earlier ORNL work. This report describes additional independent measurements and corresponding calculations as a supplement. The current work includes measured isotopic data from 19 spent fuel samples from the Italian Trino Vercelles PWR and the US Turkey Point-3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results on combination of measured-to-calculated ratios are presented. The results described herein represent an extension to a new reactor design and spent fuel samples with enrichment as high as 3.9 wt% {sup 235}U. Consistency with the earlier work for each of two different cross-section libraries suggests that the estimated biases for each of the isotopes in the earlier work are reasonably good estimates.

  10. Numerical simulation of countercurrent gas-liquid flow in a PWR hot leg using VOF method

    International Nuclear Information System (INIS)

    In order to evaluate flow patterns and CCFL (countercurrent flow limitation) characteristics in a PWR hot leg under reflux condensation, numerical simulations have been done using a two-fluid model and a VOF (volume of fluid) method implemented in the CFD software, FLUENT6.3.26. The two-fluid model gave good agreement with CCFL data under low pressure conditions but did not give good results under high pressure steam-water conditions. On the other hand, the VOF method gave good agreement with CCFL data for tests with a rectangular channel but did not give good results for calculations in a circular channel. Therefore, in this paper, the computational grid and schemes were changed in the VOF method, numerical simulations were done for steam-water flows at 1.5 MPa under PWR full-scale conditions with the inner diameter of 0.75 m, and the calculated results were compared with the UPTF data at 1.5 MPa. As a result, the calculated flow pattern was found to be similar to the flow pattern observed in small-scale air-water tests, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa except in the region of a large steam volumetric flux. (author)

  11. Power-cooling mismatch test series. Test PCM-2A; test results report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cawood, G.W.; Holman, G.W.; Martinson, Z.R.; Legrand, B.L.

    1976-09-01

    The report describes the results of an in-pile experimental investigation of pre- and postcritical heat flux (CHF) behavior of a single 36-inch-long, pressurized water reactor (PWR) type, UO/sub 2/-fueled, zircaloy-clad fuel rod. The nominal coolant conditions for pressure and temperature were representative of those found in a commercial PWR. Nine separate departures from nucleate boiling (DNB) cycles were performed by either of two different methods: (a) decreasing the coolant flow rate while the fuel rod power was held constant, or (b) increasing the fuel rod power while the coolant flow rate was held constant. DNB was obtained during eight of the nine cycles performed. For the final flow reduction, the mass flux was decreased to 6.1 x 10/sup 5/ lb/hr-ft/sup 2/ at a constant peak linear heat generation rate of 17.8 kW/ft. The fuel rod was allowed to remain in film boiling for about 210 seconds during this final flow reduction. The fuel rod remained intact during the test. Results of on-line measurements of the fuel rod behavior are presented together with discussion of instrument performance. A comparison of the data with Fuel Rod Analysis Program-Transient 2 (FRAP-T2) computer code calculations is included.

  12. Addressing the Axial Burnup Distribution in PWR Burnup Credit Criticality Safety

    International Nuclear Information System (INIS)

    This paper summarizes efforts related to developing a technically justifiable approach for addressing the axial burnup distribution in PWR burnup-credit criticality safety analyses. The paper reviews available data on the axial variation in burnup and the effect of axial burnup profiles on reactivity in a SNF cask. A publicly available database of profiles is examined to identify profiles that maximize the neutron multiplication factor, keff, assess its adequacy for general PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. For this assessment, a statistical evaluation of the keff values associated with the profiles in the axial burnup profile database was performed that identifies the most reactive profiles as statistical outliers that are not representative of typical discharged SNF assemblies. The impact of these bounding profiles on the neutron multiplication factor for a high-density burnup credit cask is quantified. Finally, analyses are presented to quantify the potential reactivity consequence of assemblies with axial profiles that are not bounded by the existing database. The paper concludes with findings for addressing the axial burnup distribution in burnup credit analyses

  13. Survey of prediction capabilities of three nuclear data libraries for a PWR application

    International Nuclear Information System (INIS)

    Highlights: • State of uncertainty quantification of three NDLs for a PWR application. • Identification of several order-of-magnitude differences in important contributors. • Impact of cross-material correlations. - Abstract: We survey prediction capabilities of ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0u nuclear data libraries (NDLs) for the application of generating two-group homogenized assembly constants for a steady state diffusion model in the context of UAM-LWR (Uncertainty Analysis in Best-Estimate Modeling for Design, Operation and Safety Analysis of LWRs) Benchmark. We consider two different fuel assembly test cases representing a PWR. State of uncertainty quantification in each NDLs is presented for the application. We expect small differences between the NDLs due to the use of expert judgment in the evaluation processes, and identify several order-of-magnitude differences between the NDLs for significant contributors to uncertainty. We also quantify the contribution from cross-material correlations to the uncertainties

  14. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    International Nuclear Information System (INIS)

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety

  15. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  16. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, Saclay (France); Mimouni, S., E-mail: stephane.mimouni@edf.fr [Electricité de France, EDF MF2E, Chatou (France); Manzini, G., E-mail: giovanni.manzini@rse-web.it [RSE, Milano (Italy); Xiao, J., E-mail: jianjun.xiao@kit.edu [IKET, KIT, Karlsruhe (Germany); Vyskocil, L., E-mail: vyl@ujv.cz [UJV Rez (Czech Republic); Siccama, N.B., E-mail: siccama@nrg.eu [NRG, Safety and Power (Netherlands); Huhtanen, R., E-mail: risto.huhtanen@vtt.fi [VTT, PO Box 1000, FI-02044 VTT (Finland)

    2015-02-15

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.

  17. Improvement of the fracture toughness in the PWR pressurizer surgeline material (Final report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Lee, Bong Sang; Oh, Yong Jun; Yoon, Ji Hyun; Oh, Jong Myung; Park, Buk Gyun; Kim, Ju Hak; Kuk Il Hyun; Byun, Taek Sang [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Fracture toughness property of PWR surgeline pipe materials is one of the most important factor for Leak Before Break(LBB) analysis. In order to improve fracture toughness of the surgeline material (SA312-TP347 stainless steel), base on the evaluation and analysis of the commercial TP347 alloys, eleven TP347 model alloys were designed and manufactured. Tensile and fracture resistance properties of the model alloys, as well as microstructure, were evaluated. It is concluded that the nitrogen shall be added more than 0.1% for high tensile property and the carbon shall be in the range of 0.02 to 0.04% for high fracture resistance. In addition, four TP316N stainless steels were manufactured and evaluated to find out the applicability as a candidate material for PWR surgeline pipe. As a conclusion, TP316N stainless steels had an excellent property to be used for surgeline piping materials, substituting the present TP347 stainless steels. (author). 11 refs., 50 figs., 12 tabs.

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

    International Nuclear Information System (INIS)

    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)

  19. A quantitative comparison of loading pattern optimization methods for in-core fuel management of PWR

    International Nuclear Information System (INIS)

    The performance of several loading pattern(LP) optimization methods was quantitatively compared through a benchmark problem of PWR LP optimization. The simulated annealing(SA) method, the genetic algorithms(GA) method, the direct search(DS) method based on assembly multiple shuffling and the binary exchange(BE) method based on fuel assembly binary exchange were investigated as candidates for the optimization techniques. Hybrid strategy which combined different optimization methods was newly proposed, and the performances of two different new hybrid methods, which combined DS with BE and GA with BE were examined. From the results of the LP optimization benchmark problem, the superiority and inferiority of each method were clarified. Furthermore, it was demonstrated that the GA+BE hybrid strategy performed best among these methods. By combining GA with BE, the weaknesses of these two methods were compensated for with each other and the optimization performance was improved significantly. Therefore, the GA+BE hybrid method is quite effective for the LP optimization problems of PWR. (author)

  20. Analytical one-dimensional frequency response and stability model for PWR nuclear power plants

    International Nuclear Information System (INIS)

    A dynamic model for PWR nuclear power plants is presented. The plant is assumed to consist of one-dimensional single-channel core, a counterflow once-through steam generator (represented by two nodes according to the nonboiling and boiling region) and the necessary connection coolant lines. The model describes analytically the frequency response behaviour of important parameters of such a plant with respect to perturbations in reactivity, subcooling or mass flow (both at the entrances to the reactor core and/or the secondary steam generator side), the perturbations in steam load or system pressure (on the secondary side of the steam generator). From corresponding 'open' loop considerations it can then be concluded - by applying the Nyquist criterion - upon the degree of the stability behaviour of the underlying system. Based on this theoretical model, a computer code named ADYPMO has been established. From the knowledge of the frequency response behaviour of such a system, the corresponding transient behaviour with respect to a stepwise or any other perturbation signal can also be calculated by applying an appropriate retransformation method, e.g. by using digital code FRETI. To demonstrate this procedure, a transient experimental curve measured during the pre-operational test period at the PWR nuclear power plant KKS Stade was recalculated using the combination ADYPMO-FRETI. Good agreement between theoretical calculations and experimental results give an insight into the validity and efficiency of the underlying theoretical model and the applied retransformation method. (Auth.)

  1. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    International Nuclear Information System (INIS)

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs

  2. Modeling of the thermo-mechanical behaviour of the PWR fuel

    International Nuclear Information System (INIS)

    This article reviews the various physical phenomena that take place in an irradiated fuel rod and presents the development of the thermo-mechanical codes able to simulate them. Though technically simple the fuel rod is the place where appear 4 types of process: thermal, gas behaviour, mechanical and corrosion that combine involving 5 elements: the fuel pellet, the fuel clad, the fuel-clad gap, the inside volume and the coolant. For instance the pellet is the place where the following mechanical processes took place: thermal dilatation, elastic deformation, creep deformation, densification, solid swelling, gaseous swelling and cracking. The first industrial code simulating the behaviour of the fuel rod was COCCINEL, it was developed by AREVA teams from the American PAD code that was included in the Westinghouse license. Today the GALILEO code has replaced the COPERNIC code that was developed in the beginning of the 2000 years. GALILEO is a synthesis of the state of the art of the different models used in the codes validated for PWR and BWR. GALILEO has been validated on more than 1500 fuel rods concerning PWR, BWR and specific reactors like Siloe, Osiris, HFR, Halden, Studsvik, BR2/3,...) and also for extended burn-ups. (A.C.)

  3. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    International Nuclear Information System (INIS)

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system

  4. Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

    International Nuclear Information System (INIS)

    In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

  5. AP1000® PWR reactor physics analysis with VERA-CS and KENO-VI. Part 2. Power distribution

    International Nuclear Information System (INIS)

    Westinghouse has applied the Core Simulator of the Virtual Environment for Reactor Applications, VERA-CS, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000® PWR. The AP1000 PWR features an advanced first core with radial and axial heterogeneities, including enrichment zoning, multiple burnable absorbers, and a combination of light and heavy control banks to enable the MSHIMTM advanced operational strategy. These advanced features make application of VERA-CS to the AP1000 PWR first core especially relevant to qualify VERA performance. A companion paper at this conference describes the results obtained with VERA-CS and the KENO Monte-Carlo code for startup physics tests simulations of the AP1000 PWR first core (critical boron, rod worth and reactivity coefficients). This paper describes the results of detailed power distribution comparisons between VERA-CS and KENO, and confirms the excellent numerical agreement reported in the companion paper for the startup physics tests simulations. (author)

  6. Diagnosis and prognosis of the source term by the French Safety Institut during an emergency on a PWR

    International Nuclear Information System (INIS)

    The French approach for the diagnosis and the prognosis of the source term during an accident on a PWR is presented and the tools which have been developed to implement this approach at the Institute for Nuclear Protection and Safety (IPSN) are described. (author). 2 refs, 3 figs

  7. Research on General Corrosion Property of 304L and 304NG Stainless Steels in Simulated PWR Primary Water

    Institute of Scientific and Technical Information of China (English)

    PENG; De-quan; HU; Shi-lin; ZHANG; Ping-zhu; WANG; Hui

    2012-01-01

    <正>The general corrosion behaviors of 304L and 304NG grade stainless steels in simulated pressurized water reactor (PWR) primary loop were studied using still autoclave, respectively, the corrosion test lasted for 1 680 hours. The corrosion oxide films were analyzed macroscopically and microscopically. The results are shown in Figs. 1, 2.

  8. Influence of localized deformation on A-286 austenitic stainless steel stress corrosion cracking in PWR primary water

    International Nuclear Information System (INIS)

    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 γ' precipitates Ni3(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 γ'-free localized deformation bands on A-286 SCC by means of comparative CERT tests performed on specimens with similar yield strength, containing or not γ'-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)

  9. Effect of surface finish and loading conditions on the LCF behavior of austenitic stainless steel in PWR environment

    International Nuclear Information System (INIS)

    ANL has issued in 2006 a NUREG/CR-6909 report that is now applicable in the US for evaluations of PWR environmental effects in the fatigue analysis of new reactor components. In order to assess the conservativeness of the application of this NUREG report, low cycle fatigue (LCF) tests were performed by AREVA NP on austenitic stainless steel specimens in a PWR environment. The selected material exhibits in an air environment a fatigue behavior consistent with the ANL reference 'air' mean curve. Tests were performed in PWR environment for two various loading conditions: for fully reverse triangular signal (for comparison purpose with tests performed by other laboratories with same loading conditions) and complex signal, simulating strain variation for actual typical PWR thermal transients. Two surface finish conditions were tested: polished and ground. This paper presents the comparison of environmental penalty factors (Fen) as observed experimentally with the ANL formulation (considering the strain integral method for complex loading), and the actual fatigue life of the specimen with the fatigue life predicted through the NUREG/CR-6909 application

  10. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation

  11. Calculation of releases of radioactive materials in gaseous and liquid effluents from pressurized water reactors (PWR-GALE Code). Revision 1

    International Nuclear Information System (INIS)

    This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50

  12. Happy (Chinese) New Year!

    Science.gov (United States)

    Johnson, Georgia G.

    1979-01-01

    Suggestions are made for a classroom celebration of Chinese New Year, including discussion of the Chinese calendar and customs, a short list of appropriate children's stories, and food ideas, including a recipe for fortune cookies. (SJL)

  13. Chinese restaurant syndrome

    Directory of Open Access Journals (Sweden)

    Balachandran C

    1991-01-01

    Full Text Available A 24-year-old Chinese student with history of recurrent attacks of flushing with burning and dryness of face of 4 years duration showed exacerbation of the symptoms after oral provocation with 1 mg of Chinese salt. Patient was treated with 50 mg pyridoxine daily and restriction of the Chinese salt in diet with moderate improvement.

  14. Epilepsy in Chinese culture.

    Science.gov (United States)

    Lee, T M; Yang, S H; Ng, P K

    2001-01-01

    This article reviewed the meaning and development of epilepsy in Chinese culture. The theories of Yin and Yang and the five elements, fundamentals of traditional Chinese medicine, were introduced, which form the context of understanding of the etiology, classification, and treatment of epilepsy in traditional Chinese medicine. PMID:11321477

  15. TRADITIONAL CHINESE HERBAL MEDICINE

    NARCIS (Netherlands)

    ZHU, YP; WOERDENBAG, HJ

    1995-01-01

    Herbal medicine, acupuncture and moxibustion, and massage and the three major constituent parts of traditional Chinese medicine. Although acupuncture is well known in many Western countries, Chinese herbal medicine, the mos important part of traditional Chinese medicine, is less well known in the We

  16. Global stratification effects in feedwater line to steam generator in PWR power plant

    International Nuclear Information System (INIS)

    Recent measurements recorded at Beaver Valley Power Station-Unit 1, a PWR plant, demonstrate the presence of significant global thermal stratification within horizontal sections of the feedwater line. Global thermal stratification has not been documented nor its effects quantified on pressurized water reactor feedwater piping. An investigation of the feedwater line at BV-1 was performed to determine the effects of global stratification. Plant specific operating data were gathered for a recent plant restart using temporary instrumentation installed to measure piping temperatures and displacements. Data reduction identified significant global thermal stratification and significant pipe movement during several operating modes. Steps were taken to incorporate its effects into the design calculations when the global thermal stratification was determined to be an unanalyzed condition which was not enveloped by the existing design basis calculations

  17. Calculation of activity content and related properties in PWR and BWR fuel using ORIGEN 2

    International Nuclear Information System (INIS)

    This report lists the conditions for calculations of the core inventory for a PWR and BWR. The calculations have been performed using the computer code ORIGEN 2. The amount (grams), the total radioactivity (bequerels), the thermal power (watts), the radioactivity from theα-decay (bequerels), and the neutron emission (neutrons/sec) from the core after the last burnup have been determined. All the parameters have been calculated as a function of the burnup and the natural decay, the latter over a time period of 0-1.0E07 years. The calculations have been performed for 68 heavy nuclides, 60 daughter nuclides, to the heavy nuclides with atomic numbers under 92, 852 fission products and 7 light nucli ides. The most important results are listed. (author)

  18. Computational analysis for prediction of pressure of PWR presurizer undertransient conditions

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A computer model has been developed for prediction of the pressure in thepressurizer undertransient conditions. In the model three separate thermodynamic regions which arenot required to be inthermal equilibrium have been considered. The mathematical model derived from the general conservation equations includesall of theimportant thermal-hydraulics phenomena occurring in the pressurizer,i.e., stratificationof the hot water andincoming cold water, bulk flashing and condensation, wall condensation, andinterfacial heat and masstransfer, etc. The bubble rising and rain-out models are developed to describe bulkflashing andcondensation, respectively. To obtain the wall condensation rate, a one-dimensionalheat conductionequation is solved by the pivoting method. The presented model will predict thepressure-time behaviorof a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreementwithavailable data on the CHASHMA nuclear power plant's pressurizer performance.

