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Sample records for chinese 300mwe pwr

  1. Structural mechanics research and development for main components of Chinese 300 MWe PWR NPPs: from design to life management

    International Nuclear Information System (INIS)

    Qinshan Nuclear Power Plant (Unit I), is a 300 MWe prototype PWR independently developed by Chinese own efforts, from design, manufacture, construction, installation, commissioning, to operation, inspection, maintenance, ageing management and lifetime assessment. Shanghai Nuclear Engineering Research and Design Institute (SNERDI) has taken up with and involved in deeply the R and D to tackle problems of this type of reactor since very beginning in early 1970s. Structural mechanics is one of the important aspects to ensure the safety and reliability for NPP components. This paper makes a summary on role of structural mechanics for component safety and reliability assessment in different stages of design, commissioning, operation, as well as lifetime assessment on this type PWR NPPs, including Qinshan-I and Chashma-I, a sister plant in Pakistan designed by SNERDI. The main contents of the paper cover design by analysis for key components of NSSS; mechanical problems relating to safety analysis; special problems relating to pressure retaining components, such as fracture mechanics, sealing analysis and its test verifications, etc.; experimental research on flow-induced vibration; seismic qualification for components; component failure diagnosis and root cause analysis; vibration qualification and diagnosis technique; component online monitoring technique; development of defect assessment; methodology of aging management and lifetime assessment for key components of NPPs, etc. (authors)

  2. IPSN expert appraisal programme on the chooz A 300 MWe PWR. Lessons learned by IPSN

    International Nuclear Information System (INIS)

    The closure of Chooz A PWR provided an opportunity to take samples of items that had aged in situ in conditions close to those encountered in PWR in operation over a period of 140.000 hours, which is far longer than the usual time-spans of simulated laboratory tests. 4 topics have been studied: 1) effect of radiation on reactor vessel internals, 2) dissimilar metal joints of reactor coolant system: pressurizer surge line, 3) cast parts of austeno-ferritic steel: hot and cold leg primary valves, and 4) ageing of cables in high temperatures and under irradiation. The examination of the lower internals on some baffle angle bracket and core shroud screws, subjected to varying amounts of irradiation, did not reveal any cracking or corrosion, and confirmed the saturation effect between 4 and 10 dpa for the hardening of 304 austenitic steel in the low temperature range. Expert appraisal of the dissimilar metal joints on the pressurizer surge line confirmed the existence of small fabrication defects due to high temperature cracking. Expert appraisal of the 3 valve body samples from the main section of the coolant system confirmed that -) thermal ageing of the valve body on the hot leg was more advanced than that of the cold leg valve, -) the material of the valve housing on the cold leg which, in theory, was not sensitive to ageing phenomena, exhibited unexpectedly low impact strength values. As for cables, measurements confirmed that their mechanical and electrical properties remained sufficient for them to carry out their functions. (A.C.)

  3. 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

  4. Comprehensive research on sealing behaviour of reactor vessel of 300 MWe nuclear power plant

    International Nuclear Information System (INIS)

    The general conception of a special research on sealing behaviour of PWR vessel is described and the major results centering on the establishment of sealing analysis program system and its experimental verification, along with the description on the development and measurement of sealing ring, the thermal sealing test and the relevant analysis are given. On the basis of the above approach, the vessel sealing behaviours of 300 MWe Qinshan Nuclear Power Plant are evaluated. A concept on the classification of pressure vessels and their sealing criteria are proposed. Two viewpoints on the analysis are suggested, which are that the vessel sealing deformation analysis should be regarded as a basis of the general stress analysis and that bolt loading increment caused by the bolt temperature lag should be taken as a key point when considering the thermo-contact coupling in transient sealing analysis. The understanding about the sealing mechanism are expounded and the thermal equivalent of hydrostatic test is discussed

  5. Qinshan 300Mwe NPP full scope simulator upgrade

    International Nuclear Information System (INIS)

    On April 28,2004, RINPO was awarded the project for Qinshan 300Mwe NPP full scope simulator upgrade, the SAT (site acceptance test) was completed on June 30 2005 and the simulator put into operator training again. Scope of upgrade includes: computer system (DGI server and workstations) all replaced by microcomputers; G2 I/O controllers all replaced by RTP EIOBC; Unix-based simulation support environment replaced by RINPO's PC-based simulation environment RINSIMTM, Instructor software replaced by RINPO's PC-based instructor software with function and diagram redesigned; DEH, Feed-water control and some other digital control systems redeveloped to follow NPP modifications; desk-top simulator with soft panel control room developed as byproduct; most of the models not changed but it is planned the reactor core and PPC model will be upgraded in near future. SAT of upgrade demonstrates that the performance of the simulator much improved after the upgrade. (author)

  6. Study and economics analysis for 18-month refueling management on power uprate of a 300 MWe NPP

    International Nuclear Information System (INIS)

    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 300 MWe NPP after power uprate. The results show that the concluded fuel management scheme for a 300 MWe NPP after power uprate achieves the projected 18- month refueling cycle design objectives with the nominal thermal power of 1250 MW and meets the design criteria. As compared to the current fuel management strategy of a 300 MWe 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 300 MWe NPP and optimization of fuel management strategy. (authors)

  7. 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.

  8. 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

  9. Improved NOx emissions and combustion characteristics for a retrofitted down-fired 300-MWe utility boiler.

    Science.gov (United States)

    Li, Zhengqi; Ren, Feng; Chen, Zhichao; Liu, Guangkui; Xu, Zhenxing

    2010-05-15

    A new technique combining high boiler efficiency and low-NO(x) emissions was employed in a 300MWe down-fired boiler as an economical means to reduce NO(x) emissions in down-fired boilers burning low-volatile coals. Experiments were conducted on this boiler after the retrofit with measurements taken of gas temperature distributions along the primary air and coal mixture flows and in the furnace, furnace temperatures along the main axis and gas concentrations such as O(2), CO and NO(x) in the near-wall region. Data were compared with those obtained before the retrofit and verified that by applying the combined technique, gas temperature distributions in the furnace become more reasonable. Peak temperatures were lowered from the upper furnace to the lower furnace and flame stability was improved. Despite burning low-volatile coals, NO(x) emissions can be lowered by as much as 50% without increasing the levels of unburnt carbon in fly ash and reducing boiler thermal efficiency. PMID:20429548

  10. 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)

  11. Experimental research progress on passive safety systems of Chinese advanced PWR

    International Nuclear Information System (INIS)

    TMI and Chernobyl accidents, having pronounced impact on nuclear industries, triggered the governments as well as interested institutions to devote much attention to the safety of nuclear power plant and public's requirements on nuclear power plant safety were also going to be stricter and stricter. It is obvious that safety level of an ordinary light water reactor is no longer satisfactory to these requirements. Recently, the safety authorities have recommended the implementation of passive system to improve the safety of nuclear reactors. Passive safety system is one of the main differences between Chinese advanced PWR and other conventional PWR. The working principle of passive safety system is to utilize the gravity, natural convection (natural circulation) and stored energy to implement the system's safety function. Reactors with passive safety systems are not only safer, but also more economical. The passive safety system of Chinese advanced PWR is composed of three independent systems, i.e. passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system. This paper is a summary of experimental research progress on passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system

  12. 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.

  13. On the domestically-made heavy forging for reactor pressure vessels of PWR nuclear power plant

    International Nuclear Information System (INIS)

    The present situation of the foreign heavy forgings for nuclear reactor pressure vessels and the heavy forgings condition which is used for the Qinshan 300MWe nuclear power plant are described. Some opinions of domestic products is proposed

  14. 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.

  15. The integrated PWR

    International Nuclear Information System (INIS)

    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.)

  16. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

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

  17. 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

  18. 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.)

  19. 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.)

  20. The PWR programme

    International Nuclear Information System (INIS)

    For fueling the PWR type reactors two types of fuel were developed: the UO2 and mixed oxide fuels. To satisfy the demand of the operators of UO2-fuelled power plants a specific industrial organization has been established by Cogema and Framatome: Framagema supplies the technical expertise and sells the fuel; FBFC (Societe Franco-Belge de Fabrication Combustible) is manufacturing the fuel by using particularly the zirconium components produced by Zircotube and Cezus. By making possible the recycling of the materials recovered from the spent fuel reprocessing the MOX (mixed oxide fuels) technology represents an important venture for the future electronuclear sector. To implement this project Cogema created together with Belgonucleaire (the administrator of the Dessel manufacture plant) the GIE COMMOX, in charge with marketing of this fuel. On the other side Cogema which produces MOX in its facility at Cadarache, is at present building the plant at Melox of a capacity of 120 tonnes/year. After presenting the present situation with UO2 and MOX fuels the paper ends with considerations concerning the future fuels and fuels for future and further future reactors

  1. System comparative analysis of the most advanced pressured water reactors (PWR, WWER) and boiling water reactors (BWR) projects with the aim to choose the reactors for NPP construction in Kazakhstan

    International Nuclear Information System (INIS)

    Full text: The official decision on construction of a Nuclear Power Plant (NPP) in Kazakhstan has been accepted by the Kazakhstan government. The results on the choice of the power reactors projects of the NPP are given in the report. The choice has been carried out with the aim to develop recommendation on reactors of the NPP for construction in Kazakhstan. The choice of the reactors was based on the system comparative analysis of the most advanced power reactors projects using 15 criteria system of the nuclear, radiating and ecological safety and economic competitiveness. Following Pressurized Water Reactor (PWR, WWR) projects have been subjected to the system comparative analysis: 1) Large Sized Reactors (700 MW(el) and up): such as EPR, developed by Germany Siemens and France Framatome companies; CANDU-9, heavy-water reactor, developed by Atomic Energy of Canada Ltd (AECL); System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation, developed by Korea Power Engineering Company, Inc.; APWR, Japanese advanced reactor, developed by Japan Atomic Power Company, Japan, Mitsubishi Heavy Industries, Japan and Westinghouse Electric Company, USA; WWER-1000 (V-392) - development by Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor, development by Westinghouse, USA/Genesi, Italy. 2) Medium Sized Reactors (300 MWe - 700 MWe): AP-600, passive PWR, developed by the Westinghouse company; CANDU-6, heavy-water reactor, developed by Atomic Energy of Canada Ltd (AECL); An-tilde-600, passive PWR, developed by Nuclear Power Institute of China; WWER-640, Russian passive reactor, developed by 0KB ''Gidropress'' Experimental and Design Office, Russian Federation; MS-600, developed by Mitsubishi Company; KSNP-600, developed by Korea Power Engineering Company, Inc., South Korea. 3) Small Sized Reactors (a few MWe- 300 MWe): IRIS, reactor of IV generation, developed by the International Corporation of 13

  2. 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.)

  3. Condensate purification in PWR reactors

    International Nuclear Information System (INIS)

    The recommendations made by the VGB task group on 'condensate purification for PWR reactors' 1976 are discussed in detail. Techniques and circuiting possibilities of condensate purification for BBR steam generators (forced circulation) and KWU steam generators (U tube with blow-down) are mentioned. (HP)

  4. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

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

  5. PWR AXIAL BURNUP PROFILE ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Acaglione

    2003-09-17

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

  6. Overview of PWR chemistry options

    Energy Technology Data Exchange (ETDEWEB)

    Nordmann, F.; Stutzmann, A.; Bretelle, J.L. [Electricite de France, Central Labs. (France)

    2002-07-01

    EDF Central Laboratories, in charge of engineering in chemistry, of defining the chemistry specifications and studying the operation feedback and improvement for 58 PWR units, have the opportunity to evaluate many options of operation developed and applied all around the world. Thanks to these international relationships and to the benefit of a large feedback from many units, some general evaluation of the various options is discussed in this paper. (authors)

  7. 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)

  8. Thermodynamic modelling of PWR coolant

    International Nuclear Information System (INIS)

    Spinel solubilities on PWR primary circuit surfaces vary with temperature, pH and coolant H2 concentration. The available solubility data are discussed for Fe, Ni, Co and Zn oxides, and species are identified where data are very limited or absent. An equilibrium thermodynamic model is described to predict the solubility, and results are described predicting relative Fe and Ni solubility under normal operating conditions and during shutdown/startup. The relative stabilities of stoichiometric and non-stoichiometric zinc ferrite spinels are also considered. (R.P.)

  9. CAREM: an innovative-integrated PWR

    International Nuclear Information System (INIS)

    Power levels. In this regard the cost effective sizes are 100 MWe (maintaining natural circulation for primary core cooling) and 300 MWe (using integrated primary pumps). (author)

  10. 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

  11. Technical support to an operating PWR vis-a-vis safety analysis

    International Nuclear Information System (INIS)

    Currently a PWR of 300 MWe capacity CHASNUPP-I is in operation since the year 2000. Technical support being provided includes In-core fuel management and corresponding safety analysis for the reshuffled core for the next cycle. Prior to start of cycle six an extension in cycle five based on coast down technique was achieved of almost 30 effective full power days. Cycle 6 was designed to achieve the safe and economical loading pattern. The technique used is designated as out-in mode (modified). In this technique, most of the fresh fuel assemblies are not directly located at the periphery of the core, but near the boundary. This technique has the advantage that no burnable absorbers are used in each cycle and we get less radial neutron leakage and increased discharge burnup and cycle length. Operating experience/feedback shows that this type of loading pattern gives better economy without resorting to the conventional in-out technique. The lifetime of the cycle is predicted as 10371 MWD/MTU or 373 Effective Full Power Days (EFPD at 998.6 MWth). In design calculations, the end of cycle is reached at 10 ppm critical boron concentration in the unroded core. Measured critical boron concentration at HZP, BOL is 1453 ppm compared with the calculated value i.e 1457 ppm, is within the acceptable limits. It is also observed that the calculated reactivity worth of Tl is -1771 pcm as compared to measured value i.e -1802 pcm with difference of only 1.6 % showing the reliability of the design value. The measured Moderator temperature coefficient (MTC) is 2.52 pcm/deg. C at all rods out (ARO) and critical boron concentration (CBC) condition whereas the calculated value is 3.36 pcm/deg. C (at predicted CBC of 1457) having a good agreement with design value. Safety evaluation of cycle 6 was carried out for the reshuffled core. All the probable accident scenarios based on initiating events as given in the FSAR were evaluated with respect to input parameters. For a specific event, the

  12. PWR fuel: experience and development

    International Nuclear Information System (INIS)

    The start-up of the large French nuclear program has rapidly led FRAGEMA to be one of the first PWR fuel suppliers. FRAGEMA is a joint subsidiary of two companies whose scopes of supply are fully complementary: FRAMATOME (NSSS vendor) and COGEMA (nuclear fuel cycle service supplier). At the center of these two activities FRAGEMA is in charge of designing and marketing fuel assemblies. Assistance is also offered to nuclear power plant operators in all fuel related fields by providing a wide range of services and a number of specialized components. Over the past years a statistical data base has been accumulated on fuel assembly behaviour under various operating conditions. At the same time extensive experimental programs have been, set up to develop advanced products to cope with utilities needs in the future. An overview of these two sides of our experience is presented in the following

  13. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  14. Design and construction of the PCPV for the 300 MWe THTR nuclear power station in West Germany

    International Nuclear Information System (INIS)

    In July 1972 the order was placed in Germany for the first PCPV comprising concrete structure, liner, cooling system and insulation to a consortium under the direction of KRUPP UNIVERSALBAU. The prestressed concrete structure itself was designed and constructed by this company. Extensive tests were carried out on the limestone concrete to establish all the physical properties. Special efforts were made to produce a mix which was both pumpable and generated a minimum amount of heat of hydration. As a departure from normal practice, the cylindrical parts of the vessel are constructed in complete rings up to 2m in height and of the full wall thickness. Experiments showed that, for this method of construction, the temperature difference between the old and the new concrete should not be allowed to exceed 50C. To achieve this, ice cooled water is used in the concrete mix and, in the summer time, liquid nitrogen is added at the time of mixing. The thermal behavior of the concrete has been monitored throughout the construction period. A novel construction feature worth mentioning is that the internal insulation and parts of the core structure were already erected before the construction of the concrete cylinder was complete. This was achieved by providing a temporary closure at the top of the cylinder to maintain clean conditions below. The overall stress calculations and the detailed stress pattern for the lower half of the vessel were carried out by using an axi-symmetric computer program but, for the upper half of the cylinder, a three-dimensional analysis was necessary (due to its geometric arrangement). To prove the safety of the vessel a structural model was used from which the mode of failure was found using a kinematic chain and thus the factor of safety established. A secondary line of safety is the integrity of the liner. (author)

  15. 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.

  16. Maintenance robot for PWR plant

    International Nuclear Information System (INIS)

    The remote operation, automatic machines utilized in the field of the maintenance of component machinery and equipment in nuclear power plants, so-called maintenance robots, have produced effects in the reduction of radiation exposure, the improvement of the quality of working, the shortening of working time and so on, but still many robots have their specialized functions. The expectation of present day society to robots has been diversified, and the technical development of high function robots is advanced positively. In this report, the recent examples of the high function robots developed for PWR power stations with the support of technical progress and the trend of the technical development are explained. The needs and seeds of maintenance robot development are discussed. As the examples of heightening the functions of maintenance robots, the next generation ultrasonic testing machine highly advanced by sensor technology and size and weight reduction mechanism technology, the intelligent monitoring system for welding using AI technology and other manpower-saving robots are shown. (K.I.)

  17. Thermodynamic modelling of PWR coolant

    International Nuclear Information System (INIS)

    Corrosion products released from PWR and VVER primary circuit surface oxides are transported in the coolant to the core, where they deposit and are activated to form radioactive corrosion products, which can be re-released to re-deposit on out-of-core surfaces. Spinel solubilities vary with the pH, temperature and sometimes the hydrogen concentration of the coolant. This paper describes the development of an equilibrium thermodynamic model to predict such changes, and discusses the extent of the available solubility data for Fe, Ni, Co and Zn oxides. Results are described on the relative solubility of Fe and Ni under both normal operating conditions and during shutdown/start-up, and on the relative stabilities of stoichiometric and non-stoichiometric zinc ferrite spinels. Comparison of the calculated corrosion product concentrations with reactor measurements indicates that, in reactors with low Ni content in the steam generator alloys, the concentration of Ni in the coolant is limited by its availability in the surface oxide. In reactors with high-Ni alloys, the circulating Ni concentrations may be dominated by colloidal material. The calculated changes in Ni and Fe concentrations during the acid-reducing phase of shutdown are in reasonable agreement with measurements from Sizewell B. The paper highlights the need for a more comprehensive open corrosion product data base, the need to consider both boiling and radiolysis in the core on corrosion product solubility in different parts of the primary circuit and, finally, the importance of kinetic factors at low temperature behaviour during shutdown and start-up. (author)

  18. 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.)

  19. 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)

  20. 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.)

  1. 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)

  2. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL)

  3. Full MOX core design for PWR

    International Nuclear Information System (INIS)

    Full MOX core design for APWR was analyzed in nuclear design, fuel integrity analysis, thermal hydraulic design and safety analysis et. al. Feasibility of Full MOX core was confirmed from these analyses without any large modifications. Full MOX PWR core has very good characteristics in which single Pu content in an assembly, burnable poison free, higher burnup and longer cycle operation are feasible. (author)

  4. 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.)

  5. REWET, PWR LOCA accident experiments

    International Nuclear Information System (INIS)

    1 - Description of test facility: The REWET-II facility was designed for the investigation of the reflooding phase of a LOCA. The main design principle is the accurate simulation of the rod bundle geometry and the primary system elevations. This is necessary in order to have the correct flow channels and hydrostatic pressures for the reflooding process. The reactor vessel is simulated by a stainless steel U-tube construction consisting of downcomer, lower plenum, core and upper plenum. The primary loops contain a pipe simulating the broken loop and a connection line between the upper plenum and the downcomer simulating five intact loops. The containment is simulated by a pressure vessel (not in scale). Steam generators and primary pumps are simulated with flow resistances. The ECC-water can be injected to the downcomer and/or to the upper plenum by a pump or from an accumulator. All the elevation in the reactor vessel simulator are scaled to 1:1 (except the reactor upper head). The scale of the volumes and flow areas is 1:2333 referring to the number of the fuel rod simulators in the facility and the fuel rods in the reference reactor. The rod bundle is either in a hexagonal shroud, the inside distance of the opposite walls is 54.3 mm and the wall thickness 2 mm, or in a round shroud, the inner diameter is 66.0 mm and the wall thickness 2 mm. The simulation of the fuel-rod bundle consists of 19 indirectly electrically heated simulator rods. The heating coils are inside stainless steel claddings in magnesium oxide insulation. The heated length, the outer diameter and the lattice pitch of the fuel-rod simulators as well as the number (= 10) and construction of the rod bundle spacers are the same as in the reference reactor. The upper ends of the rods are attached to the upper tie plate. 2 - Description of test: Pressurized water reactors in use in Finland reactors have certain unique features which make them different from most other PWR designs. The 6 horizontal

  6. [Methodologies for optimization of maintenance and testing of safety related equipment at NPPs in Pakistan

    International Nuclear Information System (INIS)

    In Pakistan, a 137 MWe PHWR type NPP (KANUPP) is in operation since 1971, and a 300 MWe Chinese design PWR (CHASNUPP) is under construction. The under construction PWR is planned to be connected to the national grid in 1998. Under this Coordinated Research Project, the work is planned to be carried out for improvement and optimization of the maintenance and surveillance programme for safety related systems and equipment of the above mentioned two NPPs. Efforts will be directed to acquire latest knowledge regarding various methods and strategies for surveillance testing and plant technical specifications through exchange of information. This project will provide a good opportunity to the regulatory body regarding development of acceptance criteria for testing and maintenance of safety related systems and equipment. 8 refs

  7. Horizontal Drop of 21- PWR Waste Package

    International Nuclear Information System (INIS)

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in-terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 1 1) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design

  8. Progress of PWR reactor fuels: OSIRIS equipments

    International Nuclear Information System (INIS)

    The experimental reactor Osiris situated at the Saclay Nuclear Centre is a reactor fitted with tests and monitoring facilities. Of the pool and open core type, it can test the test fuel of PWR power stations under high neutron flux. The characteristic stresses of the operating states of power reactors can be reproduced in experimental devices suited to the various study subjects, be this the creep and deformation of zircaloy claddings, the behavior of fuel rods to power ramps, to load following, to remote regulation, to the cooling state in double phase or just analytical tests. The experimental irradiation devices extend from the single static coolant capsule, such as the NaK alloy, to the dynamic coolant test loop that operates in the cooling conditions representative of PWR's including water chemistry. Ancillary devices make it possible to carry out examinations and non-destructive testing: immersed neutron radiography, gamma scanning visualization monitoring device, eddy currents, profilometering

  9. Thorium fuel cycle study for PWR applications

    International Nuclear Information System (INIS)

    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 UO2 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 ThO2-PuO2 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 UO2 fuel. (author). 6 refs., 3 tabs., 6 figs

  10. Horizontal Drop of 21- PWR Waste Package

    International Nuclear Information System (INIS)

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 11) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design

  11. 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.