  19. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail: rochkhudson@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  20. Representing Operational Knowledge of PWR Plant by Using Multilevel Flow Modelling

    DEFF Research Database (Denmark)

    Zhang, Xinxin; Lind, Morten; Jørgensen, Sten Bay;

    2014-01-01

    The aim of this paper is to explore the capability of representing operational knowledge by using Multilevel Flow Modelling ( MFM ) methodology. The paper demonstrate s how the operational knowledge can be inserted into the MFM models and be used to evaluate the plant state, identify the current...... systems. And the sub - systems’ functions will be decomposed into sub - models according to different operational situations. An operational model will be developed based on the operating procedure by using MFM symbols and this model can be used to implement coordination rules for organize the utilizati...... situation and support operational decisions. This paper will provide a general MFM model of the primary side in a standard Westinghouse Pressurized Water Reactor ( PWR ) system including sub - systems of Reactor Coolant System, Rod Control System, Chemical and Volume Control System, emergency heat removal...

  1. Pressurizer level measurement inside PWR nuclear plant using resistance type heat sensors

    International Nuclear Information System (INIS)

    The accident that occured in 1979 to the PWR type nuclear reactor, Three-Mile Island 2, has drawn attention to the maladjustement of the differentiel pressure level detector installed in nuclear plants on the market. A system is presented here for measuring the level in pressurizers based on measurements of the heat resistance of the boundary layer existing between the heated sensor and the fluid mass in the vessel. The sensor consists of a 3 cm diameter cylindrical insulator support around which a 0.1 mm diameter platinum filament is wound. This filament simultaneously fulfills heating and transducer functions. To verify the feasibility of the resistant type heat sensor a test system, which provides water and steam under pressure was realised. Static and dynamic tests have shown that the principle of the resistant heat sensor is viable and can be used to obtain level informations

  2. Monte-Carlo determination of dose rates in spherical PWR shield

    International Nuclear Information System (INIS)

    An operating nuclear reactor is a source of potentially dangerous nuclear radiation. It emits many different types of nuclear radiations. However, the neutrons and gamma rays are the main sources of radiation to give the contributions to the radiation situation in the reactor during operation, and the principal concern of reactor shielding. In current study the neutron and gamma radiation dose rates at different depths in concrete bio-shield of a PWR are calculated in spherical model by M-C simulation with using MCNP4C2. The simulation results are compared with the results obtained from similar calculations based on S8P3 spherical approximation with using the ANISN code. (author)

  3. Parametric study on parallel flow induced damping of PWR fuel assembly

    International Nuclear Information System (INIS)

    This paper reports on a mechanism of parallel flow-induced changes in vibrational characteristics of PWR fuel assemblies that has been studied through a series of hydraulic tests using reduced-and full-scale prototype mockups. Measured data and analytical evaluations showed the phenomenon stands on essentially the same basis as the dynamics and stability of flexible cylinders subjected to a parallel flow. In the mathematical model, the effects of rod bundle geometries and boundaries formed by walls or adjacent bundles can be exactly incorporated in the form of added mass coefficients, velocity coupling coefficients and other fluid forces. From a full scale test, it has been shown that coolant temperature has little effect up to reactor operating conditions. The updated FEM model has been verified to be applicable in describing the vibrational characteristics of from an isolated cylinder to a full scale fuel assembly in terms of the consistent properties

  4. Considerations on the manner to account for fast fracture risk in the design of PWR vessels

    International Nuclear Information System (INIS)

    The way followed in France for analyzing fast fracture resistance of PWR primary components is the one of a deterministic analysis with safety coefficients imposed in the fracture criteria. The study of margins towards fast fracture of the 900 MWe program vessels undertaken in 1982 includes parametric evaluations of the influence of essential variables. It has stimulated further thoughts on the level of safety to fix in the analysis methodology, on the orientations for the choice of safety factors and on the manner to introduce them in the analysis. A first chapter tries to characterize the French approach in comparison to those of other countries. A second chapter examines the manner according to which safety factors can be introduced in the deterministic analysis. It presents the principle for a logical approach accounting for the interdependency of all factors and variables. It establishes criteria for the selection of defect kind and size for the computation

  5. PWR refill-reflood analysis with experimental loop and calculation model. Pt. 2

    International Nuclear Information System (INIS)

    Equations for control volumes varying in the time have been applied. The bottom and length of the bubble and film boiling region in the core are specified by a correlation and time constant based on our measurements. The boiling volume is divided into two parts, saturated water and steam volume. The hydraulic processes are calculated to the average fuel rod, but for the temperatures also the hot rod is calculated. Some parameters have been determined by comparison of measured and calculated results. Sensitivity analyses were made for a PWR, and the hydraulic resistance of the pump (and water stopper evtl. in the loop) was found as the most important factor to ensure a sufficient reflood. (orig.)

  6. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

  7. Dilute chemical cleaning of PWR steam generators off-line cleaning process evaluation

    International Nuclear Information System (INIS)

    This project evaluated the feasibility of using a low-concentration (approx. 0.5 wt %) chemical cleaning process to remove corrosion product deposits from steam generator surfaces and magnetite from tube-to-support plate crevices of PWR steam generators. The primary objective was to develop a dilute process that could be safely applied at scheduled intervals, such as during normal refueling outages, to maintain a clean operating condition in the steam generator. The dilute chemical cleaning process developed in this project was demonstrated successfully on two model generators which were operated on faulted chemistry by DOE/CRC at Commonwealth's State Line Facility. Unit 5 was cleaned after 48 days of operation with 1% seawater fouling, and Unit 6 was cleaned after 112 days of operations with Lake Michigan water. This report describes work leading to the model generator cleaning demonstrations and provides details of the cleaning operation for each model steam generator

  8. Assessment of Welding System Modification of The Candu and PWR Fuel Element Types end Plug

    International Nuclear Information System (INIS)

    To anticipate future possibility of a nuclear fuel element industry in Indonesia, research on other types of nuclear fuel element beside Cirene type has to be done. It can be accomplished, one of them, by modifying the already available equipment. Based on the sheath material, the sheath dimension and the welding process parameters such as welding current and welding cycles, the available Magnetic Force Welding can be used for welding end plug of Candu nuclear fuel element by modifying some of its components (tube clamp, plug clamp, etc). The available Pellet drying and element filling furnace with its supporting system with includes helium gas filling, welding chamber, argon gas supply, vacuum system, sheath clamp and sheath driving system can be used for welding end plug with sheath of PWR nuclear fuel element by adding og Tungsten inert Gas (TIG) welding machine in the welding chamber and modifying a few components (seal clamp, sheath clamp)

  9. INETEC new system for inspection of PWR reactor pressure vessel head

    International Nuclear Information System (INIS)

    INETEC Institute for Nuclear Technology developed new equipment for inspection of PWR and VVER reactor pressure vessel head. The new advances in inspection technology are presented in this article, as the following: New advance manipulator for inspection of RPVH with high speed of inspection possibilities and total automated work; New sophisticated software for manipulator driving which includes 3D virtual presentation of manipulator movement and collision detection possibilities; New multi axis controller MAC-8; New end effector system for inspection of penetration tube and G weld; New eddy current and ultrasonic probes for inspection of G weld and penetration tube; New Eddy One Raster scan software for analysis of eddy current data with mant advanced features which allows easy and quick data analysis. Also the results of laboratory testing and laboratory qualification are presented on reactor pressure vessel head mock, as well as obtained speed of inspection and quality of collected data.(author)

  10. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1999-09-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

  11. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses -- Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    2000-03-16

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotopes) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data, usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, was considered to be of sufficient quality for depletion code validation.

  12. Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Taniguchi, T.; Ligon, D.; Stamatelatos, M.

    1983-04-01

    Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights in the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.

  13. Production of 232U from irradiation of standard and thorium-based fuels in PWR reactors

    International Nuclear Information System (INIS)

    The production of small quantities of 232U can induce radiation protection issues in the back end of the fuel cycle, particularly for thorium-based fuels. This is due to its relatively short half life (69 years) and the emission of a high energy gamma ray of 2.6 MeV at the end of its decay chain. With the depletion code MURE, we determine the different reactions pathways, and their proportions, leading to the synthesis of 232U in UO2 and (Th,Pu)O2 fuels irradiated in a PWR. Moreover, the impact, on the 232U production, of cycle times such as time separating the fabrication of the fuel and its irradiation as well as influence of the fissile content has been investigated for UO2 fuel. The impact of the thorium ore provenance and of the plutonium quality has been studied for the (Th,Pu)O2 case. (author)

  14. Verification of ASTRA Code with PWR MOX/UO2 Transient Benchmark Problem

    Energy Technology Data Exchange (ETDEWEB)

    Han, Tae Young; Yoon, Joo Il; Kim, Jae Hee; Lee, Chang Kyu; Cho, Beom Jin [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2010-10-15

    In recent, ASTRA (Advanced Static and Transient Reactor Analyzer) has been successfully developed by KNF (KEPCO Nuclear Fuel) as a nuclear design code for commercial reactor core. This code has the capability of the multi-group analysis because of the requirement of a neutron flux solver to simulate a core not only for UO2-fueled but also MOX-UO2 fueled. In addition, ASTRA has been designed to analyze the core characteristics under transient condition as control rod ejection accident. In this paper, we have performed the benchmark analysis with the PWR MOX/UO2 control rod ejection problem provided by OECD/NEA and U.S. NRC for the purpose of verifying these capabilities of ASTRA

  15. A PWR plant model for the analysis of large amplitude transients

    International Nuclear Information System (INIS)

    The PWR transient code ALMOD has been developed to cover a wide range of transient and accident simulation in safety analysis, comprising failure of safety system components (e.g. analysis of anticipated transients without scram=ATWS). Because of the large amplitudes to be expected during the transients, simplified models such as linearized models, used in control system analysis, are not applicable here. As the transients have to be analyzed over minutes, feedback from the entire coolant system becomes effective, thus requiring the simulation of core and both primary and secondary coolant system. Because of the long duration of the transients special emphasis has been put on computational speed. Key variables of interest in transient analysis are fuel and cladding temperature as well as primary and secondary system pressure. Extreme plant conditions such as two phase flow in the primary coolant system, filling of the pressurizer with water etc. have to be simulated with sufficient accuracy. (orig.)

  16. The typical aging problem of the main component materials in PWR nuclear steam supply system

    International Nuclear Information System (INIS)

    The aging problem of equipment, components, and materials, is a systematic problem which exists throughout the whole process of NPP from design to retirement .In this paper, the basic information and current status of aging research and aging management is introduced, with detailed information about the typical aging problem and research work of main component materials in PWR nuclear steam supply system. Systematical studies have shown that: 1) the aging management work of NPP in China has started a little late and the research foundation is relatively weak; 2)the technology system and management system of aging management and lifetime assessment has not yet been formed; 3) the research work about aging problem of component materials concentrates only on the design verification except for the reactor pressure vessel , and there's no specialized research work on aging problem; 4)there's a lack of important data on aging evaluation and lifetime enhancement analysis . A systematical research work on component aging problem is suggested. (authors)

  17. Design and static simulation of secondary loop of small PWR nuclear power plants

    International Nuclear Information System (INIS)

    A computer program that has been developed with the purpose of making easier the decisions concerning the design of the secondary loop of small PWR nuclear power plants through numerical experiments of low running costs and short time is presented. Initially, the first part of the computer program is described. It aims to preliminarily design several major components of the secondary circuit from user-defined design conditions. Next, the second part of the computer program is presented. It simulates the steady state operation at part-load conditions of the preliminary design of the plant by generating and solving systems of simultaneous nonlinear algebraic equations, their number varying from 17 to 107. The computer program has been tested for several application cases. The program results are discussed in the last part of the work, along with several aspects to be added to the program in future works. (author)

  18. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  19. Study on temperature and flow fluctuation phenomena in pressurizer spray pipe of PWR

    International Nuclear Information System (INIS)

    Thermal fatigue cracking may initiate in pressurizer spray pipe of PWR where oscillations of water surface may exist. In order to clarify the flow and thermal conditions in the pressurizer spray pipe, an experiment was conducted for a steam-water flow using a mock-up pressurizer spray pipe. It was shown that the fluid temperature fluctuations were not caused by the waves on the water surface, but were caused by temperature fluctuations in water layer below the interface. Visualization tests were conducted to investigate the detail mechanism of the fluctuation using the rectangular pipe made of polycarbonate resin. The clouds were observed below the steam-water interface and it seemed like an internal wave. It was shown that the maximum of the temperature fluctuations increases with increase the steam-water temperature difference and the prominent frequency of the fluctuations was nearly equal to 0.1Hz∼1Hz. (author)

  20. Numerical modeling of in-vessel melt water interaction in large scale PWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Siemens AG, KWU NA-M, Erlangen (Germany)

    1998-01-01

    This paper presents a comparison between IVA4 simulations and FARO L14, L20 experiments. Both experiments were performed with the same geometry but under different initial pressures, 51 and 20 bar respectively. A pretest prediction for test L21 which is intended to be performed under an initial pressure of 5 bar is also presented. The strong effect of the volume expansion of the evaporating water at low pressure is demonstrated. An in-vessel simulation for a 1500 MW el. PWR is presented. The insight gained from this study is: that at no time are conditions for the feared large scale melt-water intermixing at low pressure in force, with this due to the limiting effect of the expansion process which accelerates the melt and the water into all available flow paths. (author)

  1. HEXBU-3D, a three-dimensional PWR-simulator program for hexagonal fuel assemblies

    International Nuclear Information System (INIS)

    HEXBU-3D is a three-dimensional nodal simulator program for PWR reactors. It is designed for a reactor core that consists of hexagonal fuel assemblies and of big follower-type control assemblies. The program solves two-group diffusion equations in homogenized fuel assembly geometry by a sophisticated nodal method. The treatment of feedback effects from xenon-poisoning, fuel temperature, moderator temperature and density and soluble boron concentration are included in the program. The nodal equations are solved by a fast two-level iteration technique and the eigenvalue can be either the effective multiplication factor or the boron concentration of the moderator. Burnup calculations are performed by tabulated sets of burnup-dependent cross sections evaluated by a cell burnup program. HEXBY-3D has been originally programmed in FORTRAN V for the UNIVAC 1108 computer, but there is also another version which is operable on the CDC CYBER 170 computer. (author)

  2. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  3. PRETTA:A COMPUTER PROGRAM FOR PWR PRESSURIZER’S TRANSIENT THERMODYNAMICS

    Institute of Scientific and Technical Information of China (English)

    阿谢德; 徐济鋆

    2001-01-01

    A computer program PRETTA “Pressurizer Transient Thermodynamics Analysis” was developed for the prediction of pressurizer under transient conditions. It is based on the solution of the conservation laws of heat and mass applied to the three separate and non equilibrium thermodynamic regions. In the program all of the important thermal-hydraulics phenomena occurring in the pressurizer: stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer have been considered. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented computer program will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant's pressurizer performance.