  12. Space-dependent dynamics of PWR

    International Nuclear Information System (INIS)

    The azimuthal dependent reactor dynamics coupled to thermohydraulics are studied by using the neutron-flux and coolant temperature signals measured at an actual PWR. The second azimuthal mode of neutron-flux fluctuation was found, and the coupling of the mode to thermohydraulics of the coolant was suggested. The coherent coolant flow in the reactor core seems to sustain this spatial oscillation mode. (authors)

  13. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  14. 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

  15. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.)

  16. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    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.)

  17. Steam shut-off valves for PWR type reactors

    International Nuclear Information System (INIS)

    Fast acting closure means are requested in PWR type reactors as well as in BWR to safely shut-off the live steam at the turbine input in the event of accident. The design and control system of steam shut-off valves acted by the fluid system and intended for PWR type reactors, are described. The role of these valves in a PWR is discussed with the specified requirements involved

  18. 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)

  19. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  20. Dismantling and decommissioning experience of commercial PWR

    International Nuclear Information System (INIS)

    Regarding the relatively youthness of FRAMATOME PWR's in operation none of these reactor needs to be decommissioned before 1992. However feasibility studies have been carried out by FRAMATOME for an on site entombment of active components and heavy equipments. In the past, partial dismantling of the reactor internals of the CHOOZ reactor: PWR of 320 MWe and a complete removal of the thermal shield protecting the reactor vessel were conducted successfully. After repair, the reactor power output has been upgraded of 10% and the reactor operates satisfactorily since 1970. More recently the discovery of scarce defects affecting centering pins of control guide tube located in the upper reactor internals of 900 MWe plants has initiated the construction of several ''Hot stand equipments'' for the systematic replacement of these centering pins. FRAMATOME is presently actively studying possible options consisting either to extend the plant life beyond its initial licence life, or to convert classical PWR into an advanced reactor more economical in terms of uranium consumption

  1. Pu-breeding feasibility in PWR

    International Nuclear Information System (INIS)

    This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO2 or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived

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

    International Nuclear Information System (INIS)

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

  3. 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.)

  4. PWR Core 2 Project accident analysis

    International Nuclear Information System (INIS)

    The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16,500 kilograms of depleted natural uranium (as UO2), 139 kilograms of plutonium, 2.8 megacuries of mixed fission products, and 14 kilograms of Zircaloy-4 (cladding and hardware) into the 221-T Canyon Building. Event sequences for potential accidents that were considered included (1) leaking fuel assemblies, (2) fire and explosion, (3) loss of coolant or cooling capability, (4) dropped and/or damaged fuel assemblies, and extrinsic occurrences such as loss of services, missile impact, and natural occurrences (e.g., earthquake, tornado). Accident frequencies were determined by formal analysis to be very low. Accident consequences are greatly mitigated by the safety and containment features designed into the fuel modules and shipping cask, the long cooling time since reactor discharge, and the redundant safety features designed into the facilities, equipment, and operating procedures for the PWR Core 2 Project. Possible hazards associated with the handling of these fuels have been considered and adequate safeguards and storage constraints identified. The operations of M-160 cask unloading and module storage will not involve identifiable risks as great or significantly greater than those for comparable licensed nuclear facilities, nor will hazards or risks be significantly different from comparable past 221-T Plant programs. Therefore, it is concluded that the operations required for receipt, handling, and defueling of the M-160 cask and for the storage and surveillance of the PWR Core 2 fuel assemblies at the 221-T Canyon Building can be performed without undue risk to the safety of the involved personnel, the public, the environment or the facility

  5. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Although 14C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO2), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14C in reactor coolant. A simple chemical kinetic model predicts that CH3OH would be the initial product from radiolytic reactions of 14C following its formation from 17O. CH3OH is predicted to arise as a result of reactions of OH. with CH4 and CH3, and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH3OH can be thermally reduced to CH4 in PWR conditions, although formation of CO2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH4 is the dominant form in PWR and CO2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble material or suspended

  6. 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

  7. 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

  8. 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)

  9. 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

  10. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  11. Navy lifts veil on PWR research

    International Nuclear Information System (INIS)

    The author describes the experience of Rolls Royce in developing nuclear reactors for the Navy. Reference is made to the commissioning of HMS Sceptre in February 1978, Britain's 14th nuclear submarine. This event coincided with a decision to lift the veil somewhat on a Research and Development programme that has remained secret for nearly 20 years. Factors that have inhibited progress in this field are mentioned. One of these factors has been the high cost of marine nuclear propulsion systems, tending to limit interest to very large vessels or some special purpose craft. Another factor has been slowness to develop universally acceptable safety criteria, to allow for free and ready access of nuclear vessels to ports. A third factor has been the military origins of much of the development work. A new factor that has arisen recently is the development of the Westinghouse PWR (pressurised water reactor) for marine use in the UK. This has involved collaboration with the US Westinghouse Electric Corporation. Rolls Royce and Associates were chosen to manage this work, which is here described, including the first PWR to be designed and built in Britain and incorporated into a submarine (HMS Vulcan). Much of the design work has been concerned with development of the reactor core and increasing the endurance of the vessel between refuellings. Another aspect was less noise and vibration. Costs of this work are stated, and new test facilities are described. (U.K.)

  12. Workers doses in central European PWR NPPs

    International Nuclear Information System (INIS)

    As is stated, the ISOE database which was established in 1992 forms an excellent basis for studies and comparisons of occupational exposure data between nuclear power plants. In the year 2001, 69% of all participating reactors were pressurised water reactors. The ISOE database presents workers' exposure from 213 participating pressurised reactors (PWR) from 27 countries in that year. Among these 32 PWRs belong to six Central European Countries. The analysis of the exposure of workers based on radiation protection performance indicators (collective dose, average dose etc.) in these PWRs could be related to some nuclear safety performance indicators for recent years using ISOE database. The comparison is made to ISOE world - wide data. In the six Central European Countries altogether 32 PWR operated in the year 2001.The international databases of performance indicators related to radiation protection as for example the ISOE or the UNSCEAR database can be use as an efficient tool in the management of radiation protection of workers in a nuclear facilities and regulatory bodies. The databases enable the study of performance trends and the improvement of radiation protection. (authors)

  13. PWR and WWER thorium cycle calculation

    International Nuclear Information System (INIS)

    The first step of the investigation of the thorium fuel cycle with HELIOS 1.8 is validation of the results obtained from the code for this particular type of fuel. To complete this first task we performed calculation of the benchmark announced by IAEA in 1995. The benchmark was based on a simplified PWR model of the assembly with reduced fuel composition. This calculation was focused on a comparison of the methods and basic nuclear data. After successful validation of the code we focused our work on calculating the PWR and WWER thorium fuel cycles. The thorium cycle begins after the first use of UO2 fuel in the reactor as separation of plutonium from the burnt fuel. Separated plutonium is mixed with thorium and used as a new nuclear fuel in the reactor. For our calculation we prepared two variants of the assembly - the first variant is a homogeneous distribution and the second one is a non-homogenised distribution of thorium fuel in the assembly. The model of non-homogenised distribution of Pu-Th fuel was designed by replacing selected rods of the classical UO2 assembly by Pu-Th rods. These selected rods are distributed symmetrically in the assembly. Other rods in the assembly remain the same as in the classical UO2 assembly. The calculated and compared values are criticality and fuel composition as a function of burnup (Authors)

  14. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  15. 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.)

  16. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235U/UO2 : Pu/ThO2 : 233U/ThO2 - and the conventional recycle-mode uranium system - 235U/UO2 : Pu/UO2. The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

  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. 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)

  19. 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)

  20. Evaluation model for PWR irradiated fuel

    International Nuclear Information System (INIS)

    The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author)

  1. 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

  2. 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)

  3. 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)

  4. Subcooled decompression analysis in PWR LOCA

    International Nuclear Information System (INIS)

    The thermo-hydraulic behavior of the coolant in the primary system of a nuclear reactor is important in the core heat transfer analysis during a hypothetical loss-of-coolant accident (LOCA). The heat transfer correlations are strongly dependent on local thermo-hydraulic conditions of the coolant. The present work allows to calculate such thermo-hydraulic behavior of the coolant during subcooled decompression in PWR LOCA by solving the mass, momentum, and energy conservation equations by the method of characteristics. Detailed studies were made on the transient coolant outflow at the pipe rupture and the effect of frictional loss and heat addition to the coolant on the decompression. Based on the studies, a digital computer code, DEPCO-MULTI, has been prepared and numerical results are compared with the ROSA (JAERI) and the LOFT (NRTS) semiscale test data with various coolant pressures, temperatures, pipe break sizes, and complexity of flow geometry. Good agreement is generally obtained

  5. Recriticality risk in PWR spent fuel pools

    International Nuclear Information System (INIS)

    In this paper we investigated the situation in a PWR Spent Fuel Pool (SFP) following a long-term loss of power / loss of cooling accident. In the SFP there is a large amount of water with soluble boron between the fuel assemblies. There may be a problem from the point of view of criticality safety if the water of the SFP starts to boil and evaporate. A thermal-hydraulic analysis was performed using a simplified model of the SFP. The thermal-hydraulic analysis shows that in all cases a chaotic boiling phenomenon develops. This indicates that even if there is an issue of (near-)criticality, it will have a very intermittent nature. The multiplication factor of the SFP was evaluated with a Monte Carlo calculation. The neutronic analysis was performed for several representative cooling situations. In all cases, the system remains (deeply) subcritical. (author)

  6. Exponential experiments on PWR spent fuel assemblies

    International Nuclear Information System (INIS)

    An Exponential experiment system which is composed of a neutron detector, a signal analysis system and a neutron source, Cf-252 has been installed in order to experimentally determine the neutron effective multiplication factor for PWR spent fuel assembly. The axial background neutron flux is measured as a preliminary performance test. From the results, the spacer grid position is determined to be consistent with the design specifications within a 2.3% relative error. The induced fission neutron for four of the assemblies is also measured by scanning the neutron source, Cf-252 or the neutron detector. The exponential decay constants have been evaluated by the application of Poisson regression to the net induced fission neutron counts. It was revealed that the average exponential decay constants for the C15, J14, G23 and J44 assemblies were determined to be 0.130, 0.127, 0.125 and 0.121, respectively. (author)

  7. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

  8. Radiation embrittlement of PWR vessel supports

    International Nuclear Information System (INIS)

    Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

  9. Changes in 900 MW PWR alarm processing policy

    International Nuclear Information System (INIS)

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs

  10. 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.

  11. Analysis and study on nuclear safety of Mitsubishi PWR

    International Nuclear Information System (INIS)

    Theme of safety analysis and study are changing to reflect the needs at the time. This paper introduces the overall aspects of transient and accident analysis performed and presents typical researches related to safety analysis for Mitsubishi PWR. (author)

  12. 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)

  13. 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

  14. Overview of US research related to PWR sump clogging

    International Nuclear Information System (INIS)

    In the framework of research of researches related to the PWR sump clogging in Usa, the author presents the history of GSI-191 (assessment of debris accumulation on PWR sump performance), the research to date (technical assessment, regulatory guide and evaluation guidance, model validation), the current and planned tests (chemical effect and calcium silicate tests, latent debris and downstream effect tests, integrated chemical effect tests, EPRI coatings study). (A.L.B.)

  15. Pressure-relieving devices and it's arrangement for PWR

    International Nuclear Information System (INIS)

    There are four types of PWR pressure-relieving devices: direct acting safety valve, pilot-operated pressure relief valve, power-operated pressure relief valve and safety valve with auxiliaries. The principle of operation, characteristics, arrangement of the pressure-relieving devices for PWR recently used at home and abroad, confidence of discharge, experience in service and developing trend of the devices are introduced. The first and second type of the devices are emphasised

  16. 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)

  17. 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

  18. 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…

  19. Westinghouse advanced passive 600 MWe PWR design

    International Nuclear Information System (INIS)

    Although there has been a sharp downturn in the ordering of commercial nuclear power plants throughout the world, it is nonetheless anticipated that this form of energy will remain vital to the economy of many nations in a long term. One of the important new development activities is that of small plants incorporating passive safety features. The small plants have the merits in terms of low total capital requirement and potentially short lead time. The Electric Power Research Institute sponsored the development of an advanced LWR plant in a nine month Westinghouse program, which terminated in March, 1986. Further development at Westinghouse is now in progress on this design called AP 600 under the sponsorship of the U.S. Department of Energy. On the basis of the proven 600 MWe PWR plant design, the specific design improvement for increased safety and operational margin, reduced plant capital and operating cost, simplified plant systems and components, and increased certainty of meeting construction schedule and cost is pursued. The Westinghouse two-loop plants are very competitive, and the operating performance is outstanding by the comparison of plant capacity factor. The operation and maintenance costs are low. The specific design and the features of modification and improvement are discussed. (Kako, I)

  20. 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

  1. Computer aided information system for a PWR

    International Nuclear Information System (INIS)

    The computer aided information system (CAIS) is designed with a view to improve the performance of the operator. CAIS assists the plant operator in an advisory and support role, thereby reducing the workload level and potential human errors. The CAIS as explained here has been designed for a PWR type KLT- 40 used in Floating Nuclear Power Stations (FNPS). However the underlying philosophy evolved in designing the CAIS can be suitably adopted for other type of nuclear power plants too (BWR, PHWR). Operator information is divided into three broad categories: a) continuously available information b) automatically available information and c) on demand information. Two in number touch screens are provided on the main control panel. One is earmarked for continuously available information and the other is dedicated for automatically available information. Both the screens can be used at the operator's discretion for on-demand information. Automatically available information screen overrides the on-demand information screens. In addition to the above, CAIS has the features of event sequence recording, disturbance recording and information documentation. CAIS design ensures that the operator is not overburdened with excess and unnecessary information, but at the same time adequate and well formatted information is available. (author). 5 refs., 4 figs

  2. Aging effects in PWR power plants components

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Diogo da S.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

  3. Enriched Gadolinium as burnable absorber for PWR

    International Nuclear Information System (INIS)

    This paper is a summary of a master of thesis work in reactor physics made by Ola Seveborn. The work was done at Vattenfall Braensle AB and Ola was guided through the work by the corresponding author of this paper. The results presented are calculations for Ringhals 3, which is a Westinghouse 3-loop PWR within the Vattenfall Group. The fuel is characterized by 17x17 assemblies of AFA type containing 3.80-3.95 w/o 235U and 8 rods containing 2 w/o Gadolinium with an enrichment of 70 w/o 157Gd. The calculations were performed with the Studsvik-Scandpower code package based on the CASMO-4 lattice code and the SIMULATE-3 nodal code. The results are compared to the corresponding calculations for fuel with 5 w/o gadolinium with natural isotopic constitution. The depletion of the cores was done separately for the reference and enriched case. The results show that the gains in average for the five cycles studied are about 70 EFPH per cycle. This is an effect of the lower gadolinium content needed. Also less parasitic absorption of enriched gadolinium in the end of the fuel life contributes to the increased cycle lengths. The abruptly increased reactivity and internal power peaking factor around 10 MWd/kgU do not affect the core design negatively. (authors)

  4. 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.)

  5. Conceptual study on advanced PWR system

    International Nuclear Information System (INIS)

    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

  6. PWR core monitoring system and benchmarking

    International Nuclear Information System (INIS)

    The PWR Power Shape Monitoring System (PSMS) provides site engineers with new capabilities for monitoring and predicting core power distributions. These capabilities can lead to increased plant output as a result of greater operating margins, better load maneuvering, earlier detection of anomalies, and improved fuel reliability. The heart of the PSMS consists of nodal code (NODEP-2/THERM-P) that computes the 3-D core power distribution. This code is coupled to a simplified nodal version of the COBRA-IIIC/MIT-2 thermal-hydraulic model to determine the DNBR. These calculations can be completed in about 30 seconds on a PRIME-750 mini computer. Activation of the calculations and review of the results is through user-friendly interactive software that can be tailored to the requirements and capabilities of the different categories of users through table-driven menus. The PSMS provides unique advances over core power monitoring systems based purely on measurements. The PSMS approach permits the three-dimensional core simulation model to be routinely corrected with in-core/ex-core measurements while simultaneously identifying consistent instrument errors

  7. Tritium management in PWR fuel reprocessing plants

    International Nuclear Information System (INIS)

    Activity, quantity and nature of tritium compounds obtained during head end process (cutting and dissolution) are determined to estimate environmental release hazards in fuel reprocessing plants. Measurements on representative PWR reactor fuels (burnup 33,000 MWdt-1, specific power 30 MW dt-1) show that about 60% of the tritium produced in the reactor diffuses in the cladding where it is fixed. Remaining tritium stays in the irradiated oxide and is found as tritiated water in the solution obtained during fuel dissolution. In the UP3 plant at La Hague (France) tritiated water is disposed into the sea without environmental problems. In the case of a reprocessing plant far from the sea, the PUREX process is slightly modified for concentration of tritium in a limited amount of water (TRILEX process). It is verified experimentally in αβγ lab on actual fuel and by simulation at the pilot seale that the supplementary step ''tritium washing'' of the solvent can be obtained in pulsed columns. 4 tables, 7 figs

  8. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  9. Scaling Analysis for PWR Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yuquan [State Nuclear Power Technology R and D Center, Beijing (China); Ye, Zishen [Tsinghua University, Beijing (China)

    2011-08-15

    To test the nuclear power plant safety system performance and verify the relative safety analysis code, a widely used approach is to design and construct a scaled model based on a scaling methodology. For a pressurized water reactor (PWR), the SG scaling analysis is important before designing a scale model which is expected to well simulate the system response to the accident. In this work, a review of the transient process in SG during a loss of coolant accident (LOCA) is first presented, and then a brief natural circulation scaling analysis is performed to get the basic SG scaling design rules. The U-tube scaling design shows the scaling will enlarge the thermal center height ratio while keeping the length ratio when the scale model uses a different diameter ratio and the height ratio, which causes distortion in natural circulation simulation. And then, by the heat transfer scaling analysis, a relation between the U-tube diameter ratio and model height ratio is obtained, and it shows the diameter ratio decreases with the decreasing model height ratio. In the end, the SG transition from the heat sink to the heat source is analyzed, and the results show the SG secondary inventory and the total material heat capacity need to be properly scaled to represent the transition correctly.

  10. Scaling Analysis for PWR Steam Generator

    International Nuclear Information System (INIS)

    To test the nuclear power plant safety system performance and verify the relative safety analysis code, a widely used approach is to design and construct a scaled model based on a scaling methodology. For a pressurized water reactor (PWR), the SG scaling analysis is important before designing a scale model which is expected to well simulate the system response to the accident. In this work, a review of the transient process in SG during a loss of coolant accident (LOCA) is first presented, and then a brief natural circulation scaling analysis is performed to get the basic SG scaling design rules. The U-tube scaling design shows the scaling will enlarge the thermal center height ratio while keeping the length ratio when the scale model uses a different diameter ratio and the height ratio, which causes distortion in natural circulation simulation. And then, by the heat transfer scaling analysis, a relation between the U-tube diameter ratio and model height ratio is obtained, and it shows the diameter ratio decreases with the decreasing model height ratio. In the end, the SG transition from the heat sink to the heat source is analyzed, and the results show the SG secondary inventory and the total material heat capacity need to be properly scaled to represent the transition correctly

  11. Analysis of reactivity accidents in PWR'S

    International Nuclear Information System (INIS)

    This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

  12. 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.

  13. Aging effects in PWR power plants components

    International Nuclear Information System (INIS)

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

  14. 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

  15. 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

  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. Seismic analysis for safety related structures of 900MWe PWR NPP

    International Nuclear Information System (INIS)

    Nuclear Power Plant aseismic design becomes more and more important in China due to the fact that China is a country where earthquakes occur frequently and most of plants arc unavoidably located in seismic regions. Therefore, Chinese nuclear safety authority and organizations have worked out a series of regulations and codes related to NPP anti-seismic design taking account of local conditions. The author presents here an example of structural anti-seismic design of 90GM We PWR NPP which is comprised of: ground motion input, including the principles for ground motion determination and time history generation; soil and upper-structure modelling, presenting modeling procedures and typical models of safety related buildings such as Reactor Building, Nuclear Auxiliary Building and Fuel Building; soil-structure interaction analysis; and in-structure response analysis and floor response spectrum generation. With this example, the author intends to give an overview of Chinese practice in NPP structure anti-seismic design such as the main procedures to be followed and the codes and regulations to be respected. (author)

  18. 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)

  19. Seismic qualification of PWR plant auxiliary feedwater systems

    Energy Technology Data Exchange (ETDEWEB)

    Lu, S.C.; Tsai, N.C.

    1983-08-01

    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.

  20. 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

  1. 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

  2. The development of flow test technology for PWR fuel assemblies

    International Nuclear Information System (INIS)

    The objective of this project is to design and construct a high temperature and pressure flow test facility and to develop flow test technology for the evaluation of PWR fuel performance. For the nuclear fuel safety aspect it is of importance to evaluate the thermalhydraulic compatibility and mechanical integrity of a newly designed fuel through the design verification test. The PWR-Hot Test Loop facility is under construction to be used to perform a pressure drop test, a lift force test and a fretting corrosion test of a fullsize PWR fuel assembly at reactor operating conditions. This facility was designed to be used to produce the hydraulic parameters of the existing PWR fuel assemblies(14x14FA, 16x16FA, 17x17FA) and to verify a design of advanced fuel assemblies (KAFA-I and KAFA-II) developed by KAERI. The PWR-Cold Test Loop facility with the 5x5 Rod Bundles in the test section was designed and installed to carry out the flow distribution study by means of Laser Doppler Velocimeter. The LDV techniques have been developed and used to measure the flow velocity and turbulent intensity for evaluating mixing effects of a newly designed spacer grid with and without mixing vanes, cross flow between the fuel assemblies and a turbulent model. (Author)

  3. 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.

  4. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    International Nuclear Information System (INIS)

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

  5. Affecting factors analysis of major equipment erection key path in PWR NPP

    International Nuclear Information System (INIS)

    The affecting factors of major equipment erection in PWR NPP exist impersonally, especially the design and equipment supply has produced some effects on major equipment erection of PWR nuclear power plant. Through the analysis of key path and affecting factors on major equipment erection of PWR NPP, the paper puts forward some countermeasures. (authors)

  6. 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

  7. 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.)

  8. PWR fuel performance and future trend in Japan

    International Nuclear Information System (INIS)

    Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of RandD on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important RandD targets are the burnup extension, Gd contained fuel, utilization and the load follow capability

  9. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  10. Timing analysis of PWR fuel pin failures

    International Nuclear Information System (INIS)

    This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PF1/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design burnup. Using peaking factors commensurate with actual burnups would result in longer intervals for both reactor designs. This document provides appendices K and L of this report which provide plots for the timing analysis of PWR fuel pin failures for Oconee and Seabrook respectively

  11. 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)

  12. Load-following operation of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    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).