  4. A neural networks based ''trip'' analysis system for PWR-type reactors

    International Nuclear Information System (INIS)

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients'inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author)

  5. The key to superior water chemistry at a PWR nuclear station

    International Nuclear Information System (INIS)

    This paper demonstrates how a condensate polishing unit can be successfully used to treat the feedwater for circulating-type pressurized water reactors (PWRs). Water chemistry at the Salem Generating Station, a two-unit, four-loop Westinghouse PWR located in New Jersey, is discussed. Topics considered include a plant description and the history of early operation, the role of constant surveillance, makeup water quality, the effect of freezing on gel-type anion exchange resin, a total organic carbon (TOC) survey, steam generator chemistry, steam generator inspection, condensate polisher operation, and management philosophy. The SEPREX condensate polishing process, in which the complete separation of the anion exchange resin from the cation exchange resin is achieved by flotation separation, is examined. It is concluded that the utilization of a condensate polishing process such as SEPREX provides the operating personnel at the plant with the necessary means to maintain the minimum desired level of contaminants within the steam generator

  6. Preliminary evaluation of the transportation cost for PWR spent fuels in Korea

    International Nuclear Information System (INIS)

    It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper we established a cost estimate model based on the conceptual design of a transportation system and a logistics analysis. Using the developed model, the transportation cost for PWR spent fuels from the Yonggwang nuclear power plants to a centralized interim storage facility and a final repository was estimated. The result shows that the unit transportation cost is around 36 million won per one TU of spent fuel

  7. Effect of Weld Properties on the Crush Strength of the PWR Spacer Grid

    Directory of Open Access Journals (Sweden)

    Kee-nam Song

    2012-01-01

    Full Text Available Mechanical properties in a weld zone are different from those in the base material because of different microstructures. A spacer grid in PWR fuel is a structural component with an interconnected and welded array of slotted grid straps. Previous research on the strength analyses of the spacer grid was performed using base material properties owing to a lack of mechanical properties in the weld zone. In this study, based on the mechanical properties in the weld zone of the spacer grid recently obtained by an instrumented indentation technique, the strength analyses considering the mechanical properties in the weld zone were performed, and the analysis results were compared with the previous research.

  8. Calculation and analysis of heat source of PWR assemblies based on Monte Carlo method

    International Nuclear Information System (INIS)

    When fission occurs in nuclear fuel in reactor core, it releases numerous neutron and γ radiation, which takes energy deposition in fuel components and yields many factors such as thermal stressing and radiation damage influencing the safe operation of a reactor. Using the three-dimensional Monte Carlo transport calculation program MCNP and continuous cross-section database based on ENDF/B series to calculate the heat rate of the heat source on reference assemblies of a PWR when loading with 18-month short refueling cycle mode, and get the precise values of the control rod, thimble plug and new burnable poison rod within Gd, so as to provide basis for reactor design and safety verification. (authors)

  9. Fatigue Crack Growth Rate of Type 347 Stainless Steel at the PWR Environment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Materials used in nuclear power plants are low alloy steel, stainless steel, and superalloy steel. Understanding the characteristics of these materials is important in the development of nuclear power plant related technology. Nb-stabilized Type 347 stainless steel is used for the coolant pressurizer surge line of Korea Standard Nuclear Power Plant (KSNPP). Surge line of PWR nuclear reactor are damaged by thermal fatigue due to thermal gradient during heat-up and cool-down, mechanical fatigue due to mechanical stress, and corrosion fatigue due to nuclear reactor water environment. Fatigue is an important factor which limits the life of structure. Fatigue crack growth rate curves in nuclear reactor environment are needed to evaluate the integrity of nuclear reactor structure but that result is not sufficient. In this study, fatigue crack growth rates at nuclear reactor environment are produced to evaluate integrity of nuclear power plant section 5

  10. Optimization of thermal efficiency of nuclear central power like as PWR

    International Nuclear Information System (INIS)

    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

  11. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    CERN Document Server

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  12. The radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR

    CERN Document Server

    Kelly, G N; Charles, D; Hemming, C R

    1983-01-01

    This report contains an assessment of the radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR. Three of the degraded core accident releases postulated by the CEGB are analysed. The consequences, conditional upon each release, are evaluated in terms of the health impact on the exposed population and the impact of countermeasures taken to limit the exposure. Consideration is given to the risk to the Greater London population as a whole and to individuals within it. The consequences are evaluated using the NRPB code MARC (Methodology for Assessing Radiological Consequences). The results presented in this report are all conditional upon the occurrence of each release. In assessing the significance of the results, due account must be taken of the frequency with which such releases may be predicted to occur.

  13. Application of the ballooning analysis code MATARE on a generic PWR fuel assembly

    International Nuclear Information System (INIS)

    The MATARE (MAbel-TAlink-RElap) code is a new multi-pin deformation analysis code created through the dynamic coupling between the thermal-hydraulic code RELAP5 and multiple instances of the single-pin thermal-mechanics code MABEL. A multi-pin representation of different zones of a typical PWR fuel assembly under post-LOCA reflooding conditions was analysed including some of the most relevant features that characterise a typical nuclear reactor fuel assembly and evaluate their effect on the behaviour of the fuel rods under conditions leading to clad ballooning. The code was able to simulate the deformation of wide regions of a fuel assembly under reflood conditions and has shown how differences in pin pressure and the presence of rod with burnable poisons and control rod guide thimbles also contribute to a substantial incoherent ballooning in agreement with the experimental data. (author)

  14. CAE advanced reactor demonstrators for CANDU, PWR and BWR nuclear power plants

    International Nuclear Information System (INIS)

    CAE, a private Canadian company specializing in full scope flight, industrial, and nuclear plant simulators, will provide a license to IAEA for a suite of nuclear power plant demonstrators. This suite will consist of CANDU, PWR and BWR demonstrators, and will operate on a 486 or higher level PC. The suite of demonstrators will be provided to IAEA at no cost to IAEA. The IAEA has agreed to make the CAE suite of nuclear power plant demonstrators available to all member states at no charge under a sub-license agreement, and to sponsor training courses that will provide basic training on the reactor types covered, and on the operation of the demonstrator suite, to all those who obtain the demonstrator suite. The suite of demonstrators will be available to the IAEA by March 1997. (author)

  15. Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference

    CERN Document Server

    Castro, Emilio; Buss, Oliver; Garcia-Herranz, Nuria; Hoefer, Axel; Porsch, Dieter

    2016-01-01

    A Monte Carlo-based Bayesian inference model is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous...

  16. New technical knowledge to be implemented to the revision of rules on pipe wall thinning management for PWR Plants. 2006 edition

    International Nuclear Information System (INIS)

    Rules for PWR plant pipe wall thinning management were formulated by the Japan Society of Mechanical Engineers in 2006. Since then thinning management of Japanese PWR plants has been carried out based on this rule. Pipe wall thinning phenomena to be dealt with in this rule have been identified in many piping components of power plants. New technical knowledge has been accumulated since the issuance of 2006 edition. We have formulated these knowledge and information about the thinning phenomena in PWR power plants. Given the history of application of this rule, we have to make our best effort to carry out a study of latest technical knowledge and implement them to the revision of rule and improve pipe wall thinning management. This paper summarizes the new technical knowledge and basis to be implemented to the revision of rules on pipe wall thinning management for PWR plants in Japan. (author)

  17. Singular deposit formation in PWR due to electrokinetic phenomena - application to SG clogging

    Energy Technology Data Exchange (ETDEWEB)

    Guillodo, M.; Muller, T.; Barale, M.; Foucault, M. [AREVA NP SAS, Technical Centre (France); Clinard, M.-H.; Brun, C.; Chahma, F. [AREVA NP SAS, Chemistry and Radiochemistry Group (France); Corredera, G.; De Bouvier, O. [Electricite de France, Centre d' Expertise de I' inspection dans les domaines de la Realisation et de l' Exploitation (France)

    2009-07-01

    The deposits which cause clogging of the 'foils' of the tube support plates (TSP) in Steam Generators (SG) of PWR present two characteristics which put forward that the mechanism at the origin of their formation is different from the mechanism that drives the formation of homogeneous deposits leading to the fouling of the free spans of SG tubes. Clogging occurs near the leading edge of the TSP and the deposits appear as diaphragms localized between both TSP and SG tubing materials, while the major part of the tube/TSP interstice presents little or no significant clogging. This type of deposit seems rather comparable to the ones which were reproduced in Lab tests to explain the flow rate instabilities observed on a French unit during hot shutdown in the 90's. The deposits which cause TSP clogging are owed to a discontinuity of the streaming currents in the vicinity of a surface singularity (orifices, scratches ...) which, in very low conductivity environment, produce local potential variations and/or current loop in the metallic pipe material due to electrokinetic effects. Deposits can be built by two mechanisms which may or not coexist: (i) accumulation of particles stabilized by an electrostatic attraction due to the local variation of electrokinetic potential, and (ii) crystalline growth of magnetite produced by the oxidation of ferrous ions on the anodic branch of a current loop. Lab investigations carried out by AREVA NP Technical Centre since the end of the 90's showed that this type of deposit occurs when the redox potential is higher than a critical value, and can be gradually dissolved when the potential becomes lower than this value which depends on the 'Material - Chemistry' couple. Special emphasis will be given in this paper to the TSP clogging of SG in PWR secondary coolant dealing particularly with the potential strong effect of electrokinetic phenomena in low conductive environment and in high temperature conditions

  18. Overview of the Vercors Programme Devoted to Safety Studies on Irradiated PWR Fuel

    International Nuclear Information System (INIS)

    The first objective of the Heva-Vercors Program is to improve the data of fission product release and behaviour after an extensive fuel temperature increase and loss of integrity of the fuel elements that occur in case of severe PWR accident. The program is co-funded by the French Nuclear Protection and Safety Institute (IPSN) and Electricite de France (EDF). The experiments are conducted in a shielded cell of the French Grenoble Nuclear Centre. For these tests, industrial fuel from French PWR reactor plants is used. In order to rebuild the short lived fission product inventory, a reirradiation is performed in the experimental Siloe reactor, prior to the test. Eight tests have been conducted in the frame of the Heva Program up to 2370 K in the 1983-1988 period. The main outcomes of these studies were linked to the volatile fission product release. This program has been extended by the Vercors one with higher fuel temperature (2600 K) and improved instrumentation: gamma spectrometry, emission tomography, metallography, scanning electron microscopy, energy dispersive X-ray analysis, X-ray diffraction are some of the experimental techniques used for on line and post test characterization. The knowledge of the behavior of low volatile fission product has been significantly improved with the six Vercors tests. The results of the Vercors 4 test (38 GWd/t(U), 2570 K, reducing atmosphere) are presented here as an example. The key parameters are exhibited. The next step of these studies will use the Vercors HT loop that is planned to be operational at the beginning of 1996 to reach fuel melting temperature. The first aim of these future tests is to study the behaviour of non volatile and transuranic elements. An even more sophisticated instrumentation is implemented to reach the goal. The use of MOX fuel, the interaction between fission product aerosols and structural materials (Ag-In-Cd) and the fuel granulometry effect will be the next steps of the experimental program

  19. AREVA's fuel assemblies addressing high performance requirements of the worldwide PWR fleet

    International Nuclear Information System (INIS)

    Taking advantage of its presence in the fuel activities since the start of commercial nuclear worldwide operation, AREVA is continuing to support the customers with the priority on reliability, to: >participate in plant operational performance for the in core fuel reliability, the Zero Tolerance for Failure ZTF as a continuous improvement target and the minimisation of manufacturing/quality troubles, >guarantee the supply chain a proven product stability and continuous availability, >support performance improvements with proven design and technology for fuel management updating and cycle cost optimization, >support licensing assessments for fuel assembly and reloads, data/methodologies/services, >meet regulatory challenges regarding new phenomena, addressing emergent performance issues and emerging industry challenges for changing operating regimes. This capacity is based on supplies by AREVA accumulating very large experience both in manufacturing and in plant operation, which is demonstrated by: >manufacturing location in 4 countries including 9 fuel factories in USA, Germany, Belgium and France. Up to now about 120,000 fuel assemblies and 8,000 RCCA have been released to PWR nuclear countries, from AREVA European factories, >irradiation performed or in progress in about half of PWR world wide nuclear plants. Our optimum performances cover rod burn ups of to 82GWD/tU and fuel assemblies successfully operated under various world wide fuel management types. AREVA's experience, which is the largest in the world, has the extensive support of the well known fuel components such as the M5'TM'cladding, the MONOBLOC'TM'guide tube, the HTP'TM' and HMP'TM' structure components and the comprehensive services brought in engineering, irradiation and post irradiation fields. All of AREVA's fuel knowledge is devoted to extend the definition of fuel reliability to cover the whole scope of fuel vendor support. Our Top Reliability and Quality provide customers with continuous

  20. Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2002-12-17

    This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

  1. Evaluation of Physical Characteristics of PWR Cores with Accident Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

    2015-10-15

    The accident tolerant fuels (ATF) considered in this work includes metallic microcell UO{sub 2} pellets and outer Cr-based alloy coating on cladding, which is being developed in KAERI (Korea Atomic Energy Research Institute). Chromium metals have been used in many fields because of its hardness and corrosion-resistance. The use of the chromium metal in nuclear fuel rod can enhance the conductivity of pellets and corrosion-resistance of cladding. The objective of this work is to study the neutronic performances and characteristics of the commercial PWR core loaded the ATF-bearing assemblies. In this work, we studied the PWR cores which are loaded with ATF assemblies to improve the safety of reactor core. The ATF rod consists of the metallic microcell UO2 pellet which includes chromium of 3.34 wt% and the outer 0.05mm thick coating of Cr-based alloy with atomic number ratio of 85:15. We performed the cycle-by-cycle reload core analysis from the cycle 8 at which the ATF fuel assemblies start to be loaded into the core. The target nuclear power plant is the Hanbit-3 nuclear power plant. From the analysis, it was found that 1) the uranium enrichment is required to be increased up to 5.20/4.70 wt% in order to satisfy a required cycle length of 480 EFPDs, 2) the cycle length for the core using ATF fuel assemblies with the same uranium enrichments as those in the reference UO{sub 2} fueled core is decreased from 480 EFPDs to 430 EFPDs.