  13. 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)

  14. Chinese astronomy

    OpenAIRE

    Macfarlane, Alan; Cullen, Christopher

    2004-01-01

    Standing in the observatory in Beijing, Christopher Cullen discusses the nature and sophistication of Chinese astronomy in the medieval period. The political as well as the intellectual interest in astronomy is outlined.

  15. 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.

  16. Evolution of reactor monitoring and protection systems for PWR

    International Nuclear Information System (INIS)

    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.)

  17. Corrosion product transfer in PWR primary circuits during cold shutdowns

    International Nuclear Information System (INIS)

    Two experimental tests have been performed to study the corrosion product transfer during PWR cold shutdowns: one with nickel ferrite and the other one with metallic nickel. The temperature evolution together with boron and oxygen concentration evolutions are similar to those obtained during cold shutdowns of PWR primary circuits. With metallic nickel, the increase of the Ni concentration occurs during the decrease of the primary temperature and mainly when the oxygenation is realised. Whereas, with nickel ferrite, the Ni concentration increase occurs during the 24 hours after the oxygenation. These results compared to the plant data lead to conclude that metallic nickel presence in the core is the most probable hypothesis. (author)

  18. The latest full-scale PWR simulator in Japan

    International Nuclear Information System (INIS)

    The latest MHI Full-scale Simulator has an excellent system configuration, in both flexibility and extendability, and has highly sophisticated performance in PWR simulation by the adoption of CANAC-II and PRETTY codes. It also has an instructive character to display the plant's internal status, such as RCS condition, through animation. Further, the simulation has been verified to meet a functional examination at model plant, and with a scale model test result in a two-phase flow event, after evaluation for its accuracy. Thus, the Simulator can be devoted to a sophisticated and broad training course on PWR operation. (author)

  19. 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

  20. 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)

  1. Examination of dissimilar metal welds in BWR and PWR piping

    International Nuclear Information System (INIS)

    This paper addresses dissimilar metal weld examinations at PWRS. Surveys were conducted to document the dissimilar metal weld configurations at PWR plants and to update the information known about dissimilar metal weld configurations at BWR plants. The experiences which BWR utilities have had with dissimilar metal weld examinations are documented and include: correct identification of IGSCC, indications thought to be IGSCC but were actually fabrication flaws, and difficulties encountered with the examination of dissimilar metal welds after stress improvement. An experimental program was conducted which verified that the longitudinal wave procedures developed for BWRs are also applicable to PWR designs

  2. 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

  3. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Post DNB heat transfer experiments for PWR fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation (NUPEC) and Mitsubishi performed heat transfer experiments on post DNB (departure from nucleate boiling) for the pressurized water reactor (PWR) fuel assemblies under the sponsorship of the Japanese Ministry of Economy, Trade and Industry (METI) as one of a series of fuel assembly verification tests. Based on the obtained experimental data, a new evaluation model for the fuel rod heat transfer behavior after DNB was developed. A large safety margin, which had remained in the present thermal-hydraulic design that did not allow DNB, was confirmed by applying the developed model to the PWR plant safety analysis. (author)

  6. 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.

  7. Advanced ion exchange resins for PWR condensate polishing

    International Nuclear Information System (INIS)

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  8. 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)

  9. 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

  10. Methodology for the LABIHS PWR simulator modernization

    International Nuclear Information System (INIS)

    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

  11. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 2800C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 2800C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  12. Simulations of PWR spray systems by ASTEC computer code

    International Nuclear Information System (INIS)

    In this paper, sequence of loss of feedwater of steam generators on a PWR 900 MWe is performed by application of integral code ASTEC. The influence of the spray on evolution and source term of severe accident in the containment and in the environment is mainly studied. The results are helpful for the investigation of mitigative measures of severe accident. (authors)

  13. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    International Nuclear Information System (INIS)

    A statistical procedure has been used to evaluate several alternative descriptions of pressurized water reactor (PWR) cladding corrosion behavior, using an extensive database of Improved (low tin) Zr-4 cladding corrosion measurements from fuel irradiated in commercial PWRs. The in-reactor corrosion enhancement factors considered in the model development are based on a comprehensive review of the current literature for PWR cladding corrosion phenomenology and models. In addition, because prediction of PWR cladding corrosion behavior is very sensitive to the values used for the oxide surface temperatures, several models for the forced convection and sub-cooled nucleate boiling (SNB) coolant heat transfer under PWR conditions have also been evaluated. This evaluation determined that the choice of the forced convection heat transfer has the greatest impact on the ability to fit the data. In addition, the SNB heat transfer model used must account for a continuous transition from forced convection conditions to fully developed SNB conditions. With these choices for the heat transfer models, the evaluation determined that the significant in-reactor corrosion enhancement factors are related to the formation of a hydride rim at the cladding outer diameter, the coolant lithium concentration, and the fast neutron fluence (author) (ml)

  14. PWR power plant reactor maintenance: site experience and technology transfer

    International Nuclear Information System (INIS)

    PWR reactor maintenance activities has considerably expanded during the last few years at Framatome. The services offered by Framatome can be divided into three main categories: - Implementation of backfits aimed at performance and safety improvement and equipment reliability - Technical assistance for plant operators, especially during refuelling - Maintenance and repair services during both scheduled and unscheduled outages

  15. PWR auxiliary systems, safety and emergency systems, accident analysis, operation

    International Nuclear Information System (INIS)

    The author presents a description of PWR auxiliary systems like volume control, boric acid control, coolant purification, -degassing, -storage and -treatment system and waste processing systems. Residual heat removal systems, emergency systems and containment designs are discussed. As an accident analysis the author gives a survey over malfunctions and disturbances in the field of reactor operations. (TK)

  16. 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

  17. Sub Channel Thermal hydraulics Design Analysis of PWR-KSNP

    International Nuclear Information System (INIS)

    Sub channel analysis for the fuel element of thermal hydraulics design PWR-KSNP reactor has been carried out. PWR-KSNP reactor is a kind of Pressurized Water Reactor (PWR) Nuclear Power Plant developed by Korea (Korean Standard Nuclear Plant), that produce an electricity power about 1000 MWe. In the analysis, a fuel assembly with 4 fuel rods piled up into matrix 2 x 2, and surrounding by 9 sub channels of coolant, was used as a calculation model. There are 3 models of fuel assembly, i.e. the radial factors in the first model are 1.144, 1.144, 1.120 and 1.121, in the second fuel model are 0.994 , 1.005 , 0.987 and 0.989, and in the third model are 2.500, 1.144, 1.120 and 1.121, respectively. The calculated results using the COBRA IV-I code showed that the maximum cladding temperature revolved by 340.3 - 349.0 ℃, the maximum temperature of meat surface (outer of meat) revolved by 498.1 - 758.2 ℃ and the maximum temperature of meat center revolved by 928.5 - 1843.7 ℃, respectively. Whereas the safety margin against DNBR revolved by 6.50 - 2.05. By maximum meat temperature limit of 2804 ℃ and the minimum DNBR of 1.30, it is concluded that the PWR-KSNP design was in the range of safety. (author)

  18. Latest technologies of design and construction for new PWR plants

    International Nuclear Information System (INIS)

    Mitsubishi Heavy Industries, Ltd. (MHI) has so far constructed 23 PWR plants including Mihama Unit-1 Unit, which started operation in 1970. After Genkai Unit-4, which started operation in 1997, construction of PWR plants temporarily halted in Japan. However, the safety assessment of Tomari Unit-3 (3 loops, output 912 MWe) of Hokkaido Electric Power Company started in Nov. 2000, and a safety assessment of Tsuruga Units-3 and 4 (4 loops, output 1538 MWe/unit) of Japan Atomic Power Company, the first APWR units, are also planned to start very soon. PWR plants have now been designed full scale and construction of new PWR plants is going to commence in earnest in Japan. A brief introduction of the new design is followed by a description of the improved design method, upgrade of the 3D-CAD design system and improved project control procedure. The concept of the new construction method planned for the new plants is also introduced. Through these, and based on construction experience of 23 plants, MHI will introduce enhancements to the reliability and safety of the new plants and promote their smooth design and construction

  19. 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)

  20. 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)

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

    International Nuclear Information System (INIS)

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

  2. 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

  3. 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.)

  4. Thermohydraulic and constructional boundary conditions of an advanced PWR reactor

    International Nuclear Information System (INIS)

    The advantages and special features of an advanced PWR reactor (FDWR) have been systematically investigated for several years by the Department of Space Flight and Reactor Technology of the University of Brunswick (LRR-TUBS). The FDWR will have a homogeneous core, i.e. the fuel elements will consist of fuel rods of the same size and enrichment. (orig./GL)

  5. 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

  6. Process and device for residual heat removal in a PWR

    International Nuclear Information System (INIS)

    This new process for residual heat remover in a PWR is characterized by the use of the pressurizer relief tank. After cooling by the steam generator, the primary water is taken at the upper part of the pressurizer, then expanded and condensed in the pressurized relief tank. A pump recycles this water and introduces it in the primary circuit

  7. Safety criteria and their comparison between WWER and PWR

    International Nuclear Information System (INIS)

    Relevant PWR and WWER fuel safety related criteria are identified and classified into: (a) safety criteria, (b) operational criteria, and (c) design criteria. All criteria and their basis were reviewed in details. PWR and WWER fuel safety criteria are found in principle compatible (very similar if not identical basis) - thus, no fundamental distinction in the approach to defining fuel safety criteria between Eastern and Western Europe appears to exist. Some differences have been observed between the individual criteria and/or their numerical values. These are mostly due to differences in fuel design and materials. The variety in reactor characteristics and country specific licensing requirements sometimes lead to differences between criteria definitions. The report captures the characteristics of PWR/WWER fuel safety and serves as a basis to outline the general safety case for PWR and WWER fuel. The report also serves to highlight the basic fuel safety principles and their bases, and broadens the insight in failure. Especially in view of high burnup and new materials, there is a need to further develop the safety criteria and their numerical values. Even without such innovations, there is a clear potential for refining and improving some of the fuel safety related criteria and/or their numerical values. This report could serve as a basis for discussions on a more in-depth co-operation for future activities needed to verify the existing safety criteria or to support improvements. Closer collaboration in the review of PWR (LWR) and WWER safety criteria and in reviewing the respective safety cases is recommended

  8. 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

  9. PWR-GALE, Radioactive Gaseous Release and Liquid Release from PWR

    International Nuclear Information System (INIS)

    1 - Description of program or function: The PWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from pressurized water reactors (PWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment. 2 - Method of solution: GALE calculates expected releases based on 1) standardized coolant activities derived from ANS Standards 18.1 Working Group recommendations, 2) release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, 3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environs, and 4) information received on the operation of nuclear power plants. 3 - Restrictions on the complexity of the problem: The liquid release portion of GALE uses subroutines taken from the ORIGEN (CCC-217) to calculate radionuclide buildup and decay during collection, processing, and storage of liquid radwaste. Memory requirements for this part of the program are determined by the large nuclear data base accessed by these subroutines

  10. 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%

  11. PWR accident management realated tests: some Bethsy results

    International Nuclear Information System (INIS)

    The BETHSY integral test facility which is a scaled down model of a 3 loop FRAMATOME PWR and is currently operated at the Nuclear Center of Grenoble, forms an important part of the French strategy for PWR Accident Management. In this paper the features of both the facility and the experimental program are presented. Two accident transients: a total loss of feedwater and a 2'' cold leg break in case of High Pressure Safety Injection System failure, involving either Event Oriented - or State Oriented-Emergency Operating Procedures (EO-EOP or SO-EOP) are described and the system response analyzed. CATHARE calculation results are also presented which illustrate the ability of this code to adequately predict the key phenomena of these transients. (authors). 13 figs., 11 refs., 2 tabs

  12. 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

  13. The plutonium recycle for PWR reactors from brazilian nuclear program

    International Nuclear Information System (INIS)

    The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U3 O8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

  14. 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

  15. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

  16. Training of the PWR staff at EdF

    International Nuclear Information System (INIS)

    Electricite de France (EdF) is now operating thirty-two 900-MW and nine 1300-MW pressurized water reactor (PWR) units and a 1500-MW surgenerator; two 900-MW and ten 1300-MW PWR units are under construction. The number of persons employed to run these plants is at present 24000. A very important training program provided initial knowledge, specific training, and retraining in order to ensure nuclear safety. The following list includes those features required in the most viable and efficient training program: first contacts for familiarization with company and safety training (including radiation protection in nuclear plants); industrial adaptation, depending on the specialty; training for the function performed; skill maintenance and refresher courses in the function; and preparatory or promotion courses when performing a new function

  17. The modeling of flooding conditions in a PWR downcomer

    International Nuclear Information System (INIS)

    In this paper the modeling capability of the two fluid model is investigated with respect to the flooding phenomena that occur during the refill phase in a PWR experiencing a large cold leg break loss of coolant accident. A review of the literature indicate's that the basic modeling requirements have not been met and that it is still not apparent that a two fluid approach is acceptable. A modeling study is presented that attempts to investigate the applicability of a two dimensional two fluid model to the two phase hydrodynamic conditions that occur in the downcomer. A specific flow map has been constructed from experimental flooding tests on a 1/10 scale PWR and has been used to apply closure relationships appropriate for the type of flow regimes experienced during the experimental investigations. The model was validated by comparison with experimental test data and found to give good results when compared with experimental flooding curves

  18. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

  19. 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)

  20. 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)

  1. PWRDYN: a computer code for PWR plant dynamic analysis

    International Nuclear Information System (INIS)

    This report describes analytical models and calculated results of a PWR plant dynamic analysis code PWRDYN. The code has been developed in order to analyze and evaluate transient responses for small disturbance such as operating mode change and control system characteristic analysis. The features included in PWRDYN are 1) One loop approximation of primary loops, 2) Praimary coolant is always subcooled, 3) At the secondary side of steam generator is used one dimensional model and natural circulation is calculated assuming constant by positive driving head. 4) Main control systems are incorporated. In the transient responses caused by small perturbation, the calculated results by PWRDYN are in good agreement with the RETRAN calculations. Furthermore, computing time is very short so as about one seventh of real time, hence the code is convenient and useful for dynamic analysis of PWR plants. (author)

  2. Bond graph model for prediction of PWR natural circulation

    International Nuclear Information System (INIS)

    In operation of a Pressurized Water Reactor, natural circulation is an efficient, passive heat transfer mechanism for cooling. It is often employed for heat removal in operational transients, especially in long-term decay-heat removal operation. To simulate the dynamics of natural circulation, a bond graph representation of the PWR primary system and causal manipulation of the field equations has been modeled. The bond graph method calls for establishing the dynamic equations of multiport systems in state-variable form. Using the analogy of circuit elements in electrical networks, a bond graph consists of multiport capacitances, inertances, dissipations, sources of effort and flow, transformers, gyrators, and ideal junction elements. By treating each component in a PWR primary loop as a multiport element in the bond graph, a set of state-space equations representing the thermal/hydraulic responses of the loop is obtained. The state equations are then solved iteratively by using the program DYSIS developed by MIT

  3. 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

  4. Steam explosion simplified modeling. Pressure assessment in a PWR cavity

    International Nuclear Information System (INIS)

    A simplified approach of steam explosion modeling in a PWR cavity has been developed in the framework of an R and D program supported by the EDF Basic Design Division (SEPTEN). It can be used as a stand-alone engineering tool (called MGEV) or included in an integrated code (MAAP for example). This approach is mainly based on an expansion model with compressibility and inertial effects. Heat and mass transfers at the high pressure bubble interface are assessed by the EXCOBUL model and the corium mass in the interaction zone is estimated by the Park-Corradini model. The implementation aims at the calculation of pressure records on the cavity boundaries. These results will be used for mechanical evaluations which should demonstrate that the potential damages in a large PWR would be limited to the cavity region

  5. 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)

  6. 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)

  7. 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

  8. The development of 1530 MW steam turbine for Advanced PWR

    International Nuclear Information System (INIS)

    MITSUBISHI has been manufacturing 27 nuclear steam turbines and total output is over 20,000 MW since the 1970 first delivery of nuclear steam turbine. Based on these our successful experiences, MITSUBISHI is making a continuous effort to develop the most modern steam turbine with the lager capacity, higher efficiency and higher reliability. And now, the first Advanced PWR is being planned to be built at Tsuruga No.3 and No.4 by Japan Atomic Power Co. as the largest plant with an electric power of about 1530 MW. To apply this Advanced PWR plants, we are going forward planing and developing the largest capacity nuclear steam turbine. This paper shows the key technologies of target capacity nuclear steam turbine such as 54 inches low pressure last blade, and the advanced technologies to realize high performance and high reliability steam turbine. (author)

  9. Polynomial parameterized representation of macroscopic cross section for PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fiel, Joao Claudio B., E-mail: fiel@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

    2015-07-01

    The purpose of this work is to describe, by means of Tchebychev polynomial, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and {sup 235} U {sub 92} enrichment. Analyzed cross sections are: fission, scattering, total, transport, absorption and capture. This parameterization enables a quick and easy determination of the problem-dependent cross-sections to be used in few groups calculations. The methodology presented here will enable to provide cross-sections values to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by parameterized cross-sections functions, when compared with the cross-section generated by SCALE code calculations, or when compared with K{sub inf}, generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)

  10. Assessment of subcriticality during PWR-type reactor refueling

    International Nuclear Information System (INIS)

    During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-α and Feynman-α methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

  11. 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

  12. 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)

  13. 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)

  14. Training of PWR operators in Electricite de France

    International Nuclear Information System (INIS)

    Electricite de France has started up in 1981 eight PWR units (900 MWe) and five or six more will be commissioned in the course of each of the next few years. Such a programme will require an unprecedented effort of recruitment and training by the company. New staff will be recruited partly by taking on young school-leavers and partly by internal recruitment within EDF of staff working in conventional thermal or nuclear power stations. Training will be provided each year to around 4,000 staff. The paper reviews successively the following points: organizing the operation of a power station with four PWR units (900 MWe); the criteria for personnel selection; training programmes suited to the origin and function of operating personnel; training staff to carry out a given function; the existing training resources

  15. Calculation of drop course of control rod assembly in PWR

    International Nuclear Information System (INIS)

    The validation of control rod drop performance is an important part of safety analysis of nuclear power plant. Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line. Based on structural features of the drive line, the driving force on moving assembly was analyzed and decomposed, the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method, and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved. The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR. It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA. (authors)

  16. 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

  17. 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

  18. 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.)

  19. Fire experiences: principal lessons learned, application in PWR power plants

    International Nuclear Information System (INIS)

    The article reviews the principal design rules to be borne in mind for PWR nuclear units installation. These rule takes into account: the specific character of materials involved (safety aspect for nuclear construction), experience acquired as a result of fires in EDF production units, and the results obtained from tests carried out by the EDF at Fort de Chelles between 1980 and 1982, especially in the field of PVC cables

  20. DNB experiments for high-conversion PWR core design

    International Nuclear Information System (INIS)

    It is very important to clarify the departure from nucleate boiling (DNB) performance of core fuel assemblies for the high conversion pressurized water reactors (PWR). To investigate this, DNB experiments were performed in tight lattice rod bundles, using the model fluid Freon 12 and water under the actual operating conditions. In addition, DNB heat flux measurements in an annular-flow channel were carried out for the design of the fertile rods, which are installed in thimble tubes. (orig.)

  1. DNB experiments for high conversion PWR core design

    International Nuclear Information System (INIS)

    It is very important to clarify the departure from nucleate boiling (DNB) performance of core fuel assemblies for the high conversion PWR design. To investigate this, DNB experiments were performed in tight lattice rod bundles, using the model fluid freon-12 and the actual water. And also DNB heat flux mesurements in an annular flow channel were carried out for design of fertile rods which are installed in thimble tubes. (orig.)

  2. 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

  3. Ciclon: A neutronic fuel management program for PWR's consecutive cycles

    International Nuclear Information System (INIS)

    The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author)

  4. SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System

    International Nuclear Information System (INIS)

    1 - Description of program or function: SACHET evaluates the dynamic fission product inventories in the multiple compartment system of pressurized water reactor (PWR) plants. 2 - Method of solution: SACHET utilizes a matrix of fission product core inventory which is previously calculated by the ORIGEN code. 3 - Restrictions on the complexity of the problem: Liquid wastes such as chemical waste and detergent waste are not included

  5. Safety philosophy and concepts for future European PWR

    International Nuclear Information System (INIS)

    PWR are presently the most developed nuclear technology in the world and particularly in Europe. The existing units have shown during twenty years of operation a high level of reliability and safety. Despite the remarkable result already achieved the necessity of regaining public confidence leads nuclear industry to look for further safety improvements. The european utilities, convinced of the necessity of future nuclear capacity investments, have decided to unify their efforts for the preparation of the next nuclear generation

  6. 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.

  7. A practical method for optimization of fuel management of PWR

    International Nuclear Information System (INIS)

    A practical method for simulation of fuel management optimization of PWR cores with two-dimensional model is described. The general objective of the optimization is to choose a set of refuelling arrangement schemes, which will produce the maximum economic profit on condition that it meets the safety criteria of PWR. It oftern requires quite a lot of computer time to simulate the optimized schemes. An effective and acceptable optimization strategy, two-step search method, has been developed. The first step of algorithm consists of several approaches based on the information avilable and the past experiences with refuelling. The second step allows a further improvement of the previously determined optimum schemes. The maximum radial power peaking factor, Wp, is defined as the objective function. Several physical criteria are examined to propose the constraints. The main intention is to minimize, the objective function Wp, subjected to various constraints. Hence a computer program, 2DFEOF in FORTRAN 77, was developed. Some calculations were done for a typical PWR core on an IBM-4341 computer. The satisfactory results were obtained at reasonable low computational costs. It spent nearly 9 mins CPU time for 3 fuel cycles with a 1/8 core configuration

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

    Science.gov (United States)

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    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.

  9. 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)

  10. 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

  11. Design study of long-life PWR using thorium cycle

    Science.gov (United States)

    Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

    2012-06-01

    Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that 231Pa better than 237Np as burnable poisons in thorium fuel system. Thorium oxide system with 8% 233U enrichment and 7.6˜ 8% 231Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1% Δk/k and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53% Δk/k and reduced power peaking during its operation.

  12. 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.

  13. 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

  14. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

  15. 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.

  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. Analysis of the estimated isotopic concentration of PWR spent fuel

    International Nuclear Information System (INIS)

    Using SCALE4.4 SAS2H, HELIOS and CASMO codes, isotopic inventories in PWR spent fuel have been calculated and compared with the reported experimental data. 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. Influences of correction factors to the multiplication factor have also been investigated. The calculated biases and uncertainties of U-235 in PWR spent fuel seem to be 2.8 % / 3.9 %, -2.0 % / 4.1 % and 5.0 % / 4.5 %. In the case of transuranium isotopes and fission products, the results calculated by HELIOS and CASMO codes show a large discrepancy from the reported experimental data in comparison of SAS2H results. In general it is believed that SAS2H is better than HELIOS and CASMO for estimating isotopic inventory in PWR spent fuel. It is revealed that correction factors obtained by codes of interest give rise to the maximum difference of about 0.05 in the multiplication factor

  18. Study on virtual simulation technology for operation and control of PWR

    International Nuclear Information System (INIS)

    The way to build graphical models of PWR with MultiGen Creator is discussed, and the three-dimensional model used in the virtual simulation is built. The mathematical simulation model for PWR based on the platform of MFC and Vega is built through the analysis of the mathematical simulation of PWR. The way to perform the virtual effect is introduced associating with the Pressurizer. And, all above parts are connected in one with VC++ to perform the whole virtual simulation of PWR. (authors)

  19. 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 ...

  20. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

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

  3. Computer analyses on loop seal clearing experiment at PWR PACTEL

    International Nuclear Information System (INIS)

    Highlights: • Code analyses of loop seal clearing experiment with PWR PACTEL are introduced. • TRACE and APROS system codes are used in the analyses. • Main events of the experiment are well predicted with both codes. • Discrepancies are observed on the secondary side and in the core region. • Loop seal clearing phenomenon is well simulated with both codes. - Abstract: Water seal formation in the loop seal in pressurized water reactors can occur during a small or intermediate break loss-of-coolant accident, causing temporary fuel overheating. Quantification of the accuracy of overheating prediction is of interest in the best-estimate safety analyses, even though the peak cladding temperatures due to the water seal formation in the loop seal seldom approach acceptance criteria as such. The aim of this study was to test and evaluate the accuracy with which the thermal–hydraulic system code nodalizations of the PWR PACTEL predict loop seal clearing in a small break loss-of-coolant-accident test performed with the PWR PACTEL facility. PWR PACTEL is a thermal–hydraulic test facility with two loops and vertical inverted U-tube steam generators. Post-test simulations were performed with the TRACE and APROS system codes. In the post-test simulations, the main events of the transient such as the decrease in the core water level, depressurization of the primary circuit, and the behavior of the water seal formation and clearing in the loop seal were predicted satisfactorily by both codes. However, discrepancies with the experiment results were observed in the analyses with both codes, for example the core temperature excursions were halted too early and the peak temperature predictions were too low. The core water level increase caused by loop seal clearing was overestimated with both codes, and the pressure and temperature were overestimated on the secondary side of the steam generators. Loop Seal 2 was evidently cleared out while Loop Seal 1 remained closed

  4. The Chinese Banking System

    OpenAIRE

    Grant Turner; Nicholas Tan; Dena Sadeghian

    2012-01-01

    The Chinese banking system is critical to the functioning of the Chinese economy, being the main conduit through which savings are allocated to investment opportunities. Banking activity in China has grown rapidly over the past decade in association with the expansion of the Chinese economy, and the Chinese banking system now includes some of the world’s largest banks. Chinese banks have become more commercially orientated over this period, although the Chinese Government retains considerable...

  5. 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)

  6. 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)

  7. The evaluation of erosion-corrosion problems in Taiwan PWR carbon steel piping

    International Nuclear Information System (INIS)

    Taiwan PWR Nuclear Power Plant Units 1 and 2 implemented the projects of Pipe Wall Thinning Measurement under the request of ROCAEC to prevent the events due to the piping erosion/corrosion. The purpose of this paper is to present the improvements in the evaluation method for the identification of the potential piping systems and components in PWR

  8. PREP-PWR-1.0: a WIMS-D/4 pre-processor code for the generation of data for PWR fuel assemblies

    International Nuclear Information System (INIS)

    The PREP-PWR-1.0 computer code is a substantially modified version of the PREWIM code which formed part of the original MARIA System (Report J.E.N. 543). PREP-PWR-1.0 is a comprehensive pre-processor code which generates input data for the WIMS-D/4.1 code (Report PEL 294) for PWR fuel assemblies, with or without control and burnable poison rods. This data is generated at various base and off-base conditions. The overall cross section generation methodology is described, followed by a brief overview of the model. Aspects of the base/off-base calculational scheme are outlined. Additional features of the code are described while the input data format of PREP-PWR-1.0 is listed. The sample problems and suggestions for further improvements to the code are also described. 2 figs., 2 tabs., 12 refs

  9. Integral type small PWR with stand-alone safety

    Energy Technology Data Exchange (ETDEWEB)

    Makihara, Yoshiaki [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2001-09-01

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

  10. 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)

  11. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    International Nuclear Information System (INIS)

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial pressurized water reactor (PWR) assembly enrichment that would permit loading of spent nuclear fuel into the 21 PWR waste package with absorber plates design as provided in Attachment IV. This calculation is an example of the application of the methodology presented in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent U-235, and a burnup range of 0 through 45 GWd/MTU. Higher burnups were not necessary because 45 GWd/MTU was high enough for the loading curve determination. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing PWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials. This calculation is subject to the ''Quality Assurance Requirements and Description'' (QARD) (DOE 2004) because it concerns engineered barriers that are included in the ''Q-List'' (BSC 2004k, Appendix A) as items important to safety and waste isolation

  12. 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)

  13. Effect of startup ramp rate on PWR fuel reliability

    International Nuclear Information System (INIS)

    A wide range of startup strategies and restart times currently exists for commercially operated pressurized water reactors (PWRs). The variability in PWR restart strategies is a function of several factors, including reactor system instrument calibration, primary and secondary water chemistry control, and vendor specified fuel rod ramp rate limitations. Fuel vendors, as a means to mitigate pellet-cladding interaction (PCI) leading to fuel rod failures, specify reactor power ramp rate limitations following a refueling outage. Typical restart ramp rates range between 3% per hour and 4% per hour of full reactor power above a threshold reactor power level between 20% and 40% full power. This paper summarizes an analytical evaluation performed to assess the technical basis for PWR restart ramp rate restrictions and to provide the technical justification to propose less restrictive power ramp rate conditions. Two combinations of PWR reactor types (Yonggwang Unit 2 and 4) and fuel rod designs were used to evaluate the impact of ramp rate and threshold power conditions on the PCI behavior of once-burned and twice-burned fuel rods. The fuel rod condition at the reactor restart of interest was established using the ESCORE steady state fuel performance program. Detailed PCI calculations were performed using the FREY fuel rod behavior program. The assessment identified significant margin to PCI failure for current ramp rate conditions used in YGN Unit 2 and 4. Based on the analytical evaluation presented, ramp rates up to 5% per hour above threshold power levels up to 60% of full reactor power can be used without concern for fuel rod integrity during reactor restarts following a refueling outage

  14. Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

    International Nuclear Information System (INIS)

    An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U3O8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U3O8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author)

  15. Short-term calculations to supplement the RS 16 B PWR experiments with internals (PWR1 to PWR5), using the LECK 4 computer code

    International Nuclear Information System (INIS)

    Within the framework of research project RS 16 B sponsored by the German BMFT a series of a blowdown experiments, DWR1 to DWR5, were performed using a vessel with dummy internals under conditions similar to those in a PWR. The prime objective of these experiments was the investigation of the highly transient blowdown phenomena in the discharge nozzle and the determination of the induced loads on the internals. As a partner in the project, KWU carried out both pre-test predictions and post-test analyses of these experiments using, among others, the computer code LECK 4. For the most severe blowdown test DWR5, the influence of the most important model parameters on the blowdown analysis was investigated in detail. These investigations suggest that, similar to the long-term analyses, calculations using the homogeneous critical flow model would improve agreement between calculation and experiment. (orig./RW)

  16. Nuclear power perspective in China

    International Nuclear Information System (INIS)

    China started developing nuclear technology for power generation in the 1970s. A substantial step toward building nuclear power plants was taken as the beginning of 1980 s. The successful constructions and operations of Qinshan - 1 NPP, which was an indigenous PWR design with the capacity of 300 MWe, and Daya Bay NPP, which was an imported twin-unit PWR plant from France with the capacity of 900 MWe each, give impetus to further Chinese nuclear power development. Now there are 8 units with the total capacity of 6100 MWe in operation and 3 units with the total capacity of 2600 MWe under construction. For the sake of meeting the increasing demand for electricity for the sustainable economic development, changing the energy mix and mitigating the environment pollution impact caused by fossil fuel power plant, a near and middle term electrical power development program will be established soon. It is preliminarily predicted that the total power installation capacity will be 750-800GWe by the year 2020. The nuclear share will account for at least 4.0-4.5 percent of the total. This situation leaves the Chinese nuclear power industry with a good opportunity but also a great challenge. A practical nuclear power program and a consistent policy and strategy for future nuclear power development will be carefully prepared and implemented so as to maintain the nuclear power industry to be healthfully developed. (author)

  17. 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.)

  18. Vertical Drop Of 21-PWR Waste Package On Unyielding Surface

    International Nuclear Information System (INIS)

    The objective of this calculation is to determine the structural response of a 21-PWR (pressurized-water reactor) Waste Package (WP) subjected to the 2-m vertical drop on an unyielding surface at three different temperatures. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities in two different WP components. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only

  19. Evaluation of a PWR assembly subjected to air-storage

    International Nuclear Information System (INIS)

    Most dry storage tests and demonstrations involving Zircaloy-clad fuel have been conducted in inert cover gases. The availability of irradiated PWR fuel and an instrumented dry storage test module (Fuel Temperature Test or FTT Facility) at the Engine Maintenance and Disassembly (EMAD) site in Nevada prompted a dry storage test in air. Not included in the prior publication are results of the post-test fuel examination. To date, that examination has involved visual inspection, photography, and analysis of swipes taken from fuel assembly surfaces. This paper summarizes results of the examination and an overview of the test interpretation and significance

  20. Probabilistic consequence analysis of ATWSs in a PWR plant

    International Nuclear Information System (INIS)

    PWR responses (in terms of overpressures, DNBR and other safety-related quantities) to ATWSs are being probabilistically investigated by applying response surface methodology to ALMOD, a computer program for simulation of large amplitude transients. The reactor considered for the analysis is the 1300 MWel reference KWU reactor plant. A comprehensive set of input quantities--including operational, engineering and physical variables--is taken into account. Results are presented for the first phases of station-blackout and loss-of-heat sink ATWSs

  1. 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)

  2. 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.

  3. Application of H∞ control theory to PWR power control

    International Nuclear Information System (INIS)

    In this paper, a robust controller is designed by the use of H∞ control theory for the PWR power control. The design specification is incorporated by the frequency weights using the mixed-sensitivity problem. The robustness of H∞ control is verified by comparing with the classical output feedback control and LQG control in the case of measurement delay of the power measurement system. The H∞ optimal control shows excellent stability-robustness and performance-robustness for external disturbances and noises, model parameter variations, and modeling errors. It also provides a practical design method because the design specification can be easily implemented

  4. 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)

  5. Design build process flow visualization model plant PLTN PWR type

    International Nuclear Information System (INIS)

    Scale-down version of nuclear power plant type PWR model and process flow visualization has been design and constructed. This scale-down model includes primary and secondary cooling systems, and transmission line in three dimensional layout with a 1: 33,33 scale. The construction of scale model has been done in five steps that are study literature, field survey, drawing scale design, construction, and test. The results is scale-down model integrated with monitoring system using lab view and interlock system using PLC. The test result shows that process flow has operated as required in design specification. (author)

  6. Friends of the Earth case against the PWR

    International Nuclear Information System (INIS)

    Friends of the Earth's case against Sizewell B has been summarised in a report entitled 'Critical Decision: Should Britain buy the Pressurised Water Reactor?'. This showed that on economic and safety grounds, Sizewell B would not be a good choice for the electricity consumer or the country at large. Events since the end of the Inquiry, particularly those affecting the economic case, have confirmed this conclusion. This paper will summarise the case, both during the Inquiry and subsequent to this, as well as make reference to the long-term environmental implications of the Central Electricity Generating Board's PWR programme. (author)

  7. 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

  8. 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)

  9. 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.)

  10. 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

  11. Neutronal aspects of PWR control for transient load following

    International Nuclear Information System (INIS)

    The purpose of this thesis is to qualify the CRONOS diffusion code on a load transient in grey mode control. First of all, we have established a general axial calculational model and studied the important physical phenomena: xenon oscillation, grey rods absorption, radial leaks modelling, effect of the initial conditions in Iodine and Xenon. In a second stage, a three dimensional calculation has been performed, the results of which have been compared to a PWR 900 TRICASTIN 3 experiment and have been in good agreement. In the last part, we show that the results of the axial model using one-dimensional CRONOS calculations are quite consistent with the three-dimensional calculation

  12. Quality surveillance for PWR power plant reactor internals manufacturing

    International Nuclear Information System (INIS)

    The structure and the function of the reactor internals of the improved generation Ⅱ PWR power plant is instructed briefly, the critical factors and difficulties in the manufacture process for reactor internals are analyzed, the quality control and surveillance of reactor internals manufacturing is discussed, especially the critical factors and difficulties of the quality control in the manufacture process for the main parts of reactor internals and in the reactor internals assembling process are represented in detail, the key points of the resident manufacture supervision is presented, and other key points of quality control in the manufacture process are also given, such as the documents control and personnel control. (author)

  13. For sale: 7 AGR stations and a brand new PWR

    International Nuclear Information System (INIS)

    Britain's seven AGR stations and the Sizewell B PWR will pass to private ownership under the UK government's plan to privatise the two nuclear generators, Nuclear Electric and Scottish Nuclear, sometime next year. Under the new set-up, the two generators will become operating subsidiaries of a holding company which will be headquartered in Scotland. The companies' ageing Magnox gas-cooled reactors will remain in a separate public sector company before being transferred to British Nuclear Fuels (BNFL) at the time of privatisation. (author)

  14. Concept of safety systems for next generation PWR (APWR+)

    International Nuclear Information System (INIS)

    The concept of the next generation PWR, which is expected to come after the APWR and is named the APWR+, is being studied, considering that the light water reactors are seemed to be dominant also in the 21st century. The APWR+ is designed to have the features of four-train safety systems, divergent emergency electrical sources, and passive core cooling system using steam generators at early stage of the Loss of Coolant Accident. The basic concept has been made, and more detailed investigation is scheduled in near future. (author)

  15. Recent progress in SG level control in French PWR plants

    International Nuclear Information System (INIS)

    Controlling the steam generator (SG) level is of major importance in a large PWR plant. This has led to extensive work on SG computer models. This paper presents results of the comparison between calculations and tests on the first four-loop plant in France. Four-loop plants started up after 1985 will be equipped with digital instead of analog controllers. A new SG level control has been designed and then optimised using the validated SG model. A prototype of this new system has been successfully tested on a three-loop plant. 4 refs

  16. 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

  17. Improvement in PWR flexibility the french program 1975-1995

    International Nuclear Information System (INIS)

    Between 1975 and 1985, a substantial effort was launched in France to greatly improve PWR's flexibility, resulting in the current situation where both frequency control and load follow are now routinely performed on most plants in operation. Based on rapidly accumulating operational experience and on all expertise acquired in the past decade, a second-generation core control strategy is now being finalized for application on all future 1400 MW plants (with commercial operation scheduled in 1992 for first unit). This 20-year program is discussed

  18. Optimization of control area ventilation systems for Japanese PWR plants

    International Nuclear Information System (INIS)

    The nuclear power plant has been required to reduce the cost for the purpose of making the low-cost energy since several years ago in Japan. The Heating, Ventilating and Air Conditioning system in the nuclear power plant has been also required to reduce its cost. On the other hand the ventilation system should add the improvable function according to the advanced plant design. In response to these different requirements, the ventilation criteria and the design of the ventilation system have been evaluated and optimized in Japanese PWR Plant design. This paper presents the findings of the authors' study

  19. 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)

  20. 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, ...

  1. 21-PWR Waste Package Side and End Impacts

    International Nuclear Information System (INIS)

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1

  2. PWR fuel performance and burnup extension programme in Japan

    International Nuclear Information System (INIS)

    Since the first PWR nuclear power plant Mihama Unit 1 initiated commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts on improving the technology of PWRs. The results can already be seen by the significantly improved performance of the PWR plants now in operation. Mitsubishi Heavy Industries, Ltd supplied the nuclear fuel assemblies, which now amount to almost 5000. Although some trouble with fuel was experienced in the beginning, the progressive efforts made to improve the fuel design and manufacturing technology have resulted in the superior performance of Mitsubishi fuels. Since fuel of current design should comply with the limitation set in Japan for a maximum discharged fuel assembly average burnup of less than 39,000 MW·d/t, the maximum burnup is now around 37,000 MW·d/t. However, an increase in this burnup limitation has been strongly requested by Japanese utilities in order to make nuclear power more economic and thus more competitive with other power generation methods. A summary is given of the design improvements made on Mitsubishi fuel, as well as demonstration programmes of current design fuel to prove its superior reliability and to prepare the database for a future extension of burnup. (author)

  3. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    International Nuclear Information System (INIS)

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M and O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated

  4. 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.

  5. 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)

  6. Modifications needed to operate PWR's plants in G-Mode

    International Nuclear Information System (INIS)

    The production of electricity from PWR nuclear plants represents 44% of the total production of electricity in France for 1984, and 68% of the electricity produced by Thermal power plants (127 TWh over 187 TWh). These data show clearly that the French PWR plants do not work in ''base mode'' anymore but have to fit production with consumption, in other words to assume the frequency control. To participate permanently to the load follow and frequency control, it appeared that some improvements in the field of pressurizer level and pressure control were necessary as well as in the field of operator aids computer. It should be noted that these improvements are useful even without taking into account the constraints due to load follow and frequency control because of the mechanical stress in the CVCS piping, for instance. Some additional tests are planned to better identify this specific problem. The need of a more flexible operating mode than ones given by the initial system (black control rods), significantly reduced in 1973 due to the application of the ECCS criterion, led EDF and Framatome to develop a new operating mode (G. Mode) allowing a faster power escalation (5% PN/mn) whatever the fuel burn-up. This new operating mode improves significantly also the flexibility of operation when the frequency control is needed, and helps a lot the operators in such cases. All the 900 MWe Nuclear plants will be able to operate in ''G mode'' before the end of 1984

  7. 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)

  8. 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)

  9. DNB analysis with mechanistic models for PWR fuel assemblies

    International Nuclear Information System (INIS)

    In order to predict the DNB heat flux of PWR fuel assemblies and the critical power of BWR fuel bundles, the Boiling Transition Analysis Code CAPE' has been developed in the IMPACT project. The CAPE code for PWR includes three analysis modules, subchannel analysis module, three-dimensional two-phase flow analysis module, and DNB evaluation module. The subchannel module uses drift-flux model to identify the hottest subchannel. The three-dimensional two-phase flow analysis module uses nonhomogeneous and nonequilibrium two fluid model to analyze the detailed three-dimensional two-phase flow behaviors such as void distribution. For DNB heat flux prediction, the DNB evaluation module uses the Weisman model in which is a mechanistic DNB evaluation model. This paper describes the analysis models, analysis techniques and the results of validation by rod bundle test analysis. To date, the average difference between calculated and 11 measured values was -0.6% with a standard deviation of 7.0%. (author)

  10. 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

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

    International Nuclear Information System (INIS)

    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

  12. 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

  13. 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

  14. 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)

  15. 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.

  16. Thermal-hydraulic model verification calculation of PWR tests

    International Nuclear Information System (INIS)

    Test PWR 5 determined forces and pressure differences across the reactor pressure vessel and the internals in the test vessel during the first 80 ms by means of the present test data tapes. Furthermore a comparison between the measured data and those determined with the aid of the LECK program system is carried out. The following results were obtained in this connection: The qualitative pattern as compared between calculation and measurement shows a good agreement. Higher pressure differences resulted across the components due to the higher pressure gradients in the initial phase of the blowdown verification in the calculations. The best agreement of the pressure gradients was obtained with the verification calculations for a rupture opening time of 6 ms. Since there was no fluid/structural-dynamic coupling it was not possible to simulate the premature pressure reduction within the core barrel. The distribution of the initial temperature in the calculation did not always agree with that during the test. (orig.)

  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. Optimization of chemistry in PWR and VVER nuclear power plants

    International Nuclear Information System (INIS)

    This paper, based on international feedback and studies, proposes potential improvements for PWR and VVER operation: pH optimization in the primary coolant in order to minimize corrosion product transport/deposition and associated radiation exposure, crud induced power shifts (previously called axial offset anomaly), and fuel failure; use of enriched boron acid (enriched with 10B) to easily optimize the above described pH, particularly with the increased use of higher fuel enrichments; zinc addition in the reactor cooling system; establishment of secondary water chemistry specifications which take into consideration the steam generator tubing materials and design to minimize corrosion risk while keeping sufficient plant availability and decreasing environmental impact; amine selection for the secondary system aimed at mitigating steam generator tube fouling, power loss and maintenance costs as well as corrosion risks; overall operating chemistry options designed to minimize environmental impact, such as elimination of condensate polishers and optimum ion exchange resin use. (orig.)

  19. Optimization of chemistry in PWR and VVER Nuclear Power Plants

    International Nuclear Information System (INIS)

    Potential improvements for PWR and VVER operation are proposed: pH optimization in the primary coolant in order to minimize corrosion product transport/deposition and associated dosimetry, Crud Induced Power Shifts (CIPS, previously named AOA), fuel failure; use of EBA (enriched Boron 10) to easily optimize above described pH, particularly with the larger use of higher fuel enrichments; zinc addition in the RCS; secondary water chemistry specifications depending on the steam generator tubing materials and design for minimizing corrosion risk while keeping a sufficient plant availability and decreasing environmental impacts; amine selection for the secondary system for mitigating steam generator tube fouling, power loss and maintenance costs as well as corrosion risks; overall operating chemistry options to minimize environmental impacts, such as elimination of condensate polishers, optimum ion exchange resin use. (N.T.)

  20. 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

  1. 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.)

  2. 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.

  3. 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)

  4. Siemens Nuclear Power Corporation experience with BWR and PWR fuels

    International Nuclear Information System (INIS)

    The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

  5. 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

  6. Ultrasonic Backscattering in Polycrystalline Materials of Pwr Components

    Science.gov (United States)

    Chassignole, B.; Dupond, O.; Fouquet, T.; Rupin, F.

    2011-06-01

    The ultrasonic examination of metallic components of Pressurized Water Reactors (PWR) is an important challenge for the nuclear industry. During the past decades, EDF R&D has undertaken numerous studies in order to improve the NDT process on these applications and to help to their qualification. The present paper deals with the problem of the structural noise which can potentially disturbs the ultrasonic inspection. In particular, this study proposes a modeling approach to simulate the ultrasonic scattering due to coarse grain structures of polycrystalline materials. The methodology is based on the mixing of a grain scale description of the material and a 2D finite element code (ATHENA) developed by EDF to simulate the ultrasonic propagation in isotropic and anisotropic elastic media. The modeling results are compared to experimental acquisitions on mock-ups containing artificial defects.