  2. Single PWR spent fuel assembly heat transfer data for computer code evaluations

    International Nuclear Information System (INIS)

    The descriptions and results of two separate heat transfer tests designed to investigate the dry storage of commercial PWR spent fuel assemblies are presented. Presented first are descriptions and selected results from the Fuel Temperature Test performed at the Engine Maintenance and Disassembly facility on the Nevada Test Site. An actual spent fuel assembly from the Turkey Point Unit Number 3 Reactor with a decay heat level of 1.17 KW, was installed vertically in a test stand mounted canister/liner assembly. The boundary temperatures were controlled and the canister backfill gases were alternated between air, helium and vacuum to investigate the primary heat transfer mechanisms of convection, conduction and radiation. The assembly temperature profiles were experimentally measured using installed thermocouple instrumentation. Also presented are the results from the Single Assembly Heat Transfer Test designed and fabricated by Allied General Nuclear Services, under contract to the Department of Energy, and ultimately conducted by the Pacific Northwest Laboratory. For this test, an electrically heated 15 x 15 rod assembly was used to model a single PWR spent fuel assembly. The electrically heated model fuel assembly permitted various ''decay heat'', levels to be tested; 1.0 KW and 0.5 KW were used for these tests. The model fuel assembly was positioned within a prototypic fuel tube and in turn placed within a double-walled sealed cask. The complete test assembly could be positioned at any desired orientation (horizontal, vertical, and 250 from horizontal for the present work) and backfilled as desired (air, helium, or vacuum). Tests were run for all combinations of ''decay heat,'' backfill, and orientation. Boundary conditions were imposed by temperature controlled guard heaters installed on the cask exterior surface

  3. Study of power peak migration due to insertion of control bars in a PWR reactor

    International Nuclear Information System (INIS)

    This paper aims to present a study on the power distribution behavior in a PWR reactor, considering the intensity and the migration of power peaks as is the insertion of control rods in the core banks. For this, the study of the diffusion of neutrons in the reactor was adopted by computer simulation that uses the finite difference method for numerically solving the neutron diffusion equation to two energy groups in steady state and in symmetry of a fourth quarter core. We decided to add the EPRI-9R 3D benchmark thermal-hydraulic parameters of a typical power PWR. With a new configuration for the reactor, the positions of the control rods banks were also modified. Due to the new positioning of these banks in the reactor, there was intense power gradients, favoring the occurrence of critical situations and logically unconventional for operation of a nuclear reactor. However, these facts have led interesting times for the study on the power distribution behavior in the reactor, showing axial migration of power peaks and mainly the effect of the geometry of the core on the latter. Based on the distribution of power was evident the increase of the power in elements located in the central region of the reactor core and, concomitantly, the reduction in elements of its periphery. Of course, the behavior exhibited by the simulated reactor is not in agreement with that expected in an actual reactor, where the insertion of control rods banks should lead to reduced power throughout the core as evenly as possible, avoiding sharp power peaks, standardizing the burning fuel, controlling reactivity deviations and acting in reactor shutdown

  4. Effect of aging on the PWR Chemical and Volume Control System

    International Nuclear Information System (INIS)

    The PWR Chemical and Volume Control System (CVCS) is designed to provide both safety and non-safety related functions. During normal plant operation it is used to control reactor coolant chemistry, and letdown and charging flow. In many plants, the charging pumps also provide high pressure injection, emergency boration, and RCP seal injection in emergency situations. This study examines the design, materials, maintenance, operation and actual degradation experiences of the system and main sub-components to assess the potential for age degradation. A detailed review of the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Report (LER) databases for the 1988--1991 time period, together with a review of industry and NRC experience and research, indicate that age-related degradations and failures have occurred. These failures had significant effects on plant operation, including reactivity excursions, and pressurizer level transients. The majority of these component failures resulted in leakage of reactor coolant outside the containment. A representative plant of each PWR design (W, CE, and B and W) was visited to obtain specific information on system inspection, surveillance, monitoring, and inspection practices. The results of these visits indicate that adequate system maintenance and inspection is being performed. In some instances, the frequencies of inspection were increase in response to repeated failure events. A parametric study was performed to assess the effect of system aging on Core Damage Frequency (CDF). This study showed that as motor-operated valve (MOV) operating failures increased, the contribution of the High Pressure Injection to CDF also increased

  5. Evaluation of the environmental quality of cables at PWR plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Nakanishi, S. [Kansai Electric Power Co., INC., Osaka (Japan); Mito, K. [Hokkaido Electric Power Co., INC., Sapporo (Japan); Kikuchi, K. [Shikoku Electric Power Co., INC., Takamatsu (Japan); Konishi, M. [Kyushu Electric Power Co., INC., Fukuoka (Japan); Suzuki, S. [The Japan Atomic Power Co., INC., Tokyo (Japan); Yamamoto, T. [Mitsubishi Heavy Industries, LTD., Kobe (Japan)

    2001-07-01

    Safety-related cables used at PWR plants in Japan must pass the environmental qualification test in accordance with IEEE Std. 323 and 383. IEEE Std.383 describes the methods of accelerated aging of cables that corresponds to the aging during the in-service period. However, IEEE Std. 775, etc. raise questions about the process in which: 1) conditions for the accelerated thermal aging are estimated by extrapolating activation energy, which is evaluated by the Arrhenius plot obtained in relatively high temperature range, 2) irradiation at high dose rate (<10 kGy/h) is permitted, and 3) irradiation is carried out sequentially after the thermal treatment. As a result, it was found that the activation energies obtained through the thermal aging test in a relatively low temperature range were considerably smaller than those obtained through the thermal aging test in a high temperature range. As to the method of the superposition of time dependent data proposed in IEC1244-2, we obtained good prospects that it would be possible to predict deterioration in an irradiation atmosphere at a very low dose rate of an actual plant with a considerably high accuracy, though it is necessary to obtain more data. Furthermore, we conducted LOCA exposure tests after the long-term thermal/radiation aging tests on EP rubber insulated cables and confirmed that cable soundness was maintained even after the LOCA. As a result of evaluating the long-term thermal/radiation aging period under the environmental conditions of an actual PWR plant by means of the shift factors obtained by the method of the superposition of time dependent data, it was evaluated that it would be equivalent to the aging of maximum 145 years.

  6. Singular deposit formation in PWR due to electrokinetic phenomena - application to SG clogging

    International Nuclear Information System (INIS)

    The deposits which cause clogging of the 'foils' of the tube support plates (TSP) in Steam Generators (SG) of PWR present two characteristics which put forward that the mechanism at the origin of their formation is different from the mechanism that drives the formation of homogeneous deposits leading to the fouling of the free spans of SG tubes. Clogging occurs near the leading edge of the TSP and the deposits appear as diaphragms localized between both TSP and SG tubing materials, while the major part of the tube/TSP interstice presents little or no significant clogging. This type of deposit seems rather comparable to the ones which were reproduced in Lab tests to explain the flow rate instabilities observed on a French unit during hot shutdown in the 90's. The deposits which cause TSP clogging are owed to a discontinuity of the streaming currents in the vicinity of a surface singularity (orifices, scratches ...) which, in very low conductivity environment, produce local potential variations and/or current loop in the metallic pipe material due to electrokinetic effects. Deposits can be built by two mechanisms which may or not coexist: (i) accumulation of particles stabilized by an electrostatic attraction due to the local variation of electrokinetic potential, and (ii) crystalline growth of magnetite produced by the oxidation of ferrous ions on the anodic branch of a current loop. Lab investigations carried out by AREVA NP Technical Centre since the end of the 90's showed that this type of deposit occurs when the redox potential is higher than a critical value, and can be gradually dissolved when the potential becomes lower than this value which depends on the 'Material - Chemistry' couple. Special emphasis will be given in this paper to the TSP clogging of SG in PWR secondary coolant dealing particularly with the potential strong effect of electrokinetic phenomena in low conductive environment and in high temperature conditions. (author)

  7. Core design study for power uprating of integral primary system PWR

    International Nuclear Information System (INIS)

    Highlights: • We propose a 20% power uprate of an integral primary system PWR for a better economic performance. • The power uprate is achieved mainly by optimizing core radial power peaking. • By enrichment zoning, power peaking of the proposed core is found to be around 1.43. • The predicted steady state MDNBR of the proposed core is found to be 3.454. • The results show that the proposed core design satisfies all design criteria defined in this study. - Abstract: Core design study for power uprating of integral primary system PWR has been performed. The selected reference core design is a four-year straight burn option of the International Reactor Innovative and Secure (IRIS). The objective of this study is to evaluate the possibility of increasing thermal power of the IRIS reactor by 20%, for a better economic performance. This study includes core neutronic and thermal hydraulic aspects. The power uprate is achieved by minimizing core radial power peaking, and by increasing fuel enrichment and coolant mass flow rate. Parametric calculations have been carried out to optimize the power-uprated core design, with the objective of obtaining relatively low core power peaking and similar initial reactivity with that of the reference core. The optimization is achieved by means of enrichment zoning and application of burnable poison, with different configuration from that of the reference core. The performance of power-uprated core is compared with the reference core. The calculation results show that the proposed core-uprated design with thermal power of 1200 MWt shows preferable characteristics, such as low power peaking of around 1.43, negative reactivity coefficients, and relatively high MDNBR of 3.454

  8. Chinese Food in America

    OpenAIRE

    Jou, Diana T.

    2011-01-01

    How did Chinese food get to look like this? With more than 41,000 Chinese restaurants in America - 3 times the number of McDonald’s restaurants - Chinese food is one of the most accepted and misunderstood cuisines in the United States. From large cities to small towns, locals can always count on an order of orange chicken in a takeout box, with a few fortune cookies thrown in the bag. But what Americans view as Chinese food is far from a traditional Chinese meal, wh...

  9. Chinese Lunar Calendar

    Institute of Scientific and Technical Information of China (English)

    方陵生

    2005-01-01

    @@ Background and Concept The Chinese animal signs2 are a 12-year cycle used for dating the years. They represent a cyclical concept of time, rather than the Western linear concept of time. The Chinese Lunar Calendar is based on the cycles of the moon, and is constructed in a different fashion than the Western solar calendar3. In the Chinese calendar, the beginning of the year falls somewhere between late January and early February. The Chinese have adopted the Western calendar since 1911,but the lunar calendar is still used for festive occasions such as the Chinese New Year. Many Chinese calendars will print both the solar dates and the Chinese lunar dates.

  10. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  11. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

    2014-07-01

    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  12. Chinese Wrestling,Chinese Traditional Spirit to Be Succeeded

    Institute of Scientific and Technical Information of China (English)

    Sun Yongjian; Guo Yan

    2006-01-01

    @@ Chinese Wrestling which has been exercised for thousands of years, has a history of as long as the Chinese People China's Foreign Trade exclusively interviewed 72-year old Mr. Li Baoru, Head of the Chinese Wrestling Team. He stressed: "Chinese Wrestling will not become extinct, because it is the symbol of the Chinese Spirit."

  13. Qualitative analysis of the maintenance politics of the systems of a typical PWR by artificial neural networks; Analise qualitativa da politica de manutencoes dos sistemas de um PWR tipico por redes neurais artificiais

    Energy Technology Data Exchange (ETDEWEB)

    Lourenco, Victor Hugo Moreno

    2010-02-15

    Proceedings and techniques in order to maximize the reliability and the availability of industrial plants have been used along the last decades by specialists and professionals of maintenance. However, the modem industrial systems' sizing, and the increasing complexity and interdependence among its components have become this activity's planning a more and more difficult task. Considering this scenario, the objective of the present work is to provide a computational tool which is able to help about the taking decision's task, and about planning policies of maintenance practiced in thermonuclear plants. The tool developed is based on the artificial neural networks (ANN) for the recognition of standards and establishment of correlations among events occurred in the components of pressurized water reactor (PWR) typical systems. The ANN work as miners of database of failure events, and are able to identify connections and to establish imperceptible inferences even for the most experienced specialists in maintenance of nuclear systems. The results were attained from realistic data and are confronted against the maintenance's classic policies which are practiced nowadays on PWR thermonuclear plants. These results show the solidity of the technique in valuing and predicting failures in a real power plant, and is able to be used as a tool for supporting decisions about planning maintenance policies on a typical PWR. (author)

  14. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume III. Checkout applications. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Obenchain, C. F.; Ramsthaler, J. H.; Eales, E. P.; Charlton, T. R.; Childs, F. W.; Giles, M. M.; Good, E. G.; Gruen, G. E.; Guttman, J.; Johnsen, G. W.; Katsma, K. R.; Keeler, C. D.; Lawford, T. W.; Mohr, C. M.; Singer, G. L.; Townsend, W. C.

    1976-09-01

    Checkout problems presented include the following: PWR large cold leg break; PWR small cold leg break; PWR intermediate sized cold leg break; BWR large recirculation line break; BWR small recirculation line break; INEL Semiscale small cold leg break; INEL LOFT large cold leg break and INEL Semiscale large cold leg break. Also included is Update 2 of the RELAP 4/M0D5 code.

  15. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall.

  16. Comparison of Computational Estimations of Reactivity Margin From Fission Products and Minor Actinides in PWR Burnup Credit

    International Nuclear Information System (INIS)

    This paper has presented the results of a computational benchmark and independent calculations to verify the benchmark calculations for the estimation of the additional reactivity margin available from fission products and minor actinides in a PWR burnup credit storage/transport environment. The calculations were based on a generic 32 PWR-assembly cask. The differences between the independent calculations and the benchmark lie within 1% for the uniform axial burnup distribution, which is acceptable. The Δk for KENO - MCNP results are generally lower than the other Δk values, due to the fact that HELIOS performed the depletion part of the calculation for both the KENO and MCNP results. The differences between the independent calculations and the benchmark for the non-uniform axial burnup distribution were within 1.1%

  17. Application of RELAP5/MOD1 for calculation of safety and relief valve discharge piping hydrodynamic loads. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    A series of operability tests of spring-loaded safety valves was performed at Combustion Engineering in Windsor, CT as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of PWR Utilities in response to the recommendations of NUREG-0578 and the requirements of the NRC. Experimental data from five of the safety valve tests are compared with RELAP5/MOD1 calculations to evaluate the capability of the code to determine the fluid-induced transient loads on downstream piping. Comparisons between data and calculations are given for transients with discharge of steam, water, and water loop seal followed by steam. RELAP5/MOD1 provides useful engineering estimates of the fluid-induced piping loads for all cases.

  18. PFM Analysis for Pre-Existing Cracks on Alloy 182 Weld in PWR Primary Water Environment using Monte Carlo Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Phil; Bahn, Chi Bum [Pusan National University, Busan (Korea, Republic of)

    2015-10-15

    Probabilistic Fracture Mechanics (PFM) analysis was generally used to consider the scatter and uncertainty of parameters in complex phenomenon. Weld defects could be present in weld regions of Pressurized Water Reactors (PWRs), which cannot be considered by the typical fracture mechanics analysis. It is necessary to evaluate the effects of the pre-existing cracks in welds for the integrity of the welds. In this paper, PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out using a Monte Carlo simulation. PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out. It was shown that inspection decreases the gradient of the failure probability. And failure probability caused by the pre-existing cracks was stabilized after 15 years of operation time in this input condition.

  19. Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2015-09-01

    Full Text Available In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238 and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242. The main results were presented as a calculated-to-experimental ratio (C/E for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55. The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

  20. Coupled neutronics and thermal-hydraulic solution of a full-core PWR using VERA-CS

    International Nuclear Information System (INIS)

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating pressurized water reactors (PWRs) with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry. (author)

  1. Contamination of a green algae (Scenedesmus obliquus) from fresh water by radionuclides typical of PWR effluents: culture in a turbidostat

    International Nuclear Information System (INIS)

    The fixation of radioactive polluents, typical for PWR effluents, by the green soft water algae Scenedesmus obliquus is studied by means of a continuous culture method and in controlled conditions (turbidostat). Transfer factors are obtained. The elimination of radiocesium occurs in two distinct phases characterized respectively by a short biological period of less than one hour and a long period of the order of one day. The photosynthetic metabolism of the algae accounts for 25% of the decorporation. (Author)

  2. Numerical simulation of natural circulation phenomena in a PWR during TMLB' transients prior to core damage

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Y S; Domanus, H M; Liu, K V; Schmitt, R C; Sha, W T; Shah, V L

    1989-05-01

    Multidimensional thermal-hydraulic simulation of natural circulation phenomena during a TMLB' transient prior to core damage has been performed by using the COMMIX computer code. Both a three-loop and a four-loop PWR plant are analyzed. Effects of various parameters on natural circulation are investigated. The most important result is that natural circulation exists during a TMLB' transient for all the cases investigated so far. 190 figs., 34 tabs.

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

    International Nuclear Information System (INIS)

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

  4. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tiving, F.; Opel, S.

    2014-07-01

    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  5. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    OpenAIRE

    Andrej Prošek; Leon Cizelj

    2013-01-01

    Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pump...

  6. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  7. 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.

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

    International Nuclear Information System (INIS)

    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

  9. Break flow modeling for a steam generator tube rupture (SGTR) incident in a pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    The design-basis steam generator tube rupture (SGTR) scenario for the pressurized water reactor (PWR) postulates an instantaneous double-ended break of a steam generator (SG) U-tube. The flow rate through the broken U-tube depends on the primary-to-secondary side differential pressure in the affected SG, the primary coolant subcooling, and the break location along the U-tube. In this report, the RELAP5/MOD2 code's capability in predicting the SGTR break flow rate is assessed against experiments conducted on the Large Scale Test Facility (LSTF). The code is then used to predict break flow rate in the PWR for typical SGTR situations. It is shown that the code simulates well the break flow rates in the LSTF experiments for both single-phase and two-phase discharges, including two-phase critical flow discharge. The calculated PWR break flow rate takes a maximum for a break occurring at the lower end of the U-tube, on its cold leg side, because of the combined influence of tube-inlet fluid subcooling and frictional pressure drop along the broken tube. Modeling the tube frictional pressure drop is important to predict the break flow rate dependence on inlet fluid sub-cooling; simplified break flow modeling which applies a constant discharge coefficient less than unity, instead of modeling explicitly the tube frictional length, fails to predict the change in break flow rate accurately if the inlet subcooling varies for a wide range. (author)

  10. Experimental investigation of reflux condensation heat transfer in PWR steam generator tubes in the presence of noncondensible gases

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen; Wu, Tiejun [Purdue Univ., West Lafayette (United States); Nagae, Takashi [Institute of Nuclear Safety System, Tokyo (Japan)

    2003-07-01

    Under certain circumstances in a Pressurized Water Reactor (PWR), the coolant system may be in a partially drained state and reflux condensation in the steam generator U-tubes can be the major heat removal mechanism. Noncondensable gases may be present and would degrade the heat transfer rate. If heat removal rates are insufficient, this situation could lead to core boil-off, fuel rod heatup, and eventually core damage. The Institute of Nuclear Safety System, Inc. (INSS) and the Nuclear Heat Transfer Systems Laboratory at Purdue University have begun a cooperative research program to investigate the effectiveness of reflux condensation in PWR steam generator U-tubes in the presence of noncondensable gases. The final objectives are to provide local heat transfer data for development of methods to analyze reflux condensation in PWR steam generator U-tubes and to investigate the potential for flooding. Key features of the experimental data reported herein are that they are local data under laminar steam/gas mixture and condensate film flow and they are taken from a test section with dimensions similar to an actual steam generator tube. Steady state data were obtained under various steam and air inlet flow rates and pressures. The data show the significant degrading effect of noncondensable gas on heat transfer coefficients. From the data, correlations for the reflux condensation local heat transfer coefficient and the local Nusselt number under laminar conditions were derived. These experiments are providing essential and unique fundamental data for development of methods to analyze reflux condensation.