  7. MAAP/PWR comparisons to other codes and plant data

    International Nuclear Information System (INIS)

    As part of a simulator qualification effort, a cooperating utility assembled an extensive collection of RETRAN and RELAP calculations and plant transient data for a four loop Westinghouse pressurized water reactor (PWR). This information was provided and formed the basis for a relatively complete set of MAAP comparison calculations. The calculations performed include actual loss of offsite power; actual reactor trip; reactor trip calculation with simplified control system operation; failed power operated relief valve; total loss of feedwater with feed and bleed; main steam line break; steam generator tube rupture; steam generator tube rupture with stuck-open steam line PORV; small loss of coolant accident (LOCA); and small local with failure of safety injection systems. The results obtained were fairly good

  8. Review of PWR-related thermal-shock studies

    International Nuclear Information System (INIS)

    Flaw behavior trends associated with pressurized-thermal-shock (PTS) loading of PWR pressure vessels have been under investigation at ORNL for approx.12 years. During that time, eight thermal-shock experiments with thick-walled steel cylinders were conducted as a part of the investigations. These experiments demonstrated, in good agreement with linear elastic fracture mechanics (LEFM), crack initiation and arrest, a series of initiation-arrest events with deep penetration of the wall, long crack jumps without significant dynamic effects at arrest, arrest in a rising K/sub I/ field, extensive surface extension of an initially short and shallow flaw, and warm prestressing with K/sub I/ equal to or less than 0. This information was used in the development of a fracture-mechanics model that is being used extensively in the evaluation of the PTS issue

  9. Shutdown Chemistry Process Development for PWR Primary System

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  10. Evaluation of fire probabilistic safety assessment for a PWR plant

    International Nuclear Information System (INIS)

    The internal fire analysis of the level 1 power operation probability safety assessment (PSA) for Maanshan (PWR) Nuclear Power Plant (MNPP) was updated. The fire analysis adopted a scenario-based PSA approach to systematically evaluate fire and smoke hazards and their associated risk impact to MNPP. The result shows that the core damage frequency (CDF) due to fire is about six times lower than the previous one analyzed by the Atomic Energy Council (AEC), Republic of China in 1987. The plant model was modified to reflect the impact of human events and recovery actions during fire. Many tabulated EXCEL spread-sheets were used for evaluation of the fire risk. The fire-induced CDF for MNPP is found to be 2.1 E-6 per year in this study. The relative results of the fire analysis will provide the bases for further risk-informed fire protection evaluation in the near future. (author)

  11. PWR fuel management optimization using continuous particle swarm intelligence

    International Nuclear Information System (INIS)

    The objective of nuclear fuel management is to minimize the cost of electrical energy generation subject to operational and safety constraints. In the present work, a core reload optimization package using continuous version of particle swarm optimization, CRCPSO, which is a combinatorial and discrete one has been developed and mapped on nuclear fuel loading pattern problems. This code is applicable to all types of PWR cores to optimize loading patterns. To evaluate the system, flattening of power inside a WWER-1000 core is considered as an objective function although other variables such as Keff along power peaking factor, burn up and cycle length can be included. Optimization solutions, which improve the safety aspects of a nuclear reactor, may not lead to economical designs. The system performed well in comparison to the developed loading pattern optimizer using Hopfield along SA and GA.

  12. Reactor building seismic analysis of a PWR type - NPP

    International Nuclear Information System (INIS)

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  13. 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)

  14. Intelligent main control room for advanced PWR plants

    International Nuclear Information System (INIS)

    The design targets of the main control room of nuclear power plants are as follows. (1) To make a good working environment where operators can operate easily. (2) To reduce the work load and operators error. To this end, MHI has been improving main control room design for advanced PWR plants. The new intelligent main control room consists of a soft operation console and a large display panel. According to our evaluation, the work load and human error of the new main control room are reduced by about 35% compared with the latest plants. This new design will be used to plan new plants and will have the additional feature of saving costs by standardizing plant design. (author)

  15. Conceptual design of SPWR, a PWR with enhanced passive safety

    International Nuclear Information System (INIS)

    A conceptual design has been carried out on a new type of integrated pressurized water reactor, SPWR (System-integrated PWR). This reactor installs a poison tank (borated water filled) in the reactor vessel instead of control rod drive system. Three hydraulic pressure valves are installed as the upper interface between the poison tank and primary coolant. A 700MWe power plant with twin 1100MWt SPWRs which are installed in a reactor building has been studied. Design and analysis have been made on the reactor core, reactor (reactor vessel, steam generator, main circulating pump, pressurizer, poison tank and their integration), plant systems (main and sub systems), layout, construction scheme, operation and maintenance, safety related components, reactor dynamics, economics and R and D needs. Passive safety features are also studied. (author)

  16. Residual heat removal in a PWR using a passive system

    International Nuclear Information System (INIS)

    The present work is made in the frame of the studies that are performed at the French Atomic Energy Commission on the innovative safety systems. The system which is discussed here is devoted to the residual heat removal. It can be used for a current french 3 loops PWR in place of the combination auxiliary feedwater system - atmospheric relief valve. A blackout transient, without auxiliary feedwater, is calculated, using the CATHARE code, in order to assess the capabilities of the system. Some complementary scenarios are calculated, assuming the intervention of other systems after a while, for example restart of the primary pumps and manual opening of the atmospheric relief valves. The influence of non condensable gases is also discussed. 7 refs., 17 figs

  17. Fuzzy logic control for PWR load-follow

    International Nuclear Information System (INIS)

    The developments of fuzzy logic control theory promote the application of fuzzy logic controller to load-follow in Pressurized Water Reactors. A control law combined fuzzy logic controller to load-follow in Pressurized Water Reactors. A control law combined fuzzy logic controller with conventional PID controller, using the strategy of output gains varying with nuclear reactor power, was proposed to control load-follow operations in PWR. This method solves the nonlinear time-varying close-loop control problem and overcomes the shortcomings and limitations of the model-based method. The simulation results show the method is of both satisfactory dynamic performance and high steady state precision. This approach will improve the automaticity of load-follow operations

  18. Development of a general nodalization scheme for PWR simulators

    International Nuclear Information System (INIS)

    The paper deals with the development of four nodalizations of PWR simulators for Cathare 2 V1.3E code. The nodalizations have been set up using the same general scheme for the considered facilities (Lobi, Spes, Bethsy, Lstf). The geometrical configuration of the various plants considered has been reproduced representing the different zones with the same elements in the code. Criteria already tested for nodalizations development have been followed to assure the geometrical fidelity to the represented systems and new criteria have been introduced to assure the maximum possible similarity among the nodalizations. This activity will lead to reduce the effect of differences in the nodalization when comparing calculations of similar experiments, in particular counterpart tests performed in differently scaled facilities. The nodalizations that have been set up are suitable for every kind of transient. The four nodalizations have been tested at a steady state level against experimental data derived from the facilities. (author)

  19. 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.

  20. Behaviour of organic iodides under pwr accident conditions

    International Nuclear Information System (INIS)

    Laboratory experiments were performed to study the behaviour of radioactive methyl iodide under PWR loss-of-coolant conditions. The pressure relief equipment consisted of an autoclave for simulating the primary circuit and of an expansion vessel for simulating the conditions after a rupture in the reactor coolant system. After pressure relief, the composition of the CH3sup(127/131)I-containing steam-air mixture within the expansion vessel was analysed at 80 0C over a period of 42 days. On the basis of the values measured and of data taken from the literature, both qualitative and quantitative assessments have been made as to the behaviour of radioactive methyl iodide in the event of loss-of-coolant accidents. (author)

  1. Numerical regulation of a test facility of materials for PWR

    International Nuclear Information System (INIS)

    The installation aims at testing materials used in nuclear power plants; tests consists in simulations of a design basis accident (failure of a primary circuit of a PWR type reactor) for a qualification of these materials. A description of the test installation, of the thermodynamic control, and of the control system is presented. The organisation of the software is then given: description of the sequence chaining monitor, operation, list and function of the programs. The analog information processing is also presented (data transmission). A real-time microcomputer and clock are used for this work. The microprocessor is the 6800 of MOTOROLA. The microcomputer used has been built around the MC 6800; its structure is described. The data acquisition include an analog data acquisition system and a numerical data acquisition system. Laboratory and on-site tests are finally presented

  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. DEPCO-MULTI, Subcooled Decompression in PWR Primary System LOCA

    International Nuclear Information System (INIS)

    1 - Nature of physical problem solved: DEPCO-MULTI is used to analyze the subcooled decompression in a multiple pipe network such as a PWR primary coolant system after an instantaneous pipe break. 2 - Method of solution: The unsteady one dimensional mass, momentum and energy conservation laws are converted to a set of compatibility conditions and characteristic directions equivalent to them. After the simplification of the compatibility conditions, they are numerically integrated by the explicit formula. 3 - Restrictions on the complexity of the problem: (1) Maximum number of different thermodynamic initial conditions: 5; (2) Maximum output points of: - pressure history: 5; - pressure difference history: 3; - flow velocity history: 1. (3) The assumptions of thermal equilibrium, no buoyancy effect, no heat transfer and one-dimensional analysis are used

  4. 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.

  5. 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

  6. 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)

  7. PWR power plant pump reliability data. Interim report

    International Nuclear Information System (INIS)

    This report represents a portion of the EPRI effort to collect data relevant for use in nuclear power plant risk studies. The work reported here involved collection and analysis of failure and repair data for pumps in PWR plant operating and safety systems. Failures are classified by failure mode, cause, and part. The failure parameters estimated for pumps are the probability of failure on demand and rate of failure per unit of running time. Different approaches to aggregating the data are presented, and failure parameters are estimated under each approach. Common cause failures are identified, and beta factors are calculated for pumps in some systems. Repair man-hours data are used to estimate a repair time distribution, and the unavailability of pumps in different systems due to repair after failure is calculated

  8. Toward an early detection of PWR control rod anomalous dropping

    International Nuclear Information System (INIS)

    Some anomalous PWR control rods dropping occurred in the past. It is assumed to be caused by a geometrical deformation of its guide tube, which might be related with neutron fluence and its sharp changes. Now at days, this problem is an open field of research, oriented to the understanding and prevention of the event. Work here is focused toward early detection. A differential equation modelling control rod free fall movement is found. There result three acceleration terms: gravity; friction with fluid; and friction with its guide tube. From recorded Plant measurements, both friction coefficients are estimated. The one from guide tube experiences a large variation in case of anomalous dropping; so relationship with neutron fluence is proposed for the prevention purpose. (Author)

  9. Influence of spectral history on PWR full core calculation results

    International Nuclear Information System (INIS)

    The few-group cross section libraries, used by reactor dynamics codes, are affected by the spectral history effect-a dependence of fuel cross sections not only on burnup, but also on local spectral conditions during burnup. A cross section correction method based on Pu-239 concentration was implemented in the reactor dynamic code DYN3D. This paper describes the influence of a cross section correction on full-core calculation results. Steady-state and burnup characteristics of a PWR equilibrium cycle, calculated by DYN3D with and without cross section corrections, are compared. A study has shown a significant influence of spectral history on axial power and burnup distributions as well as on calculated cycle length. An impact of the correction on transient calculations is studied for a control rod ejection example. (Authors)

  10. Low concentration NP preoxidation condition for PWR decontamination

    International Nuclear Information System (INIS)

    To use preoxidation condition with low concentration NP (nitric acid permanganate) instead of conventional high concentration AP (alkline permanganate ) for PWR oxidation decontamination (POD) was summarized. Experiments including three parts have been performed. The defilming performance and decontamination factor of preoxidation with low concentration NP, which is 100, 10 times lower than that of AP are better than that with high concentration AP. The reason has been studied with the aid of prefilmed specimens of corrosion potential measuring in NP solution and chromium release in NP and AP solutions. The behaviour of alloy 13 prefilmed specimen in NP preoxidation solution is different from 18-8 ss and Incoloy 800. In the low acidity, the corrosion potential moves toward positive direction as the acidity becomes high

  11. Delayed phenomena analysis from French PWR containment instrumentation system

    International Nuclear Information System (INIS)

    The analysis of the large amount of measurements which has been now gathered by EDF on its twenty two PWR 900 MW shows that the behaviour of concrete under creep and shrinkage effects is in good agreement with the values given as correct estimates by french regulations and taken into account for the design of nuclear prestressed structures. None of the containment buildings studied here showed significant differences with the regulations theoretical values and consequently all the measurements remain in the field of the allowable strain variations used for design. On the other hand, if the instant loading elastic modulus is clearly determined for each containment, and its effect on theoretical creep taken into account, it was not possible up till now to extract from measurements some particular effects such as type of concrete and agregates or climatic effects. (orig.)

  12. 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

  13. Analysis of reactivity insertion accidents in PWR reactors

    International Nuclear Information System (INIS)

    A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (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. Corrosion Resistance Evaluation of HANA Claddings in Commercial PWR

    International Nuclear Information System (INIS)

    Korea Atomic Energy Research Institute (KAERI) in collaboration with KEPCO Nuclear Fuel (KNF) developed newly-advanced alloy which are named HANA (High-performance Alloy for Nuclear Application) for high burnup PWR nuclear fuel, showed an excellent out-pile corrosion resistance in PWR simulating loop conditions. And in-pile corrosion resistance of HANA claddings, which was examined at the first provisional inspection after -185 FPD of irradiation in the Halden Reactor, and also shown superior to the other references alloy. Also, other researches showed a much better corrosion resistance when compared to the other Zr-based alloy in various corrosion conditions. In this study, the LTA program for newly-developed fuel assembly (HIPER) with the HANA claddings was implemented to justify the performance for 3 cycles of operation schedule in Hanul nuclear power plant. The objective of this study is to compare corrosion properties of reference alloy with HANA claddings loaded in Hanul nuclear power plant.. For the examination procedures, the oxide thickness measurements method and equipment of PSE are described in detail as follow in measurement methods chapter. Finally, based on the above mentioned measurements method, the summarized oxide thickness data obtained from PSE are evaluated for the corrosion resistance in commercial nuclear power plant and some discussion for the corrosion resistance are described. In the past, corrosion resistance of HANA claddings was successfully conducted in test reactor. In this study, the corrosion characteristic of HANA claddings which are applied to HIPER is examined in the commercial nuclear power plant. HANA claddings in the HIPER showed a more improved corrosion resistance than reference alloy claddings and are evaluated well with meeting the oxide thickness criteria

  16. 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

  17. 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

  18. PWR operation and reloading: EDF experience and developments

    International Nuclear Information System (INIS)

    The large experience accumulated by EDF in PWR operation and reloading for about fifteen years required reliable and industrial techniques. Presently, about 54 units of 900 MWe and 1300 MWe PWR's are being operated through various fuel managements (three-batch cycle, four-batch cycle, plutonium recycling). EDF has developed two sets of automatized computational sequences with automatic generation of input data and core calculations for both, the Loading Pattern (LP) optimization and initialization of input data (fuel reshuffling), and for reload related calculations (safety evaluation, start-up physics tests prediction, operating data). As far as the LP search is concerned, it consists in a technique of 'trial and error' based upon knowledge and which is under very severe constraints. Then, reload values prediction and core following are performed with codes and calculational methods which have a high level of qualification and calibration over the large experience of in-core measurements. With respect to these different points, continuous efforts are done aimed at improving the overall reloading methods. Developments are being achieved at different levels. Because of load following perturbations, on-line and off-line core power distribution followings are evaluated with fast nodal CAROLINE code. This one is derived from the 3D design COCCINELLE code developed by EDF, and whose main features are 3D core calculations with optimized numerical schemes and fast resolution techniques, fuel thermal and neutronic feed-back effects modelling (pin by pin). As an alternative to LP manual design used currently, EDF has examined two possible approaches: expert system and optimization package. As far as automatic sequences are concerned, a new technique of automatic generation of input files was evaluated but priority has been given to improvements in physics by more 3D extensive calculations with the new COCCINELLE code

  19. 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)

  20. 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)

  1. Integrated chemical effects test program for PWR sump performance assessment

    International Nuclear Information System (INIS)

    Products attributable to chemical interactions between the emergency core cooling system (ECCS) containment spray water and exposed materials (such as metal surfaces, paint chips, and fiberglass insulation debris) could impede the performance of ECCS recirculation following a loss-of-coolant accident (LOCA) at a pressurized-water reactor (PWR). Five tests have been conducted in the ICET (Integrated Chemical Effects Test) test loop in order to simulate the chemical environment present inside a PWR containment water pool following a LOCA. The tests were conducted for 30 days at a constant temperature of 60 Celsius degrees. The materials tested within this environment included representative amounts of submerged and un-submerged aluminum, copper, concrete, zinc, carbon steel, and insulation samples (either 100% fiberglass or a combination of 80% calcium-silicate and 20% fiberglass by volume). Representative amounts of concrete dust and latent debris were also added to the test solution. Water was circulated through the bottom portion of the test tank during the entire test to achieve representative flow rates over the submerged specimens. Overall, the ICET program provided some insights and initial understanding regarding solution chemistry, as well as the types and amounts of chemical reaction products that may form in the ECCS containment sump pool. The observed chemical products may potentially contribute to pressure losses across a debris-laden sump screen, as well as performance degradation of ECCS components downstream of the sump screen. The ICET results indicate that: -1) chemical reaction products with varied quantities, consistencies, attributes, and apparent formation mechanisms were found; -2) containment materials (metallic, non-metallic, and insulation debris), pH, buffering agent, temperature, and time are all important variables that influence chemical product formation; and -3) changes to one important environmental variable (e.g., pH adjusting agent

  2. PWR operation and reloading: EDF experience and developments

    International Nuclear Information System (INIS)

    The large experience accumulated by EDF in PWR operation and reloading for about fifteen years required reliable and industrial techniques. Presently, about 54 units of 900 MWe and 1300 MWe PWR's are being operated through various fuel management (three-batch cycle, four-batch cycle, plutonium recycling). EDF has developed two sets of automatized computational sequences with automatic generation of input data and core calculations for both, the Loading Pattern (LP) optimization and initialization of input data (fuel reshuffling), and for reload related calculations (safety evaluation, start-up physics tests prediction, operating data). As far as the LP search is concerned, it consists in a technique of ''trial and error' based upon knowledge and which is under very severe constraints. Then, reload values prediction and core following are performed with codes and calculational methods which have a high level of qualification and calibration over the large experience of in-core measurements. With respect to these different points, continuous efforts are done aimed at improving the overall reloading methods. Developments are being achieved at different levels. Because of load following perturbations, on-line and off-line core power distribution followings are evaluated with fast nodal CAROLINE code. This one is derived from the 3D design COCCINELLE code developed by EDF, and whose main features are 3D core calculations with optimized numerical schemes and fast resolution techniques, fuel thermal and neutronic feed-back effects modelling (pin by pin). As an alternative to LP manual design used currently, EDF has examined two possible approaches: Expert system and optimization package. As far as automatic sequences are concerned, a new technique of automatic generation of input files was evaluated but priority has been given to improvements in physics by more 3D extensive calculations with the new COCCINELLE code. (author). 4 refs, 3 figs

  3. 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

  4. Presentations on Ageing management of PWR piping systems

    International Nuclear Information System (INIS)

    The structural integrity of the reactor coolant system (RCS) of PWR's is a key safety issue. As high-energy piping and essential piping system for cooling the reactor core, the design basis considered different hypothetical double end guillotine break ruptures, generally in 11 locations [fig. 1]. The consequences of that importance are the design, fabrication rules and the surveillance programs of these lines in operation. The field experience is in accordance with these precautions and limited degradations have been encountered up to now. Different evolution of these initial design bases and different practices of surveillance program are used in a case by case application process in different countries: - leak before break - realistic break opening section and break opening time - risk informed in-service inspection The RCS piping systems of existing PWRs have generally a very high quality standard for design, fabrication and operation rules. The corresponding field experience confirms very limited degradations encountered in these systems. Nevertheless, some degradations appears recently (SCC of DMW in VC SUMMER and RINGHALS), some have no direct consequences encountered (as loss of toughness by thermal ageing), some are potential and not encountered for the moment (but no ISI can justify absence of early degradation due to the thickness of these piping systems); some other questions are not completely covered in this presentation, as risk of brittle fracture for some cladded ferritic pipings. The RCS reliability remains very high, but an important effort has to be maintain to understand, case by case, encountered and potential degradation mechanisms is an essential contribution to assure long term high safety level of PWR plants

  5. 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.

  6. Chinese varkens in Nederland

    OpenAIRE

    Zhang, W.L.; Huiskes, J.H.

    1992-01-01

    In China zijn in totaal 300 miljoen varkens van meer dan 100 rassen. Ze worden voor verschillende soorten productie gebruikt. Sommige Chinese rassen zijn in Frankrijk, Nederland en U.S.A. geomporteerd. De vraag is of Chinese varkens nuttig kunnen zijnvoor de varkensproductie in Nederland en zo ja welke Chinese varkens

  7. 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

  8. 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

  9. 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)

  10. Determination of uranium in PWR spent fuels by coulometric titration method

    International Nuclear Information System (INIS)

    Controlled-potential coulometric titration method was applied in 0.5M sulphuric acid medium for the determination of uranium content in samples of PWR spent fuel. In this study, we discussed some experimental conditions related to the determination of uranium in PWR spent fuel samples. Accuracy(recovery of uranium) for the coulometric determination of 1∼7mg uranium standard was 99.96∼100.88%. Precision(relative standard deviation, rsd) for the coulometric determination(n=3) of 3∼4mg uranium in PWR spent fuel samples was 0.07∼0.68%. Relative error for the results of the potentiometric and coulometric determination of uranium PWR spent fuel samples was +0.65∼-2.76%

  11. Recovery and separation for the trace amounts of iodide in PWR spent fuel

    International Nuclear Information System (INIS)

    An separation and recovery technique for iodide in spent pressurized water reactor (PWR) fuels has been established using a SIMFUEL simulated for spent PWR fuel. The spent PWR fuels were dissolved with mixture of nitric and hydrochloric acids(80; 20 mol%) which can oxidize iodide to iodate through dissolution process. Iodide in uranium matrix and co-exist fission products was separated and recovered by organic extraction of iodine with carbon tetrachloride and by back extraction of iodide with 0.1 M NaHSO3. Recovered iodide was measured using an ion chromatograph/shielding system available for analysis of radioactive materials. In practice, a spent PWR fuel whose burnup rate was 42,261 MWd/MtU was analyzed and then the relation between the burnup and the quantity of the fission products was compared to the calculated by burnup code, Origen 2

  12. Contribution to a model for stress corrosion cracking of Alloy 600 in PWR primary water

    International Nuclear Information System (INIS)

    Nickel base alloys such as Alloy 600 are widely used for Pressurized Water Reactors (PWR) components. One of the main drawbacks of Alloy 600 is its susceptibility to intergranular stress corrosion cracking (IGSCC) in PWR primary water. This phenomenon has been extensively studied since more than 30 years and a lot of data are now available in the literature. However, the models proposed are still under debate as the mechanisms of cracking are still not well-known. The aim of this study is to improve our knowledge of SCC mechanisms of Alloy 600 in PWR primary water. The influence of intergranular carbides precipitation and cold-working on intergranular oxide penetrations after exposure in simulated PWR primary environment was more specifically studied. The morphology, the chemical nature and crystalline structure of the oxide formed at the surface of the samples and inside the grain boundaries were characterized using analytical Transmission Electron Microscopy (TEM). (authors)

  13. 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.