  11. Evaluation of stress corrosion cracking of irradiated 304L stainless steel in PWR environment using heavy ion irradiation

    Science.gov (United States)

    Gupta, J.; Hure, J.; Tanguy, B.; Laffont, L.; Lafont, M.-C.; Andrieu, E.

    2016-08-01

    IASCC has been a major concern regarding the structural and functional integrity of core internals of PWR's, especially baffle-to-former bolts. Despite numerous studies over the past few decades, additional evaluation of the parameters influencing IASCC is still needed for an accurate understanding and modeling of this phenomenon. In this study, Fe irradiation at 450 °C was used to study the cracking susceptibility of 304 L austenitic stainless steel. After 10 MeV Fe irradiation to 5 dpa, irradiation-induced damage in the microstructure was characterized and quantified along with nano-hardness measurements. After 4% plastic strain in a PWR environment, quantitative information on the degree of strain localization, as determined by slip-line spacing, was obtained using SEM. Fe-irradiated material strained to 4% in a PWR environment exhibited crack initiation sites that were similar to those that occur in neutron- and proton-irradiated materials, which suggests that Fe irradiation may be a representative means for studying IASCC susceptibility. Fe-irradiated material subjected to 4% plastic strain in an inert argon environment did not exhibit any cracking, which suggests that localized deformation is not in itself sufficient for initiating cracking for the irradiation conditions used in this study.

  12. The coupling of the Star-Cd software to a whole-core neutron transport code Decart for PWR applications

    International Nuclear Information System (INIS)

    As part of a U.S.- Korea collaborative U.S. Department of Energy INERI project, a comprehensive high-fidelity reactor-core modeling capability is being developed for detailed analysis of existing and advanced PWR reactor designs. An essential element of the project has been the development of an interface between the computational fluid dynamics (CFD) module, STAR-CD, and the neutronics module, DeCART. Since the computational mesh for CFD and neutronics calculations are generally different, the capability to average and decompose data on these different meshes has been an important part of code coupling activities. An averaging process has been developed to extract neutronics zone temperatures in the fuel and coolant and to generate appropriate multi group cross sections and densities. Similar procedures have also been established to map the power distribution from the neutronics zones to the mesh structure used in the CFD module. Since MPI is used as the parallel model in STAR-CD and conflicts arise during initiation of a second level of MPI, the interface developed here is based on using TCP/IP protocol sockets to establish communication between the CFD and neutronics modules. Preliminary coupled calculations have been performed for PWR fuel assembly size problems and converged solutions have been achieved for a series of steady-state problems ranging from a single pin to a 1/8 model of a 17 x 17 PWR fuel assembly. (authors)

  13. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  14. Organ-specific gene expression in maize: The P-wr allele. Final report, August 15, 1993--August 14, 1996

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, T.A.

    1997-06-01

    The ultimate aim of our work is to understand how a regulatory gene produces a specific pattern of gene expression during plant development. Our model is the P-wr gene of maize, which produces a distinctive pattern of pigmentation of maize floral organs. We are investigating this system using a combination of classical genetic and molecular approaches. Mechanisms of organ-specific gene expression are a subject of intense research interest, as it is the operation of these mechanisms during eukaryotic development which determine the characteristics of each organism Allele-specific expression has been characterized in only a few other plant genes. In maize, organ-specific pigmentation regulated by the R, B, and Pl genes is achieved by differential transcription of functionally conserved protein coding sequences. Our studies point to a strikingly different mechanism of organ-specific gene expression, involving post-transcriptional regulation of the regulatory P gene. The novel pigmentation pattern of the P-wr allele is associated with differences in the encoded protein. Furthermore, the P-wr gene itself is present as a unique tandemly amplified structure, which may affect its transcriptional regulation.

  15. Qualitative analysis of the maintenance politics of the systems of a typical PWR by artificial neural networks

    International Nuclear Information System (INIS)

    Proceedings and techniques in order to maximize the reliability and the availability of industrial plants have been used along the last decades by specialists and professionals of maintenance. However, the modem industrial systems' sizing, and the increasing complexity and interdependence among its components have become this activity's planning a more and more difficult task. Considering this scenario, the objective of the present work is to provide a computational tool which is able to help about the taking decision's task, and about planning policies of maintenance practiced in thermonuclear plants. The tool developed is based on the artificial neural networks (ANN) for the recognition of standards and establishment of correlations among events occurred in the components of pressurized water reactor (PWR) typical systems. The ANN work as miners of database of failure events, and are able to identify connections and to establish imperceptible inferences even for the most experienced specialists in maintenance of nuclear systems. The results were attained from realistic data and are confronted against the maintenance's classic policies which are practiced nowadays on PWR thermonuclear plants. These results show the solidity of the technique in valuing and predicting failures in a real power plant, and is able to be used as a tool for supporting decisions about planning maintenance policies on a typical PWR. (author)

  16. Comparison of computational performance of GA and PSO optimization techniques when designing similar systems - Typical PWR core case

    International Nuclear Information System (INIS)

    Research highlights: → Performance of PSO and GA techniques applied to similar system design. → This work uses ANGRA1 (two loop PWR) core as a prototype. → Results indicate that PSO technique is more adequate than GA to solve this kind of problem. - Abstract: This paper compares the performance of two optimization techniques, particle swarm optimization (PSO) and genetic algorithm (GA) applied to the design a typical reduced scale two loop Pressurized Water Reactor (PWR) core, at full power in single phase forced circulation flow. This comparison aims at analyzing the performance in reaching the global optimum, considering that both heuristics are based on population search methods, that is, methods whose population (candidate solution set) evolve from one generation to the next using a combination of deterministic and probabilistic rules. The simulated PWR, similar to ANGRA 1 power plant, was used as a case example to compare the performance of PSO and GA. Results from simulations indicated that PSO is more adequate to solve this kind of problem.

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

    International Nuclear Information System (INIS)

    The increase of occupational radiation exposure (ORE) due to the increase of the operational period at existing nuclear power plant and also the publication of the new version of ICRP recommendation (ICRP publication No. 60) for radiological protection require much more strict reduction of radiation buildup in the nuclear power plant. The major sources of the radiation, i.e. the radioactive corrosion-products, are generated by the neutron activation of the corrosion products at the reactor core, and then the radioactive corrosion products are transported to the outside of the core, and accumulated near the steam generator side at PWR. Major radioactive corrosion-products of interest in PWR are Cr51,: Mn54,: Co58,: Fe59 and Co60. Among them Co58 and Co60 are known to contribute approximately more than 70% of the total ORE. Thus our main concerns are focused on predicting the transport and deposition of the Co radionuclides and suggesting the optimizing method which can minimize and control the ORE of the nuclear power plant. It is well known that Co-source is most effectively controlled by pH-solubility radiation control, and also some complex computer codes such as CORA and PACTOLE have been developed and revised to predict the corrosion product behavior. However these codes still imply some intrisic problems in simulating the real behavior of corrosion products in the reactor because of 1) the lack of important experimental data, coefficients and parameters of the transport and reactions under actual high temperature and pressure conditions, 2) no general theoretical modelling which can describe such many different mechanisms involved in the corrosion product movements, 3) the newly developed and measured behavior of the corrosion product transport mechanism. Since no sufficient and detailed information is available from the above-mentioned codes (also due to propriority problems), we concentrate on developing a new computer code, CP-TRAN (Corrosion Product

  18. Characterization of PWR vessel steel tearing under severe accident condition temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Matheron, Philippe, E-mail: philippe.matheron@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Chapuliot, Stephane, E-mail: stephane.chapuliot@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Nicolas, Laetitia, E-mail: laetitia.nicolas@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Laboratoire de Mecanique des Structures Industrielles Durables, UMR CNRS-EDF 2832, 1 avenue du General de Gaulle, F-92141 Clamart (France); Koundy, Vincent, E-mail: vincent.koundy@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France); Caroli, Cataldo, E-mail: cataldo.caroli@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer We characterized French PWR vessel steel tearing resistance at high temperatures. Black-Right-Pointing-Pointer Tearing tests on Compact Tension (CT) specimens were carried out. Black-Right-Pointing-Pointer The variability of tearing properties with PWR vessels specifications was studied. Black-Right-Pointing-Pointer We propose a tearing criterion (energy parameter Gfr) at high temperatures. - Abstract: In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment. On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance. The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens. The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead

  19. CHINESE OF HUMANITY

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    A discussion of chinese curriculum of primary school under the background of new curriculum reform Mao xinjuan Feng haiying [Abstract] in recent years, Chinese learning received more and more attention by people article mainly from the national studies this course concepts, the curriculum reform of elementary school curriculum requirements and how to effective implementation of primary national studies course several aspects under the background of curriculum reform of Chinese primary curriculum the new school

  20. Psychiatry and Chinese Culture

    OpenAIRE

    Lin, Tsung-Yi

    1983-01-01

    When we examine the cultural characteristics that influence mental disorders and related behavior among the Chinese, no major differences are found between Chinese and other groups in the range of disorders or in overall prevalence. Several cultural factors influence the recognition and treatment of mental illness, among which are attitudes toward emotional display, somatic as opposed to psychogenic disorders and features of the traditional medical belief system in Chinese culture. The Chines...

  1. In-situ chemical decontamination of a PWR primary loop large components with the MEDOC process

    International Nuclear Information System (INIS)

    Shutdown in 1987 after 25 years of operation, the BR3-PWR was selected in 1989 as one of the four pilot decommissioning projects by the European Commission, in the framework of its five-year plan of Research and Technological Development on decommissioning of nuclear installations. The dismantling of a PWR type reactor leads to the production of large masses of contaminated metallic pieces, including structural materials, primary piping, tanks and heat exchangers. One of our main objectives is to demonstrate that we can minimise the volume of radioactive waste in an economical way by privileging alternative material routes, such as the clearance of materials after thorough decontamination. Therefore the SCK-CEN has developed its own chemical decontamination process, so-called MEDOC (MEtal Decontamination by Oxidation with Cerium), based on the use of Ce(IV) as strong oxidising species in sulphuric acid, which is continuously regenerated by ozone injection at high temperature. The industrial installation, which was designed and constructed in close collaboration with Framatome-ANP (France), started operation in September 1999. Initially designed to decontaminate stainless steel pieces, the process has been easily upgraded to allow the treatment of carbon steel using simply H2SO4. Up to now, more than 30 tons of contaminated materials, including primary pipes and primary pumps housings, have been treated batch wise with success. 69% of material can be directly cleared after treatment (Activity lower than 0.1 Bq/g in 60Co) 27% will be free released after melting (activity lower than 1 Bq/g) and less than 4% have to undergo an additional physical decontamination step prior to clearance. However, the working of our process is not restricted to batch wise operations. Thanks to minor adaptations on the existing plant, the SCK-CEN has recently performed the closed loop decontamination of the large components of the primary loop, namely the steam generator and the

  2. Effects of dissolved hydrogen content in PWR primary water and PWSCC initiation property

    International Nuclear Information System (INIS)

    Environmental mitigation is being looked forward to as one of the promising preventive measures for better plant maintenance. Especially, it would be much more effective, when environmental measures are introduced in combination with mechanical measures, such as stress improvement, material replacement, and so on. Dissolved hydrogen (DH) concentration in the primary cooling system has been controlled at the level of 25-35 ml STP/kg H2O in all PWR plants in Japan, maintaining a sufficient margin to the utilities' self-regulated band 15-50 ml STP/kg H2O. The lower limit of this self-regulated band was determined to suppress reactor coolant radiolysis based on experiments using a test reactor over 50 years ago, though the test was performed at much lower than typical operational temperature and is considered over-conservative at present. On the other hand, primary water stress corrosion cracking (PWSCC) events have occurred on Dissimilar Metal Welds (DMW) at vessel nozzles or at vessel head penetration in some PWR plants in Japan in recent years. PWSCC growth test data have been showing that the current management band of DH content is around the peak of crack growth rate (CGR) at about 340degC, corresponding to the operating temperature of the pressurizer. To avoid the higher CGR levels, the DH content band should be shifted to the higher side, or to the lower side. The authors suggest much lower DH content than that at present as an alternative control, since some experimental data on PWSCC initiation show monotonically decreasing dependency of initiation time with increasing DH content. In this paper, we show the effects of DH content on PWSCC initiation property. Reverse-u-bend (RUB) tests and three-point-bend (3PB) tests were conducted for Alloy 600 base metal and Alloy 182, respectively, to examine the dependency of the lower DH content than currently controlled band on PWSCC initiation time. Test results showed that advantage of lowering DH content down to

  3. Analysis of PWR ex-vessel steam explosion for axial and side melt release

    International Nuclear Information System (INIS)

    Highlights: • PWR ex-vessel steam explosion study was performed with the MC3D code. • The conditions of the OECD project SERENA reactor exercise were considered. • The influence of jet breakup modelling on the fuel coolant interaction was analysed. • The pressure loads on the cavity wall significantly depend on the release location. • For reactor simulations it is important that FCI codes have 3-D capabilities. - Abstract: A steam explosion, in the frame of nuclear reactor safety, is a process resulting from the interaction between the core melt and water. Steam explosions are an important nuclear safety issue because they can potentially jeopardize the primary system and the containment integrity of the nuclear power plant. In the year 2007 the OECD project SERENA was launched, trying to resolve the open issues in steam explosion understanding and modelling, with the focus on reactor applications. To verify the progress made in the understanding and modelling of the fuel coolant interaction key phenomena for reactor applications a reactor exercise has been performed at the end of the project. In the paper the PWR ex-vessel steam explosion study, which was carried out with the MC3D code in conditions of the SERENA reactor exercise, is presented and discussed. The axial melt release case was simulated in 2-D and the side release case in 3-D. In reactor calculations the largest uncertainties in the prediction of the steam explosion strength may be expected due to the large uncertainties related to the jet breakup. Therefore the premixing simulations were performed with both available jet breakup models, i.e. a global and a local one, to get some insight in these uncertainties. For each premixing simulation an explosion simulation was performed, triggering the explosion around the time of the melt bottom contact. It turned out that despite the different jet breakup in both considered approaches, the calculated pressure loads are in the same range due to

  4. Analysis of PWR ex-vessel steam explosion for axial and side melt release

    Energy Technology Data Exchange (ETDEWEB)

    Leskovar, Matjaž, E-mail: matjaz.leskovar@ijs.si; Uršič, Mitja

    2015-03-15

    Highlights: • PWR ex-vessel steam explosion study was performed with the MC3D code. • The conditions of the OECD project SERENA reactor exercise were considered. • The influence of jet breakup modelling on the fuel coolant interaction was analysed. • The pressure loads on the cavity wall significantly depend on the release location. • For reactor simulations it is important that FCI codes have 3-D capabilities. - Abstract: A steam explosion, in the frame of nuclear reactor safety, is a process resulting from the interaction between the core melt and water. Steam explosions are an important nuclear safety issue because they can potentially jeopardize the primary system and the containment integrity of the nuclear power plant. In the year 2007 the OECD project SERENA was launched, trying to resolve the open issues in steam explosion understanding and modelling, with the focus on reactor applications. To verify the progress made in the understanding and modelling of the fuel coolant interaction key phenomena for reactor applications a reactor exercise has been performed at the end of the project. In the paper the PWR ex-vessel steam explosion study, which was carried out with the MC3D code in conditions of the SERENA reactor exercise, is presented and discussed. The axial melt release case was simulated in 2-D and the side release case in 3-D. In reactor calculations the largest uncertainties in the prediction of the steam explosion strength may be expected due to the large uncertainties related to the jet breakup. Therefore the premixing simulations were performed with both available jet breakup models, i.e. a global and a local one, to get some insight in these uncertainties. For each premixing simulation an explosion simulation was performed, triggering the explosion around the time of the melt bottom contact. It turned out that despite the different jet breakup in both considered approaches, the calculated pressure loads are in the same range due to