  14. PWR plant operator training used full scope simulator incorporated MAAP model

    International Nuclear Information System (INIS)

    NTC makes an effort with the understanding of plant behavior of core damage accident as part of our advanced training. For the Fukushima Daiichi Nuclear Power Station accident, we introduced the MAAP model into PWR operator training full scope simulator and also made the Severe Accident Visual Display unit. From 2014, we will introduce new training program for a core damage accident with PWR operator training full scope simulator incorporated the MAAP model and the Severe Accident Visual Display unit. (author)

  15. 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

  16. 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

  17. Application of PWR LOCA margin with the revised appendix K rule

    International Nuclear Information System (INIS)

    Today's focus for nuclear power plant utility owners is to improve plant performances such that the cost per kilowatthour is minimized with enhanced safety. This paper will discuss the impact of design and licensed margin on PWR plant performance, how these margins can be used to improve PWR performance, and how Westinghouse is addressing the regulatory design limits for large break and small break LOCA which impact core thermal design margin. (orig./GL)

  18. ORIGEN-2 libraries based on JENDL-3.2 for PWR-MOX fuel

    International Nuclear Information System (INIS)

    A set of ORIGEN-2 libraries for PWR MOX fuel was developed based on JENDL-3.2 in the Working Group on Evaluation of Nuclide Production, Japanese Nuclear Data Committee. The calculational model generating ORIGEN-2 libraries of PWR MOX is explained here in detail. The ORIGEN-2 calculation with the new ORIGEN-2 MOX library can predict the nuclides contents within 10% for U and Pu isotopes and 20% for both minor actinides and main FPs. (author)

  19. Advanced passive PWR AC-600: Development orientation of nuclear power reactors in China for the next century

    International Nuclear Information System (INIS)

    Based on Qinshan II Nuclear Power Plant that is designed and constructed by way of self-reliance, China has developed advanced passive PWR AC-600. The design concept of AC-600 not only takes the real situation of China into consideration, but also follows the developing trend of nuclear power in the world. The design of AC-600 has the following technical characteristics: Advanced reactor: 18-24 month fuel cycle, low neutron leakage, low power density of the core, no any penetration in the RPV below the level of the reactor coolant nozzles; Passive safety systems: passive emergency residual heat removal system, passive-active safety injection system, passive containment cooling system and main control room habitability system; System simplified and the number of components reduced; Digital I and C; Modular construction. AC-600 inherits the proven technology China has mastered and used in Qirtshan 11, and absorbs advanced international design concepts, but it also has a distinctive characteristic of bringing forth new ideas independently. It is suited to Chinese conditions and therefore is expected to become an orientation of nuclear power development by self-reliance in China for the next century. (author)

  20. Analysis of cobalt source into the primary coolant system of a PWR

    International Nuclear Information System (INIS)

    The understanding of deposition mechanism of 60Co in primary coolant system is very important to find ways to reduce the radiation worker's exposure. In order to develop a deposition model of 60Co in the primary coolant system, the release rate of cobalt source into the coolant in a PWR should be evaluated. By reviewing previous work regarding 60Co buildup in the primary coolant system of PWR, ionic dissolution of corrosion products from oxide films into the coolant was identified as a governing process of release mechanism of cobalt source into the coolant in PWR condition. Release rate constants, 4.16 x 10-7s-1 for stainless steel and 5.56 x 10-8 s-1 for Inconel-600, were obtained by assuming that the dissolution rate is proportional to the thickness of oxide films in a thin oxide film. Total input of cobalt from structural materials in a typical PWR evaluated from the release rate constants reveals that the main source of cobalt input into the coolant system in the PWR is the corrosion of Inconel-600, which covers more than 65% of total cobalt input the PWR primary coolant system

  1. 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...

  2. Danish-accented Chinese

    DEFF Research Database (Denmark)

    Wang, Lei; Sloos, Marjoleine 莱娜; Zhang, Chun

    In search for a linguistic basis for the education of Chinese as a foreign language CFL in Denmark, we set up a new line of investigation into CFL. This research focuses on the phonetics and phonology of Mandarin Chinese as compared to Danish. Considering the sound systems of both languages, we...... note some differences and similarities. The most remarkable differences are: -Chinese has rhotic sounds (pinyin ch, zh, sh, r) but Danish does not -Chinese has affricates (c z ch zh tɕ j) but Danish does not What Danish shares with Chinese is the contrast between aspirated and plain consonants: pa...

  3. 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)

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

    International Nuclear Information System (INIS)

    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

  5. 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

  6. PWR control rod ejection analysis with the numerical nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, M.; Kochunas, B.; Downar, T. J. [Univ. of California at Berkeley, Berkeley (Canada)

    2008-10-15

    During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of

  7. PWR control rod ejection analysis with the numerical nuclear reactor

    International Nuclear Information System (INIS)

    During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of methods. The paper then

  8. 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

  9. 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

  10. Aerosols behavior inside a PWR during an accident

    International Nuclear Information System (INIS)

    During very hypothetical accidents occurring in a pressurized water ractor, radioactive aerosols can be released, during core-melt, inside the reactor containment building. A good knowledge of their behavior in the humid containment atmosphere (mass concentration and size distribution) is essential in order to evaluate their harmfulness in case of environment contamination and to design possible filtration devices. Accordingly the Safety Analysis Department of the Atomic Energy Commission uses several computer models, describing the particle formation (BOIL/MARCH), then behavior in the primary circuits (TRAP-MELT), and in the reactor containment building (AEROSOLS-PARFDISEKO-III B). On the one hand, these models have been improved, in particular the one related to the aerosol formation (nature and mass of released particles) using recent experimental results. On the other hand, sensitivity analyses have been performed with the AEROSOLS code which emphasize the particle coagulation parameters: agglomerate shape factors and collision efficiency. Finally, the different computer models have been applied to the study of aerosol behavior during a 900 MWe PWR accident: loss-of-coolant-accident (small break with failure of all safety systems)

  11. 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

  12. 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)

  13. 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)

  14. A Feasibility Study of an Integral PWR for Space Applications

    International Nuclear Information System (INIS)

    Fission space power systems are well suited to provide safe, reliable, economic and robust energy sources, in the order of 100 KWe. A preliminary feasibility study of a nuclear fission reactor is here presented with the following requirements: i) high reliability, ii) R and D program of moderate cost, iii) to be deployed within a reasonable period of time (e.g. 2015), iv) to be operated and controlled for a long time (10 years) without human intervention, v) possibly to be also used as a byproduct for some particular terrestrial application (or at least to share common technologies), vi) to start with stationary application. The driving idea is to extend as much as possible the PWR technology, by recurring to an integral type reactor. Two options are evaluated for the electricity production: a Rankine steam cycle and a Rankine organic fluid cycle. The neutronics calculation is based on WIMS code benchmarked with MCNP code. The reactivity control is envisaged by changing the core geometry. The resulting system appears viable and of reasonable size, well fit to the present space vector capabilities. Finally, a set of R and D needs has been identified: cold well, small steam turbines, fluid leakage control, pumps, shielding, steam generator in low-gravity conditions, self pressurizer, control system. A R and D program of reasonable extent may yield the needed answers, and some demanding researches are of interest for the new generation Light Water Reactors. (authors)

  15. 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)

  16. 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

  17. 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.

  18. 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

  19. Assessment of spectral history influence on PWR and WWER core

    International Nuclear Information System (INIS)

    The few-group cross section libraries, used by reactor dynamics codes, are affected by the spectral history effect - a dependence of fuel cross sections not only on burnup, but also on local spectral conditions during burnup. Neglecting this effect leads to an additional component of error in neutron-physical characteristics. Two solution approaches to this problem implemented in the reactor dynamic code DYN3D are described and compared in this paper: a cross section correction method based on 239Pu concentration and separate cross sections treatment for each axial layer of reactor core. Steady-state and burnup characteristics of a PWR and a WWER-1000 cores, calculated by DYN3D with and without cross section corrections, are compared. An impact of the correction on transient calculations is studied for a control rod ejection example. Studies have shown a significant influence of spectral history on axial power and burnup distributions as well as on calculated cycle length. Two different correction methods have shown similar major effects. (orig.)

  20. Experiments on the load following behaviour of PWR fuel rods

    International Nuclear Information System (INIS)

    KWU had studied the effects of load following operation on fuel performance from the beginning of commercial operation of nuclear power plants: The first power cycling experiments were started in 1970 in the nuclear power plant Obrigheim (KWO) and in the High Flux Reactor (HFR) Petten. These power cycling tests performed at various power levels and burnups of up to 25 GWd/t(U) showed that the fuel rod cycling performance compares well with the performance of fuel rods operated under essentially constant load at comparable power levels. Two additional cycling tests as described in this paper were performed on the HFR Petten with preirradiated PWR fuel rods having burnups of up to 40 GWd/t(U). These experiments comprised up to 60 cycles between 250/360 W/cm and 215/320 W/cm with 10% power overshoot (400, 370 W/cm) after each cycle. Also, these experiments ended up with sound fuel rods. Moreover, detailed investigations before and between power cycles and after experiment termination showed clearly that the fuel performance corresponds to a single ramp to peak power and that the cycling effects are indeed very small. This confirmed earlier findings that due to crack reversal in the UO2 the cyclic dimensional changes mainly occur in the UO2 itself. Altogether the experiments show that power cycling does not lead to fuel rod failures, which is also confirmed by successful load follow operation in commercial power plants. (orig.)

  1. On thermodynamic advantages of hybrid PWR-desalination plants

    Energy Technology Data Exchange (ETDEWEB)

    Ansari, K.; Sayyaadi, H.; Amidpour, M.; Saffari, A.; Sabzaligoll, T. [Univ. of Technology, Tehran (Iran, Islamic Republic of). Dept. of Mechanical Engineering

    2008-07-01

    Nuclear desalination processes are used to produce power and potable water, and are regarded as more thermodynamically efficient and economically feasible than single purpose nuclear generators and water production plants. This study discussed a 1000 MW PWR nuclear power plant combined with a MED-TVC desalination unit with a capacity of 25,000 m{sup 3} per day. The dual purpose plant consisted of 3 interconnected systems, notably (1) a nuclear power plant with a conversion cycle for steam power generation and a turbo generator connection, (2) a coupling system, and (3) a thermal seawater desalination plant. Exergetic simulations were conducted to obtained energy and exergy flows for the hybrid plant. Exergetic efficiency, exergy destructions, and exergy losses were obtained for the proposed plant. Results of the analysis indicated that major exergy destructions occurred within the nuclear reactor. Turbines and steam generators were other sources of exergy destruction. It was concluded that the desalination unit was only responsible for 1.1 per cent of the total exergy destruction of the hybrid plant. 16 refs., 6 tabs., 5 figs.

  2. Localization and manufacturing technology of materials for PWR plants

    International Nuclear Information System (INIS)

    The primary coolant system of PWR type reactor consists of reactor vessel, steam generator, pressurizer, primary coolant piping, and primary coolant pump. In the case of forged metal all of required materials used in above mentioned system are being produced in Korea Heavy Industries and Construction Co.. Small quantities of raw materials of austenite series stainless rolling mill products, primary tubes, and heat transfer pipings of steam generators were imported and manufactured domestically. But the primary coolant pumps are directly imported. Structural materials being installed inside reactor vessel and control element drive mechanisms are being designed and manufactured by Korea Heavy Industries and Construction Co. from Yonggwang 5 reactor. Indigenous production of rotor, bucket, nozzle, casing, bolt, and valve was totally accomplished. The motives of indigenous production of products were the results of continued investment on research activities and equipment and instruments. The materials used in primary coolant system were produced and manufactured only by the companies which holds a stringent quality control programs and has ASME MO Stamp. But the required quantity of materials from those companies is so small that most of them are imported. Other than that almost all of materials for nuclear power reactor are domestically produced. (Hong, J. S.)

  3. 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)

  4. Nature and behaviour of particulates in PWR coolants

    International Nuclear Information System (INIS)

    Corrosion product species transported by PWR coolants are present in both soluble and insoluble form. Whereas many comparative studies of corrosion products and their activated species refer to the total concentration carried by the coolant, few specifically address the nature and behavior of the insoluble component. The information summarised here is from five Belgian PWRs where continuous-flow capillary samplers were installed as a modification to installed coolant sampling facilities. A series of sampling campaigns were undertaken covering all phases of reactor operations and transient conditions. Particulate populations can vary widely from reactor to reactor and also under steady operational conditions in the same system. Some variation of particle size is apparent with reactor age. During commissioning nickel-rich particles were dominant whereas iron was the major constituent during power operation. It was found that filterable material above 0.1 μm in size could account for between 50 and 90% of total coolant borne activity due to cobalt-58 and 60. Other characteristics of coolant particulates are considered. (author)

  5. 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

  6. 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.

  7. THERMAL EVALUATION OF PRELIMINARY 21 PWR AUCF DESIGN

    International Nuclear Information System (INIS)

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8, 4-11, 4-24, 5-1, and 5-13, Ref. 5.10) and the Waste Package Plan (pp.3-15, 3-17, and 3-24, Ref. 5.9). The design data request addressed herein is: Characterize the preliminary 21 pressurized water reactor (PWR) advanced (A) uncanistered fuel (UCF) waste package (WP) to show that the design is feasible for use in the MGDS environment. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the UCF WP do not preclude UCF WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the preliminary UCF WP design under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation and to provide the required guidance to determining the major design issues for future design efforts. Future design efforts will focus on UCF design changes as further design and operations information becomes available

  8. Study Of The PWR Fuel Bundle Characteristic With Borated Water

    International Nuclear Information System (INIS)

    Study of the PWR fuel bundle characteristic with 2,4, 2,6, 2,8, 3,0, 3,2 and 3,4 enrichment also with borated water 150 and 200 ppm has been done. The fuel bundle contained 264 fuel elements and water (no fuel elements) are arranged as 17 x 17 matrix and 30,294 cm. The fuel bundle characteristic can be seen from their group constants and the infinite multiplication factor whether more or less than one. The fuel bundle parameters can be found from cell calculation with WIMS PC version program. From the cell calculation shown that the infinite multiplication factor of the fuel bundle with 2,4% enrichment and 200 ppm borated water is 1, 01672, its shown that infinite multiplication factor will less than one with increasing borated water more than 200 ppm. From these result if we would like to design the reactor core with 2,4% minimum enrichment then the maximum borated water is 200 ppm

  9. 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.)

  10. 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)

  11. 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

  12. 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)

  13. 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

  14. The deformation of PWR fuel in a LOCA

    International Nuclear Information System (INIS)

    Available world-wide published data on the deformation of PWR fuel in a loss-of-coolant accident are reviewed. Adequate data exist for the oxidation of Zircaloy up to about 15000C; data are increasingly sparse above this temperature and lacking above the melting point. The US NRC criteria for embrittlement are discussed and considered adequate for undeformed cladding, though they may be less so for deformed thinned material. Cladding deformation and the factors controlling it are considered in the light of data from the US, Germany, Japan and the UK. It is concluded that strains in the range 30% - 70% can be produced in experiments simulating LOCA conditions. The behaviour of cladding is strongly influenced by the spatial distribution of temperature, which is in turn dependent on heat transfer mechanisms at the surfaces of the cladding. No realistic experiment, i.e. one with a multirod array and simulated cooling, has produced deformations which would inhibit quenching. Such experiments have not, however, as yet covered the entire range of conditions which might obtain following a LOCA. (author)

  15. Source term aspects associated with future PWR containment systems

    International Nuclear Information System (INIS)

    The overall objective of reactor safety is to protect the population against dangerous releases of radioactive materials from nuclear power plants. In context with a reinforcement of the defense-in-depth strategy the common safety requirements on future nuclear power plants converge in the objective that these plants should be so safe that even in case of a severe accident there will be no need of off-site emergency actions such as an evacuation or resettlement of the population from the vicinity of a nuclear power plant. It is shown by the example of a future 1400 MWe pressurized water reactor (PWR) plant that this goal can be attained in principle by providing a double containment with the annulus vented via an appropriate emergency standby filter. Within the framework of severe accident consequence mitigation a set of parameters for accident conditions and emergency filter efficiencies is elaborated under which the German lower levels of intervention for evacuation are not attained. (author). 10 refs., 3 tabs., 5 figs

  16. 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.

  17. 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

  18. A comparison of HLW-glass and PWR-borate waste glass

    Science.gov (United States)

    Luo, Shanggeng; Sheng, Jiawei; Tang, Baolong

    2001-09-01

    Glass can incorporate a wide variety of wastes ranging from high level wastes (HLW) to low and intermediate level wastes (LILW). A comparison of HLW-Glass and PWR-borate waste glass is given in this paper. The HLW glass formulation named GC-12/9B and 90-19/U can incorporate 16-20 wt% HLW at 1100°C or 1150°C. The borate waste glass named SL-1 can incorporate 45 wt% borate waste generated from PWR. Their physical properties, characteristic temperatures, chemical durability and leach behavior are summarized here. The comparison indicates: the PWR-glass SL-1 can incorporate up to 45 wt% waste oxides at lower melting temperature (1000°C) in agreement with minimum additive waste stabilization (MAWS) approach; owing to the PWR-borate glass contain less Si and more B and Na, its mass loss is higher than HWR-glass; both HLW-glass and PWR-borate glass have favorable chemical durability and the same leaching phenomena, i.e., Na is mostly depleted, but Ca, Mg, Al and Ti are enriched in the leached surface layer.

  19. UPTF experiment, flow phenomena during full-scale loop seal clearing of a PWR

    International Nuclear Information System (INIS)

    To investigate the flow phenomena in the primary system of a pressurized water reactor (PWR) during a loss-of-coolant accident (LOCA) occurring with a small or intermediate break, experiments were performed at the full-scale upper plenum test facility (UPTF). Within the transient and accident management (TRAM) program integral and separate effect tests were carried out to study loop seal clearing and to provide data for the further improvement of computer codes concerning the reactor safety analysis. This paper describes the UPTF tests that focus on the sequence of loop sealclearance in a four-loop operation for two different cold leg break sizes and the residual water levels, the flow patterns in, and the pressure drops across a single loop seal during the clearing. The UPTF results obtained from a single-loop seal operation are compared with experimental data and correlations available in the literature. Two correlations are proposed which allow the quantification of residual water levels in the loop seal under PWR conditions. It is shown that the steam-water test results gained from the full-scale UPTF with realistic PWR loop seal geometry differ from those obtained from the full or small-scale test facilities under air-water conditions. The UPTF experiments indicate the substantial need for steam-water test data from a full-scale facility with realistic PWR geometries in order to validate PWR LOCA thermal-hydraulic system codes to predict loop seal clearing correctly. (orig.)

  20. 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)

  1. The synergy of PWR and PHWR in Korean nuclear power programme

    International Nuclear Information System (INIS)

    In Korea, 12 nuclear power plants are in commercial operation as of August 1997, and about 36% of the country's total electricity demand is provided by these nuclear power plants. Korea has adopted a two reactor PWR/PHWR (Candu) policy and this unique two reactor policy has received a lot of international attention. In general, many countries have adopted a two or multi reactor policy in order to both enhance the economic use of nuclear energy through various reactor technology developments and to stabilize the nuclear electricity generating system in view of safety. Korean experience has shown more synergy effects than mentioned above. First of all, the feed-back of technological advantages of PWR and PHWR has greatly contributed to the advancement of domestic nuclear industrial capabilities. The two reactor policy, PWR and PHWR, which are the most competitive commercial reactors available these days, has attracted with regard to the economic and operating advantages of the two reactors. In addition, the two reactor policy has contributed to the efficient use of spent PWR fuel and accordingly, the various options for the nuclear fuel cycle, recycling the spent PWR fuel into PHWR, such as CANFLEX-RU and DUPIC. (author)

  2. Modified ADS molten salt processes for back-end fuel cycle of PWR spent fuel

    International Nuclear Information System (INIS)

    The back-end fuel cycle concept for PWR spent fuel is explained. This concept is adequate for Korea, which has operated both PWR and CANDU reactors. Molten salt processes for accelerator driven system (ADS) were modified both for the transmutation of long-lived radioisotopes and for the utilisation of the remained fissile uranium in PWR spent fuels. Prior to applying molten salt processes to PWR fuel, hydrofluorination and fluorination processes are applied to obtain uranium hexafluoride from the spent fuel pellet. It is converted to uranium dioxide and fabricated into CANDU fuel. From the remained fluoride compounds, transuranium elements can be separated by the molten salt technology such as electrowinning and reductive extraction processes for transmutation purpose without weakening the proliferation resistance of molten salt technology. The proposed fuel cycle concept using fluorination processes is thought to be adequate for our nuclear program and can replace DUPIC (Direct Use of spent PWR fuel in CANDU reactor) fuel cycle. Each process for the proposed fuel cycle concept was evaluated in detail

  3. A comparison of HLW-glass and PWR-borate waste glass

    International Nuclear Information System (INIS)

    Glass can incorporate a wide variety of wastes ranging from high level wastes (HLW) to low and intermediate level wastes (LILW). A comparison of HLW-Glass and PWR-borate waste glass is given in this paper. The HLW glass formulation named GC-12/9B and 90-19/U can incorporate 16-20 wt% HLW at 1100 deg. C or 1150 deg. C. The borate waste glass named SL-1 can incorporate 45 wt% borate waste generated from PWR. Their physical properties, characteristic temperatures, chemical durability and leach behavior are summarized here. The comparison indicates: the PWR-glass SL-1 can incorporate up to 45 wt% waste oxides at lower melting temperature (1000 deg. C) in agreement with minimum additive waste stabilization (MAWS) approach; owing to the PWR-borate glass contain less Si and more B and Na, its mass loss is higher than HWR-glass; both HLW-glass and PWR-borate glass have favorable chemical durability and the same leaching phenomena, i.e., Na is mostly depleted, but Ca, Mg, Al and Ti are enriched in the leached surface layer

  4. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO2–Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  5. 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.

  6. Neutronic performance of uranium nitride composite fuels in a PWR

    International Nuclear Information System (INIS)

    Highlights: • Survey and sensitivity assembly level studies for uranium nitride composite fuels. • Composites harden the neutron spectrum and decrease the worth of control rods. • Moderator temperature coefficient is more negative, soluble boron coefficient is less negative. • Similar equilibrium core power peaking and reactivity coefficient when compared to UO2. • Illustrates “do no harm” in evaluation of candidate accident tolerant fuels. - Abstract: Uranium mononitride (UN) based composite nuclear fuels may have potential benefits in light water reactor applications, including enhanced thermal conductivity and increased fuel density. However, uranium nitride reacts chemically when in contact with water, especially at high temperatures. To overcome this challenge, several advanced composite fuels have been proposed with uranium nitride as a primary phase. The primary nitride phase is “shielded” from water by a secondary phase, which would allow the potential benefits of nitride fuels to be realized. This work is an operational assessment of four different candidate composite materials. We considered uranium dioxide (UO2) and UN base cases and compared them with the candidate composite UN-based fuels. The comparison was performed for nominal conditions in a reference PWR with Zr-based cladding. We assessed the impact of UN porosity on the operational performance, because this is a key sensitivity parameter. As composite fuels, we studied UN/U3Si5, UN/U3Si2, UN/UB4, and UN/ZrO2. In the case of UB4, the boron content is 100% enriched in 11B. The proposed zirconium dioxide (ZrO2) phase is cubic and yttria-stabilized. In all cases UN is the primary phase, with small fractions of U3Si5, U3Si5, UB4, or ZrO2 as a secondary phase. In this analysis we showed that two baseline nitride cases at different fractions of theoretical density (0.8 and 0.95) generally bound the neutronic performance of the candidate composite fuels. Performance was comparable with

  7. Nuclear power R and D in China

    International Nuclear Information System (INIS)

    As one of the fastest developing countries, China is anxious for enormous electricity supply. To meet the increasing demand for electricity for the sustainable economic development, changing the energy mix and mitigating the environment pollution impact caused by fossil fuel power plant, a medium and long term electrical power development program has been established. It is estimated that the nuclear power capacity will reach at 40000 megawatts from the current 8700 megawatts by the year of 2020, the nuclear power share will account for 4.0-5.0 percent of the total installed capacity. In the 1970s, the Chinese government started developing nuclear technology for power generation, and succeeded in developing Qinshan-1 nuclear power plant with the capacity of 300 MWe, high temperature gas-cooled experiment reactor with the thermal power of 10 MW. Now China fast experiment reactor with the capacity of 50 MWe is under construction. The Chinese government will strengthen self-reliance innovation of nuclear power R and D in the medium and long term, the focal points of the works will comprise: advanced PWR nuclear power unit with 1000 MWe class (meets the requirements of URD or EUR), spent fuel disposal, high temperature gas cooled reactor, fast neutron reactor, integrated reactor, supercritical water cooled reactor, nuclear fusion etc. The government encourages and supports the international exchange and cooperation in the nuclear field. (authors)

  8. Assessment of PWR Steam Generator modelling in RELAP5/MOD2

    International Nuclear Information System (INIS)

    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

  9. Technique of chemical cleaning for removing corrosion products in nuclear reactor (PWR)

    International Nuclear Information System (INIS)

    The study of chemical cleaning technique for removing corrosion products in PWR type plant and power reactor have been carried in China Institute of Atomic Energy (CIAE), Beijing. The report summarizes the in results of screening test and qualification test of chemical cleaning technique, and the results of chemical cleaning to remove corrosion products (Fe304) in primary side of PWR type power reactor which chemical cleaning process has been carried by CIAE. The chemical cleaning agent (EDTA + assistant agent + inhibitor ) is effective for removing magnetite (Fe304 ≤ 17.5g/l).The process of chemical cleaning includes cleaning, rinse, passivation. The corrosion rate of materials is acceptable. The chemical cleaning technique is effective and safe for PWR type reactor. (author)

  10. 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

  11. Hydraulic test for non-instrumented capsule of advanced PWR fuel pellet

    International Nuclear Information System (INIS)

    This report presents the results of pressure drop test, vibration test and endurance test for Non-instrumented Capsule of Advanced PWR Fuel Pellet which were designed fabricated by KAERI. From the pressure drop test results, it is noted that the flow rate across the Non-instrumented Capsule of Advanced PWR Fuel Pellet corresponding to the pressure drop of 200 kPa is measured to be about 7.45 kg/sec. Vibration frequency for the Non-instrumented Capsule of Advanced PWR Fuel Pellet ranges from 13.0 to 32.3 Hz. RMS(Root Mean Square) displacement for the fuel rig is less than 11.6 μm, and the maximum displacement is less than 30.5 μm. The endurance test was carried out for 103 days and 17 hours

  12. Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

    International Nuclear Information System (INIS)

    Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

  13. Development of in-core fuel management scoping tools for PWR

    International Nuclear Information System (INIS)

    This paper concerns with developing a simplified in-core fuel management scoping tool for PWR. For this purpose the point reactivity model is put into a fuel cycling decision code, FCYPRM. Modified Borresen's coarse-mesh diffusion theory and nodal expansion method are utilized to form a spatial neutron analysis code, CMSNAP. Numerical experiments are performed to determine a set of empirical shuffling rules for working out an automated fuel loading pattern search code, ALPS. The numerical examples are presented for verifying effectiveness and applicability of individual codes. By structuring and applying three codes for reload core design problem of a PWR, it is demonstrated that these codes provide an effective in-core fuel management scoping tool for PWR. (Author)

  14. LOFT: a nuclear plant providing realistic answers to PWR licensing issues

    International Nuclear Information System (INIS)

    This paper discusses the role the Loss-of-Fluid Test (LOFT) Experimental Program has played and will play in addressing licensing issues of interest to the US Nuclear Regulatory Commission (NRC), nuclear steam supply system vendors, and utility power companies. The LOFT facility is an operating, prototypically scaled pressurized water reactor (PWR) system in which experiments are performed that characterize conditions for postulated accidents in a commercial PWR. The accident conditions imposed have not caused damage to the LOFT system, and the system has been succesfully stabilized and recovered in each instance. Data from these experiments have provided and will continue to provide realistic answers to PWR licensing issues both before and after the accident at Three Mile Island

  15. 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.

  16. Study on advanced nuclear fuel cycle of PWR/CANDU synergism

    International Nuclear Information System (INIS)

    According to the concrete condition that China has both PWR and CANDU reactors, one of the advanced nuclear fuel cycle strategy of PWR/CANDU synergism ws proposed, i.e. the reprocessed uranium of spent PWR fuel was used in CANDU reactor, which will save the uranium resource, increase the energy output, decrease the quantity of spent fuels to be disposed and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, the transition from the natural uranium to the recycled uranium (RU) can be completed without any changes of the structure of reactor core and operation mode. Furthermore, because of the low radiation level of RU, which is acceptable for CANDU reactor fuel fabrication, the present product line of fuel elements of CANDU reactor only need to be shielded slightly, also the conditions of transportation, operation and fuel management need not to be changed. Thus this strategy has significant practical and economical benefit

  17. Benefits of Low Boron Core Design Concept for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2009-10-15

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in {sup 10}B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts.

  18. 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)

  19. Probabilistic study of the rupture of PWR vessels

    International Nuclear Information System (INIS)

    The method used, is based on understanding of the failure modes and expressing the conventional concepts of fracture mechanics in a probabilistic form: the fatigue crack growth rate, calculated for conditions of cyclic loading, the initiation of unstable crack propagation, and the possibility of crack arrest. The analysis therefore requires the statistical expression of the factors and parameters which appear in the expressions of the law of crack growth and of toughness, and also those which are used in the calculation of the stress intensity factor K1. All input data are entered in COVASTOL code in histogram form. This code takes into account the degree of correlation between the flaw size and the Paris' law coefficients. It computes the propagation of a given defect in a given position, and the corresponding failure probability during accidental loading. The present analysis is primarily concerned with defects in welds; under cladding defects have also been considered in order to evaluate their harmfulness. Data were collected from 3 European manufacturers: BREDA (Italy) - FRAMATOME (France) and ROTTERDAM NUCLEAR (Netherlands). A total of 338 meters of PWR and BWR shell were analysed. The main conclusions are presented. Loading of the vessel has been computed for 22 observable and incidental conditions. An overall statistical interpretation of all the available (da/dN) ΔK measurement points has been made for SA 508 and SA 533 steels, using four laws: Paris, Forman, Priddle and Walker. The conclusions drawn from the calculations made with the COVASTOL code from the data compiled on French PWRs operated according to EDF rules, are finally presented

  20. 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

  1. French nuclear plants PWR vessel integrity assessment and life management

    International Nuclear Information System (INIS)

    The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

  2. PWR sump screen chemical effect test in FY 2007

    International Nuclear Information System (INIS)

    Corrosion, pressure drop and integrated chemical effect assessment on NPSH (ICAN) test of insulations used in Japanese nuclear power plants were performed in FY 2007. In order to obtain basic data needed for taking into consideration of results of ICAN test, corrosion test was taken in sodium tetraborate (insulation) solution, hydrazine solution and pure water added with hydrochronic acid solution of BWR condition. Concentration of dissolved element of rock wool insulation became higher with increase of PH value while that of calcium sulfate insulation became lower with increase of PH value and showed highest PH value in hydrochronic acid solution, which increased up to 9.2 in 3 hrs after the start of experiment. Pressure drop test was to investigate effects of debris (accumulation of sump screen) state and colloid simulating corrosion products on pressure drop. Colloid particulates were apt to increase pressure drop compared with calcium sulfate. Iron hydroxide and aluminum hydroxide increased pressure drop more than copper oxide. Test using sodium tetraborate as PH control chemical was apt to increase pressure drop while test using hydrazine was difficult to increase. Test using ICAN test solution showed wet glass wool insulation increased pressure drop in short time. ICAN test under PWR containment vessel simulated condition was also performed in sodium tetraborate solution, hydrazine solution and pure water added with hydrochronic acid solution BWR condition. Solubility of aluminum, silicon, iron and copper could be almost calculated from thermodynamics data of each element's oxide or hydroxide. Pressure drop change was so complicated as to reflect respective experimental condition. (T. Tanaka)

  3. New genetic algorithms (GA) to optimize PWR reactors

    International Nuclear Information System (INIS)

    The Haling Power Distribution (HPD) has been applied in a unique process to greatly accelerate the in-core fuel management optimization calculations. These calculations involve; the arrangement of fuel assemblies (FAs) and the placement of Burnable Poisons (BPs) in the fresh FAs. The HPD deals only with the arrangement of FAs. The purpose of this paper is to describe past uses of the HPD, provide an example selected from many similar calculations to explain why and how it can be used, and also to show its effectiveness as a filter in the GARCO GA code. The GARCO (Genetic Algorithm Reactor Core Optimization) is an innovative GA code that was developed by modifying the classical representation of the genotype and GA operators. A reactor physics code evaluates the LPs in the population using the HPD Method, which rapidly depletes the core in a single depletion step with a constant power distribution. The HPD is used basically in GARCO as a filter to eliminate invalid LPs created by the genetic operators, to choose a reference LP for BP optimization, and to create an initial population for simultaneous optimization of the LP and BP placement into the core. The accurate depletion calculation of the LP with BPs is done with the coupled lattice and reactor physics CASMO-4/SIMULATE3 package. However, the fact that these codes validate safety of the core with the added BP placement design also validates the use of the HPD method. The calculations are applied to the TMI-1 core as an example PWR providing concrete results

  4. Chinese Foods; Teacher's Handbook.

    Science.gov (United States)

    Huang, Joe, Ed.

    Different styles of Chinese cooking, traditional food items, cooking utensils, serving techniques, and the nutritional value of Chinese cooking are described in this teaching guide. Lesson plans for the preparation of simple dishes are presented. Recipes, a shopping guide to San Francisco's Chinatown, a guide to sources of supplies, and a…

  5. Confucius Teaching Chinese Abroad

    Institute of Scientific and Technical Information of China (English)

    TANG YUANKAI

    2010-01-01

    @@ On December 1, 2009, the Confucius Institute in Lyon, France, held a plaque-unveiling ceremony. Thomas Boutonnet, a Frenchman who has studied Chinese for 10 years in France and China and who is also an institute supervisor, said the institute would offer courses in Chinese language and culture covering legal, wade and cultural fields.

  6. FROM THE CHINESE PRESS

    Institute of Scientific and Technical Information of China (English)

    1998-01-01

    Equality and harmony are mainstream in Chinese marriage. The conclusion was made by a systematic investigation in 1996 on love and marriage relations between couples in Shanghai, Harbin, Guangdong, Gansu and other regions. Six thousand couples were surveyed in a multi-period, separated level probability sampling; the research was conducted by the study group, "Marriage quality during the period of Chinese social

  7. Say That in Chinese

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Demand for Chinese language learning is fueling all aspects of the market, most notably the textbook publication industry Alarge-scale series of Chinese lan-guage textbooks are to be pub-lished in the coming years jointly by the China International Publi

  8. On Developing Business Chinese.

    Science.gov (United States)

    Hong, Wei

    1996-01-01

    Examines the significance of foreign languages for business, particularly Business Chinese, in the 1990s; its curriculum requirements; and the impact of business languages on international business. The article proposes a developmental plan for Business Chinese at the college level including goals, course materials, learning activities, and…

  9. Chinese by Choice

    Science.gov (United States)

    Beem, Kate

    2008-01-01

    A 2004 College Board survey revealed that school districts around America wanted to offer Chinese, but finding qualified teachers was a problem, says Selena Cantor, director of Chinese Language and Culture Initiatives for the College Board. So last year, a new College Board program brought guest teachers from China to school districts in 31…

  10. Equilibria of Chinese Auctions

    DEFF Research Database (Denmark)

    Branzei, Simina; Forero, Clara; Larson, Kate;

    Chinese auctions are a combination between a raffle and an auction and are held in practice at charity events or festivals. In a Chinese auction, multiple players compete for several items by buying tickets, which can be used to win the items. In front of each item there is a basket, and the play...

  11. Modern Chinese History Studies

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    Famous Foreign Scholars' Lecture Tours in China Around the May Fourth Movement and Changes in Chinese Intellectual Circles From 1919 to 1924, John Dewey and four other famous foreign scholars came to China on lecture tours. These tours were great cultural undertakings to spread Western learning to the East after the First World War. The lectures these schol- ars gave helped to deepen the thoughts of Chinese people, and at the same time encouraged the diversification and evolution of Chinese intellectual circles. Firstly, the lectures hastened the birth of a contemporary Chinese wave of reflection on mo- dernity, and provided a basis for the theoretical views and cultural appeals of Liang Qichao and other members of the socalled "Orient Culture Faction," thereby increasing the tension intrinsic to the development of the New Culture Movement and to the expansion of intellectual horizons in Chinese intellectual circles.

  12. Chinese Companies in Switzerland

    Directory of Open Access Journals (Sweden)

    Esther Kessler

    2014-10-01

    Full Text Available In recent years, some of China’s leading firms have made headlines with their European expansion, by either opening new facilities or by acquiring or merging with significant enterprises in Europe. The goal of this paper is to contribute to the existing literature by examining Chinese enterprises expanding into Switzerland. The study also allows some conclusions for Chinese companies entering Central and Eastern Europe. We analyze via interviews the motivations of Chinese companies to expand into Switzerland as well as their behavior and the impediments in their internationalization process. Our findings show that Chinese companies fail to take advantage of certain benefits of western economies (such as open information and stable rule of law. To move forward efficiently, they should develop competence in dealing systematically with readily available market information, building professional networks that recognize a separation between business life and personal life, and managing their Chinese and foreign employees in the foreign cultural environment.

  13. On Chinese Parody Translation

    Institute of Scientific and Technical Information of China (English)

    熊俊

    2013-01-01

    Chinese parody, as a traditional figure of speech, has captured more and more attention from scholars. The researches conducted up to date are inadequate in theorizing and exploring its translation. This paper, based on the comparative data analysis of Chinese parody translation examples in different types of texts, attempts to probe into the means about how to achieve the clos⁃est function equivalence in rendering Chinese parody under the guidance of Sociosemiotic Approach. It is found that the nature of Chinese parody translation is to achieve the closest natural equivalence or similarity in expressive function, informative func⁃tion, vocative function and aesthetic function in its equivalents in English. And it is suggested that borrowing, imitating, para⁃phrasing and adapting are effective strategies in translating Chinese parody.

  14. 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

  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. Experiments on natural circulation during PWR severe accidents and their analysis

    International Nuclear Information System (INIS)

    Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

  17. 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.

  18. Endurance test for non-instrumented capsule of advanced PWR fuel pellet (test procedure)

    International Nuclear Information System (INIS)

    This test procedure details the test loop, test method, and test procedure for pressure drop, vibration and endurance test of Non-instrumented Capsule of Advanced PWR Fuel Pellet. From the pressure drop test, the hydraulic design requirements of the capsule are verified. HANARO limit condition is checked and the compatibility with HANARO core is verified. From flow induced vibration test vibration frequency, vibration displacement are investigated. The wear of Non-instrumented Capsule of Advanced PWR Fuel Pellet is investigated through endurance test, and these data are used to evaluate the expected wear of during maximum resident time of Non-instrumented Capsule

  19. Analysis of dynamic behavior of a PWR utilizing the computer program SARDAN 2

    International Nuclear Information System (INIS)

    In the design of a PWR nuclear plant it is necessary to verify if the design limits are respected, even under abnormal operation condition. An evolution of SARDAN code, developed to simulate transients in PWR, are presented. The new aspects incorporeted in SARDAN 2 are: the fuel ROD analysis in finite-diference, an open channel model for the critic subchannel analysis and the introduction of a simplified model for the automatic control system. The program has been tested in accident condition II, in special, uncontrolled ROD cluster assembly bank withoraw, dropped full-length assembly group, uncontrolled Boron dilution, and the results obtained were considered satisfactory. (Author)

  20. Strategies for life management of French 900 MWe PWR RPV due to neutron irradiation embrittlement

    International Nuclear Information System (INIS)

    In this paper, the situation of the long-life operation plans for French PWR nuclear power plants was presented, the RPV irradiation surveillance program and its assessment approach were analyzed, and then data obtained from 180 surveillance capsules were summarized and compared with the predictive values. It specially argued for the RPV lifetime assessment techniques and management strategies concerning the neutron irradiation embrittlement, as well as R and D activities, with an aim to provide useful reference to the RPV life management of PWR NPPs in China. (authors)

  1. The behavior of impurity around ion exchanger in PWR primary circuit

    International Nuclear Information System (INIS)

    The behavior of impurity (chloride ions were selected as a representative) around ion exchange demineralizers under PWR conditions are discussed. From laboratory tests, the selectivity coefficients for polyborate ions were obtained and the procedures for calculation of equilibrium concentration of impurities in PWR water conditions were established. Precise measurements in operating plants proved that the impurity concentrations are extremely low (less than 50ppb in Cl-). The correlation between estimated values from laboratory tests and values from in-plant measurements was satisfactory. (author)

  2. 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.

  3. 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

  4. 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)

  5. Behavior of impurity around the ion exchanger in PWR primary circuit

    International Nuclear Information System (INIS)

    The behavior of impurity (chloride ions were selected as a representative) around ion exchange demineralizers under PWR conditions are discussed. From laboratory tests, the selectivity coefficients for polyborate ions were obtained and the procedures for calculation of equilibrium concentration of impurities in PWR water conditions were established. Precise measurements in operating plants proved that the impurity concentrations are extremely low (less than 50 ppb in Cl-). The correlation between estimated values from laboratory tests and values from in-plant measurements was satisfactory. (author)

  6. 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)

  7. Simplified analysis of passive residual heat removal systems for small size PWR's

    International Nuclear Information System (INIS)

    The function and general objectives of a passive residual heat removal system for small size PWR's are defined. The characteristic configuration, the components and the operation modes of this system are concisely described. A preliminary conceptual specification of this system, for a small size PWR of 400 MW thermal, is made analogous to the decay heat removal system of the AP-600 reactor. It is shown by analytic models that such passive systems can dissipate 2% of nominal power within the thermal limits allowed to the reactor fuel elements. (author)

  8. NSSR experiment with 50 MWd/kgU PWR fuel under an RIA condition

    International Nuclear Information System (INIS)

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed. (author). 6 refs, 11 figs, 5 tabs

  9. Recent development for improving of PWR flexibility to load follow and frequency control operation

    International Nuclear Information System (INIS)

    In order to adjust the PWR electricity generation to the consumption network, new operating conditions were established. Those new conditions generate additional mechanical and thermal sollicitations due to the frequent motion of control rod banks, consisting of mechanical fatigue cycling and wear at the level of control rode drive mechanisms, control rods and guide tubes, wear and thermal fatigue cycling at the level of fuel assemblies. This paper presents the various aspects of this program including identification of the most critical areas of components, basic research in laboratories for resolving wear problems in PWR environment, improvement of local hydraulics for reducing loads, and endurance testing of full scale components on testing facilities

  10. A preliminary study of thorium and transuranic advanced fuel cycle utilization in PWR

    International Nuclear Information System (INIS)

    A typical PWR fuel element considering (TRU-Th) cycle was simulated. The study analyzed the behaviour of the thorium insertion spiked with reprocessed fuel considering different enrichments that varied from 5.5% to 7.0%. The reprocessed fuels were obtained using the ORIGEN 2.1 code from a burned PWR standard fuel (33,000 MWd/tHM burned), with 3.1% of initial enrichment, which was remained in the cooling pool for five years. The Kerf, hardening spectrum, and the fuel evolution during the burnup were evaluated. This study was performed using the SCALE 6.0. (author)

  11. Study on Translating Chinese into Chinese Sign Language

    Institute of Scientific and Technical Information of China (English)

    徐琳; 高文

    2000-01-01

    Sign language is a visual-gestural language mainly used by hearing impaired people to communicate with each other. Gesture and facial expression are important grammar parts of sign language. In this paper, a text-based transfor mation method of Chinese-Chinese sign language machine translation is proposed.Gesture and facial expression models are created. And a practical system is im plemented. The input of the system is Chinese text. The output of the system is "graphics person" who can gesticulate Chinese sign language accompanied by facial expression that corresponds to the Chinese text entered so as to realize automatic translation from Chinese text to Chinese sign language.