  5. Corrosion fatigue initiation behaviour of wrought austenitic stainless pipe steels under simulated BWR/HWC and PWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Leber, H.J.; Ritter, S.; Seifert, H.P [Paul Scherrer Institute, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen PSI (Switzerland)

    2011-07-01

    The corrosion fatigue (CF) initiation and short crack growth behavior of different low-carbon and stabilized austenitic stainless steels was characterized under simulated BWR and primary PWR conditions by cyclic fatigue tests with sharply notched fracture mechanics specimens in the temperature range from 70 to 320 C. Environmental reduction of fatigue initiation life was observed in all stainless steels at strain rates {<=} 0.1 %/s in BWR and PWR environment. The stationary short crack CF crack growth rates after crack advances of 50 to 300 {mu}m from the notch-root were in the typical range of corresponding results from tests with long cracks (pre-cracked specimens) and also showed the same system parameter response. The effect of environment on the initiation process ({Delta}a = 10 {mu}m) was relevantly stronger than on the subsequent stationary short crack growth. Both, under BWR/HWC and PWR conditions, a relevant environmental reduction of fatigue initiation life occurred for the combination of temperatures {>=} 100 C, notch strain rates {<=} 0.1 %/s and notch strain amplitudes {>=} 0.3 %. If these conjoint threshold conditions were simultaneously satisfied, the environmental enhancement increased with decreasing strain rate and increasing temperature. Material and water chemistry parameters usually only had a little effect. Sensitization affected the CF behavior under highly oxidizing BWR/NWC conditions only. Preliminary block loading experiments did not reveal significant static load hold period effects on the technical corrosion fatigue initiation life. If the critical requirements were satisfied, the BWR/HWC and PWR environments usually resulted in acceleration of short fatigue crack growth by a factor of 5 to 20 with respect to air. Solution annealed steels showed slightly shorter CF initiation lives, but also lower stationary short CF crack growth rates under BWR/HWC and PWR conditions with low ECPs than under highly oxidizing BWR/NWC conditions. A very

  6. History of Chinese medicinal wine.

    Science.gov (United States)

    Xia, Xun-Li

    2013-07-01

    Chinese medicinal wine is one type of a favorable food-drug product invented by Chinese ancestors for treating and preventing diseases, promoting people's health and corporeity, and enriching people's restorative culture. In the course of development of the millenary-old Chinese civilization, Chinese medicinal wine has made incessant progress and evolution. In different historical periods, Chinese medicinal wine presented different characteristics in basic wine medical applications, prescriptions, etc. There are many medical and Materia Medica monographs which have systemically and specifically reported on Chinese medicinal wine in past Chinese dynasties. By studying leading medical documents, this article made an outline review on the invention, development, and characteristics of Chinese medicinal wine.

  7. 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.

  8. Effect of ethanol amine injection on flow-accelerated corrosion of PWR secondary systems

    International Nuclear Information System (INIS)

    Some pressurized water reactor (PWR) plants have switched secondary system feed water treatment to ethanol-amine (ETA) injection from all-volatile treatment (AVT) to reduce iron transfer in the steam generator (SG). However, the effect of ETA injection on FAC rate has not been studied systematically. To assess the effect of ETA injection on FAC rate, the water chemistries in secondary systems were calculated by considering the thermal decomposition of hydrazine in SG and the vapor/liquid partition of ammonia, ETA, and hydrazine in SG and in a moisture-separator-and-reheater (MSR). Then, we measured the FAC rate experimentally by rotation tests to examine the effect of ETA injection. The high pH condition of ETA injection reduced the FAC rate more than the low pH condition of AVT. No chemical effect on the FAC rate was observed between ETA injection and AVT at 180degC. We also evaluated the FAC rate using magnetite solubility with and without ETA injection. The evaluation showed that ETA injection reduces the FAC rate of the secondary system. (author)

  9. Ballooning analysis for the Sizewell B PWR using symmetric MABEL calculations

    International Nuclear Information System (INIS)

    An analysis of the fuel clad ballooning potential associated with the Sizewell B PWR following a design basis large break cold leg LOCA is described. Calculations employ MABEL-2C code. No allowance has been made for asymmetries in power or geometry, thus precluding any amelioration offered by early clad rupture. Thermal hydraulic data were derived from a TRAC-PD2 best estimate analysis of the LOCA and the work includes a detailed sensitivity study which leads to a correlation between peak clad temperature and clad strain. For the best estimate start of cycle 1 peak rod rating, no loss of coolability is expected within 95 percent confidence limits on peak clad temperature. No loss of coolability is expected either for rods at the design basis peak rod rating. The temperature does not have to be much higher than the 95 percent confidence limit on the best estimate rating or much beyond that of the design basis rating for rod contact and severe blockage to follow. This indicates that to establish a complete safety case the added complexity of pellet eccentricity and rod to rod power variations must be considered. (U.K.)

  10. In-service inspection techniques for PWR steam generator feedwater and pressuriser nozzles

    International Nuclear Information System (INIS)

    Regular ultrasonic inspection of the steam generator feedwater and pressuriser nozzles of the Sizewell B Pressurised Water Reactor, under construction by the Central Electricity Generating Board (CEGB), will be carried out to detect and size any service-induced cracking in the nozzle corners and bores. External access only will be available for such inspections and, to achieve full inspection coverage, it may be necessary to scan probes under the outer blend radius and adjacent surfaces of each nozzle. As part of the PWR Safety Research Programme being conducted collaboratively by CEGB and the United Kingdom Atomic Energy Authority (UKAEA) in the UK, Risley Laboratories have been developing automated ultrasonic techniques to meet the stringent inspection standards demanded for these components. The geometrical complexity of the inspection has necessitated the use of mathematical modelling to optimise inspection techniques and coverage. Computer-based data collection, display and analysis methods have been developed for the combined pulse-echo and time-of-flight diffraction techniques selected for these inspections. In this paper, the development work performed at Risley is reviewed and examples of the application of the inspection techniques to simulated service-induced defects in full-scale test specimens presented. Work on the related problem of cracking in BWR nozzles (ref.1) has demonstrated detection capability but not accurate sizing. However sizing of defects in the nozzle corner areas has been achieved in this programme with the use of advanced reconstruction methods. (author)

  11. Design and construction planning for the U.K. standard PWR design

    International Nuclear Information System (INIS)

    Construction of the U.K.'s first pressurized water reactor, Sizewell B, started in June 1987, and the Central Electricity Generating Board (CEGB) plans to build a series of five or six stations that will replicate the Sizewell B design. The main features of the CEGB strategy are described. The construction planning is started at a very early stage of the project. For Sizewell B some of the key elements are as follows: (1) construction methods and sequences developed for the standardized nuclear unit power plant system (SNUPPS) projects and from Bechtel PWR expertise, modified only where necessary to suit the Sizewell B layout, were adopted. (2) A construction group was involved full-time during the design phase. (3) A 1:75 scale construction model of the principal building of the power block was used to confirm the major building sequences, lifting methods, and positioning of construction access openings and routes. (4) System design clearance occurred sufficiently early to ensure that plant delivery, civil interface, and civil construction design satisfy the construction sequences and timing. (5) Overseas activities were closely observed, in particular the plants at Wolf Creek and Callaway, which were constructed using Bechtel-SNUPPS sequences and methods

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

    International Nuclear Information System (INIS)

    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)

  13. Development of a parametric containment event tree model of a severe PWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1996-06-01

    The study supports the development project of STUK on `Living` PSA Level 2. The main work objective is to develop review tools for the Level 2 PSA studies underway at the utilities. The SPSA (STUK PSA) code is specifically designed for the purpose. In this work, SPSA is utilized as the Level 2 programming and calculation tool. A containment event tree (CET) model is built for analysis of severe accidents at the Loviisa pressurized water reactor (PWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to include new research results, and so it facilitates the Living PSA concept on Level 2 as well. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting at a low primary system pressure. Severe accident progression from five plant damage states (PDSs) is examined, however the integration with Level 1 is deferred to more definitive, integrated, safety assessments. (34 refs., 5 figs., 9 tabs.).

  14. Ductile crack growth resistance of PWR components. Application for structural integrity assessment

    International Nuclear Information System (INIS)

    Structural integrity assessment of PWR components, as pressure vessel and piping, needs to evaluate the ductile crack growth resistance which is generally characterized by J resistance curves (or J-R curves) based on the path-independent J Integral. These curves are more often obtained from laboratory tests with small specimens as CT-specimens and their application to large component safety analysis could be questionable Indeed, it is well known that J-R curves could depend on the specimen size and on the loading mode (i.e. bending stress versus tensile stress) but this dependency could be different from one material to another. This means that it would depend not only on the stress-strain state but also on the actual local fracture mechanisms (i. e. the damage) occurring before the crack initiation or during the crack propagation. The purpose of this paper is to gather some results of crack growth resistance measurement studied at EDF with different materials in order to show how the effect of the parameters, as specimen geometry and mode of loading, is directly related to the local fracture mechanisms or the microstructure of the materials. For that a number of results are analysed by means of the local approach of fracture which is a very useful tool to predict quantitatively the J-R curve dependency, related to fracture mechanisms (authors). 12 refs., 9 figs

  15. Instrumentation and control system upgrade plan for operating PWR plants in Japan

    International Nuclear Information System (INIS)

    Digital technology has been applied to all non-safety grade instrumentation and control (I ampersand C) systems in the latest Japanese PWR plants, and has achieved more reliable and operable systems, easier maintenance and cable reductions. In the next stage APWR plants, the digital technology will be also applied to all the I ampersand C systems including safety grade systems. Parallel to the above efforts, many backfitting programs in which the digital technology is applied to operating plants are under way to improve reliability and operability. The backfitting programs for operating plants are proceeded in two phases, synthesizing various utility's needs to improve plant availability and operability, improvement of digital technology, and complexity of the practicable replacement procedures. Phase 1 is a partial application of digital technology, while Phase 2 is a complete application of digital technology. Phase 1 has been implemented in a number of operation plants, while Phase 2 studies are in the design stage, but have not been implemented at this point. This paper presents examples of the partial application of digital technology to operating plants, and the contents of basic design for the complete application of digital technology

  16. Improvements in the modeling of crevice chemistry in PWR steam generators using the MULTEQ code

    International Nuclear Information System (INIS)

    The MULTEQ Code was developed to model the high temperature chemistry of the concentrated solutions which form in PWR Steam Generator (SG) crevices and flow restricted regions. Earlier versions of the code modeled the evaporation process as either a constant mass or constant volume system with a well-mixed liquid solution. Experiments performed in simulated heated crevices and more complex thermal hydraulic models suggest that the solution is not well-mixed. The degree of mixing strongly effects the behavior of both volatile and insoluble species in the crevice. The well-mixed models predict much lower concentrations of volatile species in the crevice than is observed in laboratory experiments and inferred from corrosion in the field. By removing both the steam and precipitates formed in a step-wise fashion, the constant mass model approximates the behavior in a solution which is not well-mixed. The basis for this new model is fully described in the paper. 9 refs., 4 figs

  17. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  18. Test requirements for the integral effect test to simulate Korean PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  19. VOF Calculations of Countercurrent Gas-Liquid Flow in a PWR Hot Leg

    Directory of Open Access Journals (Sweden)

    M. Murase

    2012-01-01

    Full Text Available We improved the computational grid and schemes in the VOF (volume of fluid method with the standard − turbulent model in our previous study to evaluate CCFL (countercurrent flow limitation characteristics in a full-scale PWR hot leg (750 mm diameter, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa. In this paper, therefore, to evaluate applicability of the VOF method to different fluid properties and a different scale, we did numerical simulations for full-scale air-water conditions and the 1/15-scale air-water tests (50 mm diameter, respectively. The results calculated for full-scale conditions agreed well with CCFL data and showed that CCFL characteristics in the Wallis diagram were mitigated under 1.5 MPa steam-water conditions comparing with air-water flows. However, the results calculated for the 1/15-scale air-water tests greatly underestimated the falling water flow rates in calculations with the standard − turbulent model, but agreed well with the CCFL data in calculations with a laminar flow model. This indicated that suitable calculation models and conditions should be selected to get good agreement with data for each scale.

  20. Restructuring of current paradigm for coolant hydrogen control in PWR primary system

    International Nuclear Information System (INIS)

    In PWR primary system, hydrogen is injected into coolant to prevent stress corrosion cracking (SCC) of stainless steel, which is one of main structural materials of the system, and attention must be paid to maintain its concentration within its target range. The SCC of stainless steel attributed to water quality has never occurred before now. However, corrosion of the materials and increase of radiation source using higher hydrogen have been observed in actual plants. Then, here were discussed on importance and feasibility on restructuring of current paradigm for coolant hydrogen control in the system to establish optimal environment for integrity of stainless steel, nickel base alloy, and fuel cladding regarding the introduction of higher burn-up fuel and extended operational cycle as well as plant aging. As a result, it was found that although optimum dissolved hydrogen level in the primary coolant was not clarified them, some recent knowledges could be suggested on importance and feasibility of restructuring current paradigm for the coolant hydrogen control from viewpoints of Inconel corrosion and fuel performance and dose reduction. (G.K.)

  1. Research on Operation and Control Strategy of 600MW PWR in Load Follow

    International Nuclear Information System (INIS)

    600MW Pressurized Water Reactor (PWR) is designed to operate in Constant Axial Offset Control (CAOC) strategy with base load originally. By calculations over a typical load follow scenario '12-3-6-3 (100-50-100%FP) via the CASMO-4E and SIMULATE-3 package, values of core operating parameter have been examined. With the progress of the nuclear power industry, advanced reactors are considered to have a good performance in load follow, economy and flexibility. Under the premise of fuel loading and structural dimensions unchanged, two independent control rod groups M and AO are used in 600MW pressurized water reactor to provide fine control of both the core reactivity and axial power distribution, which is named ' Improved G strategy .' The influences of different control rod distributions, composition materials, and overlap steps had in power changes have been examined in a comparative study to choose the optimal one.Then we simulate a range of load follow scenarios of the redesigned 600MW core without adjusting soluble boron concentration in the begin, middle and end of first cycle. This paper additionally demonstrated the moderator temperature coefficient and shutdown margin values of the reactor in Improved G strategy to compare with the thermal safety design criteria. It's demonstrated that adequate adjustment of control rod groups enable the core to perform load follow through Improved G strategy in 80% of cycle and save a large volume of liquid effluent particularly toward the end of cycle

  2. CFD Turbulence Study of PWR Spacer-Grids in a Rod Bundle

    Directory of Open Access Journals (Sweden)

    C. Peña-Monferrer

    2014-01-01

    the flow dynamics and heat transfer phenomena along the fuel rods. This work presents the analysis of the turbulence effects of a split-type and swirl-type spacer-grid geometries on single phase in a PWR (pressurized water reactor rod bundle. Various computational fluid dynamics (CFD calculations have been performed and the results validated with the experiments of the OECD/NEA-KAERI rod bundle CFD blind benchmark exercise on turbulent mixing in a rod bundle with spacers at the MATiS-H facility. Simulation of turbulent phenomena downstream of the spacer-grid presents high complexity issues; a wide range of length scales are present in the domain increasing the difficulty of defining in detail the transient nature of turbulent flow with ordinary turbulence models. This paper contains a complete description of the procedure to obtain a validated CFD model for the simulation of the spacer-grids. Calculations were performed with the commercial code ANSYS CFX using large eddy simulation (LES turbulence model and the CFD modeling procedure validated by comparison with measurements to determine their suitability in the prediction of the turbulence phenomena.

  3. ICONEL 690: a material with improved corrosion resistance for PWR steam generator tubes

    International Nuclear Information System (INIS)

    This paper reviews the major aspects of the joint program carried out by the Commissariat a l'Energie Atomique, Electricite de France, Framatome and Westinghouse on the investigation of Inconel 690 (I.690) for use as PWR Steam Generator Tubing. The program was conceived as a further step in the long-term development of improvements in S.G. tube material corrosion resistance. Although the major emphasis of the work was on the corrosion resistance it was also necessary, in preparing for the commercial use of I.690 for S.G. tubing, to verify other aspects of behaviour in respect of physical and mechanical properties and for steam generator fabrication purposes. A key activity in the preparation for commercial use was the pre-production fabrication by a tube supplier of about 200 full-length tubes from each of three different heats. These tubes not only served the need to fully characterize the tubing (a requirement of the French specifications) but also provided a source of additional fully representative material for confirmatory corrosion and fabrication evaluation

  4. Level 3 PSA and it's implementation for PWR accident

    International Nuclear Information System (INIS)

    Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA) is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidents of PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The series of calculations performed are: calculate the source terms of the core damaged, modeling of meteorological conditions and environmental site, exposure pathway modeling, analysis of radionuclide dispersion and transport phenomena in the environment, radionuclide deposition analysis, analysis of radiation dose, protection & mitigation analysis, and risk analysis. The assessment uses a series of subsystems on PC Cosyma software. The results prove that the safety assessment using Level 3 PSA methodology is very effective and comprehensive estimate the impact, consequences, risks, nuclear emergency preparedness, and the reactor accident management especially for severe accidents or beyond design basis accidents of nuclear power plants. The results of the assessment can be used as a feedback to safety assessment of Level 1 PSA and Level 2 PSA. (author)

  5. Investigation of proportional integral based technique for controlling PWR pressurizer water level

    International Nuclear Information System (INIS)

    The control system in the pressurizer water level is necessary for the safety of the operation of pressurizer water reactors (PWRs). It will compensate t the primary loop volume changes while keeping the existing pressure of the primary loop at a certain set point. Some researchers have proposed both an intelligent system of neural network and a fuzzy logic to improve the capability of the common conventional control systems used in PWR, i.e. Proportional-Integral (PI) or Proportional-Integral-Derivative (PID). However, those studies did not comprehensively assess the potential of the conventional control systems. It has been confirmed that if the parameters of the Pi based control system are determined more carefully, its results will be equivalent to the results of other control systems or even better. This study aims to address this challenging topic by examining and testing control parameters more closely to obtain the best configuration of the PI-based control system. Compared to the results of the artificial neural network-based control system, the PI results of this study provide an increase of rise time around 280 times, better settling time for approximately 293 times, a decrease of overshoot about 1.1 times, and a reduction of the peak around 0.2%. The configuration has also been validated to be stable and able to overcome disturbances for about 10 seconds with a maximum peak level of 0.005%. Moreover, it can track the set point changes very well. (author)

  6. Characterization of SCC crack tip in alloy 600 in simulated primary water of PWR

    International Nuclear Information System (INIS)

    High-resolution observations on microstructure of SCC crack tips were carried out on alloy 600 exposed to a simulated PWR primary water environment. A focused-ion beam (FIB) micro-processing has been applied to prepare electron transparent foils for cross-sectional observations of the crack tips. Cr rich oxides ((Cr,Fe)3O4 and Cr2O3) and a metallic Ni phase were identified at the crack tip and along the grain boundary ahead of the crack tip. Oxide layers on the crack walls were identified as a double layer structure. The inner layer consisted of Cr rich oxides and the outer layer consisted of NiO. Similar microstructure was observed in cross sections of surface oxide layers that were prepared by using the same FIB technique. The inner layer consisted of Cr rich oxides and a metallic Ni phase and was covered by the outer layer of NiFe2O4 and NiO. In addition, Cr rich oxides were observed along uncracked grain boundaries intercepted by the surface. It is suggested that selective oxidation of chromium occurs at the crack tips, grain boundaries and inner layers. The observations of oxygen penetration along the grain boundaries support a larger diffusivity of oxygen in the stressed grain boundary than that in the unstressed one. (author)

  7. Studies of residual stress measurement and analysis techniques for a PWR dissimilar weld joint

    International Nuclear Information System (INIS)

    For evaluation of the PWSCC crack propagation behavior, a test model was produced using the same fabrication process of Japanese PWR plants and the stress distribution change was measured during a fabrication process such as a hydrostatic test, welding a main coolant pipe to the stainless steel safe end and an operation condition test. For confirmation of validity of the numerical estimation method of the stress distribution, FE analysis was performed to calculate the stress distributions for each fabrication process. From the validation procedure, a standard residual stress evaluation method was established. Furthermore for consideration of characteristics of PWSCC’s propagation behavior of the dissimilar welding joint of the safe end nozzles, the influence coefficients at the deepest point for the stress intensity factors of axial cracks with large aspect ratio a/c (crack depth/half of surface crack length) was prepared. The crack shape was assumed a rectangular shape and the stress intensity factors at the deepest point of the crack were calculated with change of crack depth using FE analysis. By using these stress distribution and influence coefficients, a behavior of a PWSCC crack propagation at the safe end nozzles can be estimated easily and rationally.

  8. Studies of residual stress measurement and analysis techniques for a PWR dissimilar weld joint

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Naoki, E-mail: naoki2_ogawa@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., 2-1-1, Shinhama, Arai-cho, Takasago 676-8686 (Japan); Muroya, Itaru; Iwamoto, Youichi; Ohta, Takahiro; Ochi, Mayumi; Hojo, Kiminobu [Mitsubishi Heavy Industries, Ltd., 2-1-1, Shinhama, Arai-cho, Takasago 676-8686 (Japan); Ogawa, Kazuo [Japan Nuclear Energy Safety Organization, 3-17-1, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

    2012-02-15

    For evaluation of the PWSCC crack propagation behavior, a test model was produced using the same fabrication process of Japanese PWR plants and the stress distribution change was measured during a fabrication process such as a hydrostatic test, welding a main coolant pipe to the stainless steel safe end and an operation condition test. For confirmation of validity of the numerical estimation method of the stress distribution, FE analysis was performed to calculate the stress distributions for each fabrication process. From the validation procedure, a standard residual stress evaluation method was established. Furthermore for consideration of characteristics of PWSCC's propagation behavior of the dissimilar welding joint of the safe end nozzles, the influence coefficients at the deepest point for the stress intensity factors of axial cracks with large aspect ratio a/c (crack depth/half of surface crack length) was prepared. The crack shape was assumed a rectangular shape and the stress intensity factors at the deepest point of the crack were calculated with change of crack depth using FE analysis. By using these stress distribution and influence coefficients, a behavior of a PWSCC crack propagation at the safe end nozzles can be estimated easily and rationally.

  9. MELCOR 1.8.2 assessment: Surry PWR TMLB` (with a DCH study)

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

    1994-02-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified.

  10. TRITON/NEWT calculation of the Corail assembly for plutonium recycling in PWR

    International Nuclear Information System (INIS)

    The TRITON/NEWT code sequence of SCALE 5.1 and associated cross section libraries are benchmarked in this work against WIMS8 and APOLLO2 for the plutonium-bearing CORAIL assembly. The CORAIL assembly is designed for the multi-recycling of plutonium in PWR, using an heterogeneous configuration of UO2 and MOX fuel pins. The heterogeneity makes the modeling of this assembly type particularly challenging. TRITON/NEWT shows a satisfactory agreement on the evolution of k∞ with burnup, on the assembly pin-wise power peaking and on the burnup dependent concentration of most of the minor actinides. An exception is found to be the burnup-dependent evolution of 242mAm, which shows discrepancies of up to 60%. The reason was traced to the branching ratio of the (n, γ) reaction Of 241Am. Upon modification of the probability of formation of 242Am from the default 16.2% to 10% (used by WIMS8) in the ORIGEN-S binary libraries, the large discrepancy in the concentration of 242mAm disappeared. It is also found that by slightly varying the group structure of the standard 44-UOX library of SCALE 5.1 it is possible to get closer agreement with the results obtained using the standard 238-group library. (authors)

  11. Demonstration Test Program for Long-Term Dry Storage of PWR Spent Fuel

    International Nuclear Information System (INIS)

    In Japan, interim storage in multipurpose dry metal casks for maximum 50 years is planned for management of spent fuel until reprocessing. In the interim storage, cladding integrity of spent fuel will be maintained and safety of transportation will be ensured after the storage based on the knowledge and experience concerning integrity of spent fuel during dry storage in Japan and overseas. To ensure safety of transportation after storage, some of Japanese electric companies (The Japan Atomic Power Company, The Kansai Electric Company and Kyusyu Electric Company (hereinafter called “the utilities”)) are planning to conduct a long-term storage test of PWR spent fuel assemblies, which have not been used for dry storage in Japan, in the similar environment to actual casks and to confirm maintenance of the spent fuel integrity. In this test, the utilities plan to install a compact test container in the research facility, store one or two spent fuel assemblies in inert atmosphere for up to 60 years. Also, they plan to analyze the internal gas periodically and confirm the fuel cladding integrity. This document introduces the backgrounds, the overall test plan and designing of the test container. (author)

  12. ROSA-II experimental program for PWR LOCA/ECCS integral tests

    International Nuclear Information System (INIS)

    This paper is the final report of the ROSA-II experimental program, in which summary of the integral test results on thermal hydraulic behavior in a loss-of-coolant accident (LOCA) of pressurized water reactor (PWR) and on the effect of emergency core cooling system (ECCS) is presented. The ROSA-II test facility has a volume scaling factor of approximately 1/400 and core heating power of 2.4 MW. Specific feature of the facility is the versatility of the break conditions, the ECCS injection conditions and the secondary system conditions. After numbers of integral tests under various test conditions, (1) condensation-depressurization effect due to ECC water, (2) stored heat release from the structural materials and (3) counter current flow limitation (CCFL) at the specific locations were found to be important phenomena for the core cooling. To supply cooling water as soon as possible to the core was indicated to be very important for successful core cooling. Based on these results, more effective ECCS was proposed and the effectiveness of the proposed ECCS was experimentally verified. On the other hand, part of the experimental data was utilized to evaluate the predictability of RELAP-3 and RELAP-4J computer codes. (author)

  13. LiOH as corrosion inhibitor for component cooling water system in PWR

    International Nuclear Information System (INIS)

    The cooling water system for the auxiliary machines and coolers in the primary cooling system of a PWR forms closed loops, and has the function to prevent the release of fluid containing radioactive substances even if leak occurs in the primary cooling system. This system is mainly composed of carbon steel, and copper and its alloys are used for the sea water cooling. The auxiliary machines and coolers in the primary cooling system are made of stainless steel. Therefore, the rust prevention method used for this system must be effective for these metals or must not give harmful effect. As the rust prevention method for the plants in operation, chromic acid process and hydrazine process have been used, but the environmental pollution by chrome and the ammonia attack on copper alloys may occur. As the rust prevention process to improve these problems, LiOH method was examined. This cooling water system comprises four pumps, four water coolers, one surge tank, pipes and valves, and the water quality control in the system is explained. The rust prevention effect of LiOH for carbon steel, copper and its alloys was examined. Particular consideration is not required in the case of copper and its alloys, but pH higher than 12 is unsuitalbe. For the perfect rust prevention of carbon steel, the dissolved oxygen concentration must be less than 0.1 ppn, and pH must be more than 10. (Kako,I.)

  14. Out-of-pile performances of new zirconium alloys for PWR fuel cladding

    International Nuclear Information System (INIS)

    Two new zirconium alloys, N18 and N36, containing Sn, Nb, Fe and Cr have been developed to use as superior PWR fuel rod cladding materials. The results are obtained from the out-of-pile performance tests on these advanced alloy claddings or materials. Analytical electron microscopy demonstrated that the best out-of-pile corrosion resistance was obtained for microstructure containing a fine and uniform distribution of β-Nb and/or Zr(Fe, Cr)2 particles. Autoclave testing indicated that N18 and N36 alloys possessed superior corrosion resistance including uniform and nodular corrosion. It has been demonstrated that the hydrogen absorption data for all of alloys from corrosion reactions under various corrosion conditions showed a linear increase with the exposure time or oxide thickness, and hydrogen absorption rate of both alloys is quite low compared to that of Zircaloy-4. These alloys have demonstrated superior out-of-pile tensile strength, burst and creep properties relative to Zircaloy-4. In addition, the thermal physical properties, texture, Stress Corrosion Cracking (SCC) for two new zirconium alloys have been examined, which also showed a good results compared to Zircaloy-4. (author)

  15. Evaluation of hideout return data from U.S. PWR steam generators

    International Nuclear Information System (INIS)

    Since the middle to late 1970's, dramatic reductions in the quantities of impurities in the bulkwater of PWR steam generators have been made by U.S. utilities. Today most utilities operate at full power with impurity concentrations in the steam generator blowdown in the low ppb range, well within existing industry guideline control limits. Despite these efforts, some of these same utilities have subsequently encountered secondary side stress corrosion cracking (SCC) and intergranular attack (IGA) of steam generator tubing within deep tubesheet crevices and more recently at tube support intersections. It must, therefore, be concluded that either continuous low level input of contaminants within existing guideline limits, or intermittent short duration input, undetected by either current sampling and analysis techniques or procedures, are permitting ingress of corrosive impurity species which subsequently concentrate in flow-occluded regions to produce localized tube corrosion. To better understand both the quantity and composition of accumulated impurity species, more and more utilities, even those who have not experienced any steam generator corrosion, have begun to perform rigorous sampling and analysis evaluations of returning chemical contaminants each time the units are brought off-line. This paper will show examples of how these data are being used by U.S. industry to gain valuable information about accumulated contaminant inventories, to make cycle-to-cycle and plant-to-plant comparisons, and to develop plant specific actions to promote maximum contaminant removal. (author)

  16. Containment venting sliding pressure venting process for PWR and BWR plants

    International Nuclear Information System (INIS)

    In order to reduce the residual risk associated with hypothetical severe nuclear accidents, nuclear power plants in Germany as well as in certain other European countries have been or will be backfitted with a system for filtered containment venting. During venting system process design, particular importance is attached to the requirements regarding, for example, high aerosol loading capability, provision for decay heat removal from the scrubber unit, the aerosol spectrum to be retained and entirely passive functioning of the scrubber unit. The aerosol spectrum relevant for process design and testing varies depending on aerosol concentrations, the time at which venting is commenced and whether there is an upstream wetwell, etc. Because of this the Reactor Safety Commission in Germany has specified that SnO2 with a mass mean diameter of approximately 0.5 μm should be used as an enveloping test aerosol. To meet the above-mentioned requirements, a combined venturi scrubber system was developed which comprises a venturi section and a filter demister section and is operated in the sliding pressure mode. This scrubber system was tested using a full-scale model and has now been installed in 14 PWR and BWR plants in Germany and Finland

  17. Water-side oxide layer thickness measurement of the irradiated PWR fuel rod by NDT method

    International Nuclear Information System (INIS)

    It has been known that water-side corrosion of fuel rods in nuclear reactor is accompanied with the loss of metallic wall thickness and pickup of hydrogen. This corrosion is one of the important limiting factors in the operating life of fuel rods. In connection with the fuel cladding corrosion, a device to measure the water-side oxide layer thickness by means of the eddy-current method without destructing the fuel rod was developed by KAERI. The device was installed on the multi-function testing bench in the nondestructive test hot-cell and its calibration was carried out successfully for the standard rod attached with plastic thin films whose thicknesses are predetermined. It shows good precision within about 10% error. And a PWR fuel rod, one of the J-44 assembly discharged from Kori nuclear power plant Unit-2, has been selected for oxide layer thickness measurements. With the result of data analysis, it appeared that the oxide layer thicknesses of Zircaloy cladding vary with the length of the fuel rod, and their thicknesses were compared with those of the destructive test results to confirm the real thicknesses

  18. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    Science.gov (United States)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  19. Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

    Science.gov (United States)

    Vasiliev, A.; Ferroukhi, H.; Zhu, T.; Pautz, A.

    2014-04-01

    The development of a high-fidelity computational scheme to estimate the accumulated fluence at the tips of PWR control rods (CR) has been initiated at the Paul Scherrer Institut (PSI). Both the fluence from high-energy (E>1 MeV) neutrons as well as for the thermal range (Erod surfaces, especially for fully withdrawn CRs. In that case, the tip is located a short distance above the core/reflector interface and since this situation corresponds to a large part of reactor operation, the accumulated fluence will highly depend on the achieved calculation accuracy and precision in this non-fueled zone. The objective of the work presented in this paper is to quantify the influence of nuclear data on the calculated fluxes at the CR tips by (1) conducting a systematic comparison of modern neutron cross-section libraries, including JENDL-4.0, JEFF-3.1.1 and ENDF/B-VII.0, and (2) by quantifying the uncertainties in the neutron flux calculations with the help of available neutron cross-section variances/covariances data. For completeness, the magnitude of these nuclear data-based uncertainties is also assessed in relation to the influence from other typical sources of modeling uncertainties/biases.

  20. CFD simulation of fibre material transport in a PWR core under loss of coolant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, Thomas; Grahn, Alexander; Kliem, Soeren; Weiss, Frank-Peter [Forschungszentrum Dresden-Rossendorf e.V., Dresden (Germany). Inst. of Safety Research

    2010-05-15

    The aim of the numerical simulations carried out in this study was to determine how and where mineral wool fibres transported to the core by ECC water during a LOCA are deposited across the grid spacers of the fuel elements of a German PWR. The spacer grid is modelled as a strainer which completely retains the insulation material carried by the coolant and reaching the plane of the spacers. The accumulation of the insulation material gives rise to the formation of a compressible fibrous cake whose permeability to the coolant flow is calculated in terms of the local amount of deposited material and the local value of the superficial liquid velocity. The calculations showed that the fiber material at the uppermost spacer grid plane is not evenly distributed. First, it is accumulated at the positions of the break-through channels. Later when the inner circulation in the core has stopped, the insulation material can also be distributed into other regions of the spacer plane. Further investigations are necessary to determine the accumulation of insulation material for a longer period of time. Also steam production in the core or re-suspension of the insulation material during back flow should be considered. Moreover, the geometry modeling should be improved taking into account the real structures in the upper plenum and the geometry of the ECC injection nozzle ('Hutze'). (orig.)

  1. ERP-IV-A program for transient thermal-hydraulic analysis of PWR plant

    International Nuclear Information System (INIS)

    The author deal with the descriptions of physical model of transient process in PWR plant and the function of ERP-IV (ERR-IV Transient Thermo-Hydraulic Analysis Code). The code has been developed for safety analysis and design transient. The code is characterized by the multi-loop long-term, short term, wide-range plant simulation with the capability to analyze natural circulation condition. The description of ERP-IV includes following parts: reactor, primary coolant loops, pressurizer, steam generators, main steam system, turbine, feedwater system, steam dump, relive valves, and safety valves in secondary side, etc.. The code can use for accident analysis, such as loss of all A.C. power to power plant auxiliaries (a station blackout), loss of normal feedwater, loss of load, loss of condenser vacuum and other events causing a turbine trip, complete loss of forced reactor coolant flow, uncontrolled rod cluster control assembly bank withdrawal. It can also be used for accident analysis of the emergency and limiting conditions, such as feedwater line break and main steam line rupture. It can also be utilized as a tool for system design studies, component design, setpoint studies and design transition studies, etc

  2. PWR steam generator chemical cleaning process testing in model steam generators

    International Nuclear Information System (INIS)

    Corrosion related problems in PWR power plant steam generators have caused high maintenance costs, increased radiation exposure to plant personnel, and reduced unit availability. Two cleaning methods were investigated for their ability to clean deposits from steam generators thereby increasing the integrity of the steam generators and reducing personnel radiation exposure, due to reduced maintenance. First, an on-line chemical cleaning process (Chelant Addition) was tested for its ability to prevent corrosion product buildup in a steam generator. Second, an off-line dilute chemical cleaning process was tested to evaluate its ability to remove corrosion product deposits and leave minimal waste for disposal. These two processes were tested in model steam generators which simulated the operating conditions of a typical full size steam generator. Six model steam generators (MSG) were fabricated and qualified for their ability to reproduce denting at tube support plates. The results of six chemical cleaning tests and the post-cleaning destructive metallurgical evaluation of two of the model steam generators are reported

  3. Validation of the scale system for PWR spent fuel isotopic composition analyses

    International Nuclear Information System (INIS)

    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant

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

    International Nuclear Information System (INIS)

    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)

  5. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2001-03-15

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

  6. Sensitivity and uncertainty analysis for UO2 and MOX fueled PWR cells

    International Nuclear Information System (INIS)

    Highlights: • A method for calculating sensitivity coefficients has been improved. • The IR approximation was used in order to get accurate results. • Sensitivities and uncertainties are calculated using the improved method. • The method is applied for UO2 and MOX fueled PWR cells. • The verification was performed by comparing our results with MCNP6 and TSUNAMI-1D. - Abstract: This paper discusses the improvement of a method for calculating sensitivity coefficients of neutronics parameters relative to infinite dilution cross-sections because the conventional method neglects resonance self-shielding effect. In this study, the self-shielding effect is taken into account by using the intermediate resonance approximation in order to get accurate results in both high and low energy groups. The improved method is applied to calculate sensitivity coefficients and uncertainties of eigenvalue responses for UO2 and MOX (ThO2–UO2 and PuO2–UO2) fueled pressurized water reactor cells. The verification of the improved method was performed by comparing the sensitivities with MCNP6 and TSUNAMI-1D. For uncertainty, calculation comparisons were done with TSUNAMI-1D, and we demonstrate that the differences are caused by the use of different covariance matrices

  7. Test requirements for the integral effect test to simulate Korean PWR plants

    International Nuclear Information System (INIS)

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time

  8. Analysis of measured and calculated counterpart test data in PWR and VVER 1000 simulators

    Directory of Open Access Journals (Sweden)

    d’Auria Francesco

    2005-01-01

    Full Text Available This paper presents an over view of the "scaling strategy", in particular the role played by the counter part test methodology. The recent studies dealing with a scaling analysis in light water reactor with special regard to the VVER 1000 Russian reactor type are presented to demonstrate the phenomena important for scaling. The adopted scaling approach is based on the selection of a few characteristic parameters chosen by taking into account their relevance in the behavior of the transient. The adopted computer code used is RELAP5/Mod3.3 and its accuracy has been demonstrated by qualitative and quantitative evaluation. Comparing experimental data, it was found that the investigated facilities showed similar behavior concerning the time trends, and that the same thermal hydraulic phenomena on a qualitative level could be predicted. The main results are: PSB and LOBI main parameters have similar trends. This fact is the confirmation of the validity of the adopted scaling approach and it shows that PWR and VVER reactor type behavior is very similar. No new phenomena occurred during the counter part test, despite the fact that the two facilities had a different lay out, and the already known phenomena were predicted correctly by the code. The code capability and accuracy are scale-independent. Both character is tics are necessary to permit the full scale calculation with the aim of nuclear power plant behavior prediction. .

  9. Fatigue-crack growth behavior of Type 347 stainless steels under simulated PWR water conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seokmin; Min, Ki-Deuk; Yoon, Ji-Hyun; Kim, Min-Chul; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Fatigue crack growth rate (FCGR) curve of stainless steel exists in ASME code section XI, but it is still not considering the environmental effects. The longer time nuclear power plant is operated, the more the environmental degradation issues of materials pop up. There are some researches on fatigue crack growth rate of S304 and S316, but researches of FCGR of S347 used in Korea nuclear power plant are insufficient. In this study, the FCGR of S347 stainless steel was evaluated in the PWR high temperature water conditions. The FCGRs of S347 stainless steel under pressurized-water conditions were measured by using compact-tension (CT) specimens at different levels of dissolved oxygen (DO) and frequency. 1. FCGRs of SS347 were slower than that in ASME XI and environmental effect did not occur when frequency was higher than 1Hz. 2. Fatigue crack growth is accelerated by corrosion fatigue and it is more severe when frequency is slower than 0.1Hz. 3. Increase of crack tip opening time increased corrosion fatigue and it deteriorated environmental fatigue properties.

  10. Fatigue Crack Growth Rate Behavior of Type 347 Stainless Steel in Simulated PWR Water Environment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The pressurizer surge line of a Korean standard nuclear power plane uses Nb stabilized type 347 stainless steel. The pressurizer surge line is the pipe connecting the pressurizer and the hot leg line, and the path controlling the pressure and temperature of the cooling system of the nuclear reactor, operated at 316 .deg. C and in a 150atm. The pressurizer surge line operated at high temperature and high pressure receives thermal stress by a temperature change and mechanical stress by a pressure change at the same time, and by being exposed to the high temperature and high pressure cooling water environment of a nuclear power plant, environmental fatigue by stress and corrosion is the main damage instrument. As the effect of environmental fatigue has been reported, through low cycle fatigue, fatigue life evaluations of austenite stainless steel have been conducted, but evaluations of fatigue crack growth rate to evaluate the soundness are very poor. In this study, evaluated characteristics of fatigue crack growth rate base on a change of dissolved oxygen in a PWR environment

  11. Evaluation of on-line chelant addition to PWR steam generators. Steam generator cleaning project

    International Nuclear Information System (INIS)

    The investigation of chelating agents for continuous water treatment of secondary loops of PWR steam generators were conducted in two general areas: the study of the chemistry of chelating agents and the study of materials compatability with chelating agents. The thermostability of both EDTA and HEDTA metal chelates in All Volatile Treatment (AVT) water chemistry were shown to be greater than or equal to the thermostability of EDTA metal chelates in phosphate-sulfite water chemistry. HEDTA metal chelates were shown to have a much greater stability than EDTA metal chelates. Using samples taken from the EDTA metal chelate thermostability study and from the Commonwealth Research Corporation (CRC) model steam generators (MSG), EDTA decomposition products were determined. Active metal surfaces were shown to become passivated when exposed to EDTA and HEDTA concentrations as high as 0.1% w/w in AVT. Trace amounts of iron in the water were found to increase the rate of passivation. Material balance and visual inspection data from CRC model steam generators showed that metal was transported through and cleaned from the MSG's. The Inconel 600 tubes of the salt water fouled model steam generators experienced pitting corrosion. Results of this study demonstrates the feasibility of EDTA as an on-line water treatment additive to maintain nuclear steam generators in a clean condition

  12. Validation of the scale system for PWR spent fuel isotopic composition analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; Bowman, S.M.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Laboratories, Las Vegas, NV (United States)

    1995-03-01

    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant.

  13. Corrosion of nickel alloys and stainless steels in polluted or confined PWR environments

    International Nuclear Information System (INIS)

    This document addresses the issue of corrosion of materials used in PWR nuclear reactors, notably in steam generators which have been particularly affected by this kind of degradation due to a progressive accumulation of impurities. The authors first present the different materials used in secondary circuit and in auxiliary circuits of PWRs: carbon steels and low alloyed steels, nickel alloys, stainless steels, and other materials. They discuss the degradation of steam generator tubes by corrosion: corrosion environments, types of corrosion (wastage, pitting, intergranular stress corrosion cracking), and influence of the environment and of the microstructure. They also propose a brief overview of modelling efforts in the case of the 600 alloy, and indicates measures to mitigate the tube degradation by corrosion (water treatment, better design of steam generators and secondary circuit, improvement of corrosion resistance). The next part addresses the degradation by stress corrosion cracking of stainless steels in polluted environments in PWRs reactors: return on experience, stress corrosion cracking in media contaminated by impurities (intergranular corrosion of sensitized or work hardened steels, transgranular corrosion by chloride ions, corrosion by diluted sulphate + chloride, corrosion in concentrated boric acid solutions)

  14. Project management of the build of the shore test facility for the prototype of PWR II

    International Nuclear Information System (INIS)

    The PWR II is a new design of nuclear steam raising plant for the Royal Navy's submarines. It features improved engineering for safety, increased power, increased shock resistance, reduced noise transmission to sea and reduced manning requirement. It is to be tested in a new prototype testing facility, the Shore Test Facility, which is a section of submarine hull containing a prototype of the nuclear steam raising plant and its support system. It is installed at the Vulcan Naval Reactor Test establishment at Dounreay in Scotland. The function of the establishment is to test new designs of core and reactor plant, validate the mathematical models used in their design, develop improved methods of operation and maintenance of the plant and test new items of equipment. The Shore Test Facility was built in large sections at Barrow-in-Furness and transported to Scotland. The project management for the construction of the Shore Test Facility is explained. It involves personnel from the Royal Navy, and a large number of people working for the contractors involved in the buildings, transportation, operation and maintenance of the Facility. (U.K.)

  15. The effect of zinc addition on PWR corrosion product deposition on zircaloy-4

    International Nuclear Information System (INIS)

    During the period 1995 to 2001 a programme of loop irradiation tests have been performed to confirm the effectiveness of zinc additions on PWR circuit chemistry and corrosion. The programme included two loop irradiation experiments, and subsequent PIE; the experiments were a baseline test (no added zinc) and a test with added zinc (10 ppb). This paper addresses the findings regarding corrosion product deposition and activation on irradiated Zircaloy-4 surfaces. The findings are relevant to overall corrosion of the reactor primary circuit, the use of zinc as a corrosion inhibitor, and activation and transport of corrosion products. The irradiation experience provides information on the equilibration of the loop chemistry, with deliberate injection of zinc. The PIE used novel and innovative techniques (described below) to obtain samples of the oxide from the irradiated Zircaloy. The results of the PIE, under normal chemistry and zinc chemistry, indicate the effect of zinc on the deposition and activation of corrosion products on Zircaloy. It was found that corrosion product deposition on Zircaloy is enhanced by the addition of zinc (but corrosion product deposition on other materials was reduced in the presence of zinc). Chemical analysis and radioisotope gamma counting results are presented, to interpret the findings. A computer model has also been used to simulate the corrosion product deposition and activation, to assist in the interpretation of the results. (authors)

  16. Study of thermal hydraulic behaviors during steam generator multiple tube rupture events in PWR

    International Nuclear Information System (INIS)

    Since the occurrence probability of multiple steam generator tube rupture (MSGTR) in PWR is low, analytical or experimental investigation for termination of such accidents is not performed explicitly. Therefore, thermal-hydraulic analysis of the plant behavior under the MSGTR (10 and 50 tubes) in all loops or in a single loop with station black out (SBO) were made in this study with use of the code RETRAN3D, and investigations for accident management (AM) of such accidents were made in order to contribute to continuous risk reduction efforts in the future. This study indicated that the water supply function to the SG is quite important for coping with accidents involving MSGTR accompanied by SBO to prevent core damage. Further, if the auxiliary feed water system loses its function, the time to reach the core exposure is predicted to be reduced by 1 hour or more in a MSGTR case as compared to a single-tube SGTR case. Therefore, in order to prevent the core damage during MSGTR, it is desirable to have alternative water injection equipment operable to quickly replace auxiliary feed water system if it fails and to increase the reliability of the auxiliary feed water system. (author)

  17. AP1000® PWR reactor physics analysis with VERA-CS and KENO-VI. Part 1. Zero power physics tests

    International Nuclear Information System (INIS)

    Westinghouse has applied the Core Simulator of the Virtual Environment for Reactor Applications, VERA-CS, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000® PWR. The AP1000 PWR features and advanced first core with radial and axial heterogeneities, including enrichment zoning, multiple burnable absorbers, and a combination of light and heavy control banks to enable the MSHIMTM advanced operational strategy. These advanced features make application of VERA-CS to the AP1000 PWR first core especially relevant to qualify VERA performance. A companion paper at this conference describes the power distribution analysis of the AP1000 PWR with VERA-CS and the KENO Monte-Carlo code. This paper describes the results obtained for the startup physics tests simulations of the AP1000 PWR first core (critical boron, rod worth and reactivity coefficients), supporting the excellent numerical agreement reported in the companion paper for the power distribution. (author)

  18. 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)

    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)

  19. Courting Chinese Investors

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Inviting Chinese business delegations. Participating in trade fairs in China.Building an industrial park for Chinese investors. Poland is doing all theseand more to attract investment from China, according to Krzysztof Szumski,Poland ’s Ambassador to Chi

  20. Why I Learn Chinese

    Institute of Scientific and Technical Information of China (English)

    Keltoum; Otamani

    2013-01-01

    <正>I am eighteen years old. I have learned Chinese at the FFCA NordPas-de Calais for three years. I will continue to study Chinese during my five-year university life and try to get a job dealing with China in the future. I became interested in China at the age of 13, and my interest grew year by year.