  12. Chinese nuclear insurance and Chinese nuclear insurance pool

    International Nuclear Information System (INIS)

    Chinese Nuclear Insurance Started with Daya Bay Nuclear Power Station, PICC issued the insurance policy. Nuclear insurance cooperation between Chinese and international pool's organizations was set up in 1989. In 1996, the Chinese Nuclear Insurance Pool was prepared. The Chinese Nuclear Insurance Pool was approved by The Chinese Insurance Regulatory Committee in May of 1999. The principal aim is to centralize maximum the insurance capacity for nuclear insurance from local individual insurers and to strengthen the reinsurance relations with international insurance pools so as to provide the high quality insurance service for Chinese nuclear industry. The Member Company of Chinese Nuclear Pool and its roles are introduced in this article

  13. Research in Ancient Chinese Language

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    JIANG Ji-cheng, A Brief Study of Arabian-Chinese Diaphone in Huihui Prescription Abstract: Based on meterials of Arabian- Chinese diaphone in Huihui Prescription, this paper studies all Chinese phonetic initials and finals in Yuan dynasty, including 21 initials and 34 finals. Key Words: Huihui Prescription, Arabian- Chinese diaphone, transferred sound, International Phonetic Alphabet

  14. 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)

  15. PWR water chemistry controls: a perspective on industry initiatives and trends relative to operating experience and the EPRI PWR water chemistry guidelines

    International Nuclear Information System (INIS)

    An effective PWR water chemistry control program must address the following goals: Minimize materials degradation (e.g., PWSCC, corrosion of fuel, corrosion damage of steam generator (SG) tubes); Maintain fuel integrity and good performance; Minimize corrosion product transport (e.g., transport and deposition on the fuel, transport into the SGs where it can foul tube surfaces and create crevice environments for the concentration of corrosive impurities); Minimize dose rates. Water chemistry control must be optimized to provide overall improvement considering the sometimes variant constraints of the goals listed above. New technologies are developed for continued mitigation of materials degradation, continued fuel integrity and good performance, continued reduction of corrosion product transport, and continued minimization of plant dose rates. The EPRI chemistry program, in coordination with other EPRI programs, strives to improve these areas through application of chemistry initiatives, focusing on these goals. This paper highlights the major initiatives and issues with respect to PWR primary and secondary system chemistry and outlines the recent, on-going, and proposed work to effectively address them. These initiatives are presented in light of recent operating experience, as derived from EPRI's PWR chemistry monitoring and assessment program, and EPRI's water chemistry guidelines. (author)

  16. 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

  17. The inertisation of PWR containments by injection of liquid carbondioxide

    Energy Technology Data Exchange (ETDEWEB)

    Karwat, H.; Stolze, P. [Technical University Munich, Garching (Germany)

    1997-03-01

    the transient distribution until uniform concentration of the atmospheric components steam, air and carbondioxide is reached. Time intervals necessary to achieve the required inerting status of approx. 60 vol% may easily be below 30 minutes. The combination of inertisation with the installation of a number of catalytic recombiners remains as a supplementary option. A limited number of experimental results has been made available by fire fighting industry to facilitate the application of the modified code for validation purposes. Comparisons between analyses and experimental results are satisfactory. However, some tests, more typical for PWR containments would be desireable to confirm the post-accident inertisation simulations. (author)

  18. The inertisation of PWR containments by injection of liquid carbondioxide

    International Nuclear Information System (INIS)

    distribution until uniform concentration of the atmospheric components steam, air and carbondioxide is reached. Time intervals necessary to achieve the required inerting status of approx. 60 vol% may easily be below 30 minutes. The combination of inertisation with the installation of a number of catalytic recombiners remains as a supplementary option. A limited number of experimental results has been made available by fire fighting industry to facilitate the application of the modified code for validation purposes. Comparisons between analyses and experimental results are satisfactory. However, some tests, more typical for PWR containments would be desireable to confirm the post-accident inertisation simulations. (author)

  19. 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)

  20. 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)

  1. Success factors of Chinese restaurants

    OpenAIRE

    Aakala, Liwen Heli

    2009-01-01

    The objectives of the thesis fall into three aspects: 1) understanding Chinese entrepreneurship through some major aspects; such as, the characteristics of Chinese entrepreneur and successful skills needed; 2) scanning the Chinese culture that is associated with their entrepreneurial success in restaurant business as well as understanding the Finnish culture that affects Chinese restaurants’ presence in Finland; 3) acquainting with the competitive strategies that those Chinese restaurants emp...

  2. The Contrast of Chinese and English in the Translation of Chinese Poetry

    OpenAIRE

    Ning Li

    2009-01-01

    Chinese poetry is the soul of Chinese literature and Chinese culture. A good translation of a Chinese verse can promote the prevalence of Chinese culture. In the translation of Chinese poetry, translators should not only keep the characteristics of Chinese poems, but also embody the English characteristics. This article analyzed some versions of translation and proposed factors affecting the translation of Chinese poetry.

  3. 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)

  4. 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.)

  5. 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.)

  6. 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.

  7. Simulation of fission products behavior in severe accidents for advanced passive PWR

    International Nuclear Information System (INIS)

    Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

  8. Atmea launches Atmea1 the mid-sized generation 3+ PWR you can rely on

    International Nuclear Information System (INIS)

    ATMEA, a daughter company of AREVA NP and Mitsubishi Heavy Industries, is developing and will supply ATMEA1, the most advanced 1100 MWe PWR plant with the combination of the unique set of competence and experience of its parent companies. This folder presents the ATMEA1 reactor main features. (J.S.)

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

    International Nuclear Information System (INIS)

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

  10. Recommendations of the MRP-139: Inspection of Welds dissimilar in Nozzles PWR reactor vessel in Spain

    International Nuclear Information System (INIS)

    The guide EPRI MRP-139, which provides the way forward for the inspection and evaluation of dissimilar butt welds, the primary system of PWR reactors, indicating the type of nondestructive testing to be done in these areas, based on discovered several cases of default in lnconel alloys 600 and 182 in American and European plants. The phenomenon of cracking.

  11. 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

  12. ''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.)

  13. Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

    International Nuclear Information System (INIS)

    This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

  14. Achievement of a training simulator for PWR power plant: application to control parametric studies

    International Nuclear Information System (INIS)

    A simulation tool adapted to training tasks is developed. One presents the description of the simulator. One studies the management of a model by NEPTUN X2. A general description of a 900 MW PWR power station and the modelling of the power station are presented. The results obtained on the FIDIANE version of the simulator are finally analyzed

  15. Basic study on PWR plant behavior under the condition of severe accident

    International Nuclear Information System (INIS)

    In this paper, we report on the core cooling effect by natural circulation cooling of the primary cooling system in all core cooling function loss accidents caused by SBO in PWR plant compared with BWR. We also report on the core cooling effect by using air as the final heat sink in place of the seawater by opening the main steam valve of the steam generator. On the other hand, we discuss the behavior of PWR plant in the most serious case that the damage such as LOCA is caused by earthquake and that SBO due to the subsequent tsunami causes the reactor isolation and all function of reactor core cooling system loss. That is the case that LOCA and SBO occur in superimposed manner. We can show the results from the simulation experiments that, in PWR plant, even if it is fell into the reactor core cooling function loss due to SBO, natural circulation cooling can keep the reactor core cool down as long as the feed water is supplied to SG by the turbine-driven auxiliary feed-water pump and also that the cooling effect of even more is expected by ensuring the heat-pass to the atmosphere by opening the main steam valve. We also clarify the plant behaviors under the condition that LOCA and SBO occur in superimposed manner in PWR through the simulation experiments. (author)

  16. 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.

  17. 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

  18. 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)

  19. 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.

  20. Flow induced vibration analysis for preventing PWR fuel rods from excessive fretting wear

    International Nuclear Information System (INIS)

    In order to prevent PWR fuel rods excessive fretting wear, the author analysed flow induced vibration. The methods developed and used by FRAMATOME to analyze and to justify the fuel rod behaviour with respect to flow induced vibrations and wear at grid support locations were presented

  1. 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.)

  2. Design and construction of the primary containment for the Sizewell 'B' PWR

    International Nuclear Information System (INIS)

    The primary containment structure for Sizewell 'B' PWR is the result of a number of years intensive development and design by Nuclear Electric (formerly Central Electricity Generating Board) and their design consultants Nuclear Design Associates. This paper identifies the significant features of the design, and the main stages of construction achieved thus far. (author)

  3. BRAVO [test facility] puts PWR safety and relief valves to the test

    International Nuclear Information System (INIS)

    A new valve blowdown test facility is completing commissioning at the Marchwood Engineering Laboratories of Britain's Central Electricity Generating Board. BRAVO, as the facility is known, will prove the performance of the safety and relief valves to be used in Sizewell B, Britain's first PWR. (author)

  4. Rupther: a simulation approach applied to a PWR vessel failure during a severe accident

    International Nuclear Information System (INIS)

    The Rupther program (Rupture Under Thermal Conditions) is a part of the international researches on severe accidents in the PWR type reactors. The aim of the program is the definition of failure simulation validated by experimental data on vessel steel 16MND5 mechanical properties. The paper presents the program and the first results. (A.L.B.)

  5. 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

  6. 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

  7. Application of the SHM and SPSM for calculations of some problems for WWER and PWR

    International Nuclear Information System (INIS)

    Results of SHM calculations applied to WWER and PWR reactors are presented to demonstrate here the potential of the Surface Harmonics Method (SHM) and the Surface Pseudo-Sources Method (SPSM). These methods indeed were developed as a solution to the problem of improvement of neutron field calculations in the cores of nuclear reactors with nonuniform lattices. (Authors)

  8. Development of an advanced man-machine system for Japanese PWR plants

    International Nuclear Information System (INIS)

    Advanced Man Machine System for Japanese PWR plants (MMS - PWR) is a prototype system of the operator supporting system that has been developed by Mitsubishi Heavy Industries, Ltd., Mitsubishi Electric Co. and Mitsubishi Atomic Power Industries, Inc. for five years from 1987 to 1991, under the financial support of MITI (the Ministry of International Trade and Industry of Japanese government). The aim of this system development is to further increase operation reliability and operability of PWR plants. For this purpose the knowledge engineering and the up-to-date computer technologies have been introduced into the design of a prototype system that can offer the information, infer and judge according to operator's thinking process in grasping the plant status and the operations. Also the system has been verified. A prototype system has the following supporting functions: (1) Operator Supporting Function for Normal Operation: Flexible operator supporting function for re-start-up and load following operation that will be needed for Japanese PWR plants in the future. (2) Operator Supporting Function for Abnormalities and Accidents: Early detection and identification of abnormality or accident of the plant, and guidance to appropriate countermeasures. (3) Operator Supporting Function for Maintenance: Supporting function for evaluation of the influence of maintenance on plant components, as well as for the isolation and restoration procedures during plant operation. (4) Optimum Operation Surveillance Function: Intelligence man machine interface that enables operators to understand various plant data precisely and offers the proper answers of what they need to know. (author). 6 figs, 2 refs

  9. Fuel Cycle Cost Calculations for a 120,000 shp PWR for Ship Propulsion. RCN Report

    International Nuclear Information System (INIS)

    A parametric study of the fuel cycle costs for a 120,000 SHP PWR for ship propulsion has been carried out. Variable parameters are: fuel pellet diameter, moderating ratio and refuelling scheme. Minimum fuel cycle costs can be obtained at moderating ratios of about 2.2. Both fuel cycle costs and reactor control requirements favour the two batch core. (author)

  10. 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

  11. Cultural Characteristics of Chinese Cuisine:From Contrastive Studies of English and Chinese

    Institute of Scientific and Technical Information of China (English)

    乞聪妮

    2014-01-01

    Chinese cuisine plays an important role in Chinese culture. The paper illustrates the features of Chinese cuisine in Chi-nese dish naming from different perspectives, and analyze them from contrastive studies of English and Chinese.

  12. Chinese Female Creativity

    Institute of Scientific and Technical Information of China (English)

    VALERIE; SARTOR

    2011-01-01

    "Many foreigners mistakenly believe that Chinese women are creatively oppressed,that they have been oppressed for centuries," Teacher Yang said,glancing at me wryly."That’s correct," I replied, lifting my eyebrows.

  13. Traditional Chinese Biotechnology

    Science.gov (United States)

    Xu, Yan; Wang, Dong; Fan, Wen Lai; Mu, Xiao Qing; Chen, Jian

    The earliest industrial biotechnology originated in ancient China and developed into a vibrant industry in traditional Chinese liquor, rice wine, soy sauce, and vinegar. It is now a significant component of the Chinese economy valued annually at about 150 billion RMB. Although the production methods had existed and remained basically unchanged for centuries, modern developments in biotechnology and related fields in the last decades have greatly impacted on these industries and led to numerous technological innovations. In this chapter, the main biochemical processes and related technological innovations in traditional Chinese biotechnology are illustrated with recent advances in functional microbiology, microbial ecology, solid-state fermentation, enzymology, chemistry of impact flavor compounds, and improvements made to relevant traditional industrial facilities. Recent biotechnological advances in making Chinese liquor, rice wine, soy sauce, and vinegar are reviewed.

  14. Chinese remainder codes

    Institute of Scientific and Technical Information of China (English)

    ZHANG Aili; LIU Xiufeng

    2006-01-01

    Chinese remainder codes are constructed by applying weak block designs and the Chinese remainder theorem of ring theory.The new type of linear codes take the congruence class in the congruence class ring R/I1 ∩ I2 ∩…∩ In for the information bit,embed R/Ji into R/I1 ∩ I2 ∩…∩ In,and assign the cosets of R/Ji as the subring of R/I1 ∩ I2 ∩…∩ In and the cosets of R/Ji in R/I1 ∩ I2 ∩…∩ In as check lines.Many code classes exist in the Chinese remainder codes that have high code rates.Chinese remainder codes are the essential generalization of Sun Zi codes.

  15. Chinese Remainder Codes

    Institute of Scientific and Technical Information of China (English)

    张爱丽; 刘秀峰; 靳蕃

    2004-01-01

    Chinese Remainder Codes are constructed by applying weak block designs and Chinese Remainder Theorem of ring theory. The new type of linear codes take the congruence class in the congruence class ring R/I1∩I2∩…∩In for the information bit, embed R/Ji into R/I1∩I2∩…∩In, and asssign the cosets of R/Ji as the subring of R/I1∩I2∩…∩In and the cosets of R/Ji in R/I1∩I2∩…∩In as check lines. There exist many code classes in Chinese Remainder Codes, which have high code rates. Chinese Remainder Codes are the essential generalization of Sun Zi Codes.

  16. Traditional Chinese biotechnology.

    Science.gov (United States)

    Xu, Yan; Wang, Dong; Fan, Wen Lai; Mu, Xiao Qing; Chen, Jian

    2010-01-01

    The earliest industrial biotechnology originated in ancient China and developed into a vibrant industry in traditional Chinese liquor, rice wine, soy sauce, and vinegar. It is now a significant component of the Chinese economy valued annually at about 150 billion RMB. Although the production methods had existed and remained basically unchanged for centuries, modern developments in biotechnology and related fields in the last decades have greatly impacted on these industries and led to numerous technological innovations. In this chapter, the main biochemical processes and related technological innovations in traditional Chinese biotechnology are illustrated with recent advances in functional microbiology, microbial ecology, solid-state fermentation, enzymology, chemistry of impact flavor compounds, and improvements made to relevant traditional industrial facilities. Recent biotechnological advances in making Chinese liquor, rice wine, soy sauce, and vinegar are reviewed. PMID:19888561

  17. Chinese Marine Materia Medica

    OpenAIRE

    Peter Proksch

    2014-01-01

    China is one of the first countries to use marine materia medica for treating diseases. Ancient books on Chinese herbology, such as Shennong Bencaojing (Shennong’s Classic of Materia Medica), Xinxiu Bencao (Newly Revised Materia Medica) and Bencao Gangmu (Compendium of Materia Medica), have detailed more than 110 marine herbs and thousands of marine herbal formulas (including those for Chinese food therapy). A great deal of information on marine herbs and their applications in medicine, colle...

  18. The Chinese Politeness Scale

    Institute of Scientific and Technical Information of China (English)

    王喜凤

    2012-01-01

    In order to make sense of what is said in an interaction,we have to look at various factors which relate to social distance and closeness.Generally,these factors include the specific situation language takes place,the relative status of the two participants,the message being delivered and finally the age of the participants.In this article,the Chinese Politeness Scale,based on Chinese social values and tradition,will be explained and demonstrated in detail.

  19. Traditional Chinese Medicine

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    2009013 Clinical observation on treatment of active rheumatoid arthritis with Chinese herbal medicine. SHENG Zhenghe(盛正和), et al.Dept TCM, 5th Affili Hosp, Guangxi Med Univ, Guangxi 545001. Chin J Integr Tradit West Med 2008;28(11):990-993. Objective To study the efficacy and safety of Chinese drugs for expelling evil-wind, removing dampness, promoting blood circulation and invigorating yin in treating active rheumatoid arthritis (RA).

  20. Chinese New Year

    Institute of Scientific and Technical Information of China (English)

    王芳

    2005-01-01

    The Chinese New Year is now known as the Spnng Festival because it starts trom the beginning otspring. Though there are some sayings about its origin (起源), all agree that the word Nian, which inmodern Chinese means “year”, was originally the name of a beast (野兽) that started to eat people thenight before the beginning of a new year.

  1. Chinese Entrepreneurs Go Global

    OpenAIRE

    Daniel Zhou

    2012-01-01

    China may be on the tipping point of explosive global growth. In response to changes in the global economy and an economic slowdown domestically, hundreds of thousands of Chinese SMEs are being encouraged to “go global” by their central and local governments. To a Chinese company, going global requires the expansion of its existing business in other countries or the development of new ventures with partners operating in other countries. Explosive growth in China may be possible, but it will d...

  2. Country Review: Chinese Taipei

    OpenAIRE

    OECD

    2008-01-01

    This report, prepared by the Secretariat of the OECD was the basis for a peer review examination of Chinese Taipei at the OECD’s Global Forum on Competition on 9 February, 2006. Competition law in Chinese Taipei has been an important element of the program of economic reforms that moved the economy from centrally directed emphasis on manufacturing and exports to a market-driven emphasis on services and high technology. The competition law follows mainstream practice about restrictive agreemen...

  3. The Magic of Chinese

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    One of the world's oldest languages appears to have a vibrant futureClassical Greek and Latin, two languages that share an ancient history with Chinese, have been threatened with extinction, being used primarily in classic books or for special purposes. Chinese, on the other hand, is thriving as more and more people develop an interest in learning the language, and its charm has been noticed by linguists.

  4. Performance evaluation of two-stage fuel cycle from SFR to PWR

    International Nuclear Information System (INIS)

    One potential fuel cycle option being considered is a two-stage fuel cycle system involving the continuous recycle of transuranics in a fast reactor and the use of bred plutonium in a thermal reactor. The first stage is a Sodium-cooled Fast Reactor (SFR) fuel cycle with metallic U-TRU-Zr fuel. The SFRs need to have a breeding ratio greater than 1.0 in order to produce fissile material for use in the second stage. The second stage is a PWR fuel cycle with uranium and plutonium mixed oxide fuel based on the design and performance of the current state-of-the-art commercial PWRs with an average discharge burnup of 50 MWd/kgHM. This paper evaluates the possibility of this fuel cycle option and discusses its fuel cycle performance characteristics. The study focuses on an equilibrium stage of the fuel cycle. Results indicate that, in order to avoid a positive coolant void reactivity feedback in the stage-2 PWR, the reactor requires high quality of plutonium from the first stage and minor actinides in the discharge fuel of the PWR needs to be separated and sent back to the stage-1 SFR. The electricity-sharing ratio between the 2 stages is 87.0% (SFR) to 13.0% (PWR) for a TRU inventory ratio (the mass of TRU in the discharge fuel divided by the mass of TRU in the fresh fuel) of 1.06. A sensitivity study indicated that by increasing the TRU inventory ratio to 1.13, The electricity generation fraction of stage-2 PWR is increased to 28.9%. The two-stage fuel cycle system considered in this study was found to provide a high uranium utilization (>80%). (authors)

  5. 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.

  6. 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)

  7. Cost comparison of PWR and PHWR nuclear power plants in Korea

    International Nuclear Information System (INIS)

    A statistical approach is used to investigate the relative economic advantages of pressurized water reactor (PWR) and pressurized heavy water reactor (PHWR-CANDU) nuclear power plants for hypothetical 900Mwe systems with the throwaway fuel cycle to be built in the Republic of Korea. Power cost is decomposed into the cost conponents related to the plant capital, operation and maintenance, working capital requirements and fuel cycle operation. The calculation of construction cost is performed with the modified version of computer code ORCOST, and the modified POWERCO-50 is used to evaluated the cost components. Most of economic parameters are treated as statistical variables, each being given with a certain range. Through a random sampling procedures, the probability histograms on unit plant construction costs and power generating costs are obtained. The power cost probability histograms of the PWR and the PHWR plants overlap considerably, and the power costs of two systems appear to be almost same with the PHWR power cost being 0.4mi11/kwh lower compared with 39.4 mills/kwh for the PWR plant (July 1986 US-dollars). When a construction period of PHWR plant is longer by one year than that of PWR plant, there is no difference in the unit power cost of two plants. This comparison leads to no definite conclusion on the cost advantages of the PWR plant versus the PHWR plant. We conclude that the selection issue of nuclear power plants in Korea still remains an open question and that future effort to solve this question should be made toward economic quantification of those factors such as technology transfer and localization. (author)

  8. Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

    Energy Technology Data Exchange (ETDEWEB)

    Papukchiev, Angel [GRS mbH Forschungsinstitute, Garching (Germany); Schaefer, Anselm [ISaR GmbH, Garching (Germany)

    2008-07-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

  9. Chinese Advertising and Advanced Chinese Culture

    Institute of Scientific and Technical Information of China (English)

    Liu Fan

    2006-01-01

    @@ Chinese advertising has long been inseparable from the Chinese national culture from late Shang Dynasty and early Zhou Dynasty when Jiang Ziya beat sword to spread sound to the 21st century when the badge of Beijing 2008 Olympic Games sweeps the whole world. With cultural trait as one of its fundamental character,advertising naturally becomes one of the most important cultural industries in contemporary era. In recent years because of prevalent theme, "Rediscover the Brilliance of Ancient Cities," promoted by the 39th IAA World Congress and the 12th China Advertising Festival held in China, Beijing and Xi'an, two ancient cities, had been splendidly presented to the whole world.

  10. 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

  11. Chinese boxes: "Typhoon" and Conrad's history of the Chinese

    OpenAIRE

    Kerr, D.

    2009-01-01

    This essay examines the novel "Typhoon," by Joseph Conrad, as a story about Chinamen coolies and seamen onboard a steamer transporting Chinese laborers. It argues that the Chinese boxes of the coolies represent several versions of the history of the Chinese and that the coolies are also representatives of a civilization with ideas and institutions, a law and order of nature. It also describes the images of Chinese life and their association to sociality, cooperation, nourishment, simplicity, ...

  12. 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.

  13. New technical knowledge to be implemented to the revision of rules on pipe wall thinning management for PWR plants

    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)

  14. Teaching Chinese Negotiating Style through Examination of Key Chinese Categories.

    Science.gov (United States)

    Myers, Dan

    This study examined different shades of meaning that a single word may have in Chinese in an effort to better understand the relationship between language and culture. An understanding of the exact meaning of Chinese words and expressions can greatly assist non-Chinese in understanding both the language and the society as a whole. A total of 102…

  15. Contemporary American Chinese Studies

    Institute of Scientific and Technical Information of China (English)

    Qiu Huafei

    2008-01-01

    The rise of modern American scholarship on China was largely attributed to the establishment of the American Joint Committee on Contemporary China (JCCC) in 1959 which sponsored all kinds of activities to promote Chinese studies, ranging from institutional support and financial resources to training courses. Since then, American study of China has entered into a period of sustainability that features academic and group-oriented research. It has become a mainstream discipline in American social science studies.1 There are some distinctive differences between early sinology and modern Chinese Studies: the latter is much more concentrated on the study of issues, comparative historical studies, and contemporary Chinese society. American Chinese studies stresses empirical research, textual data, and the application of theory to practice.Shanghai. He was a Fulbright visiting professor at State University of New York at Geneseo from 2006-2007. This treatise is one of a series of studies for China's National Research Foundation of Philosophy and Social Science (05BGJ012), "American Chinese Studies."

  16. 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)

  17. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    OpenAIRE

    2014-01-01

    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 t...

  18. Reactivity and neutron emission measurements of burnt PWR fuel rod samples in LWR-PROTEUS phase II

    International Nuclear Information System (INIS)

    Measurements have been made of the reactivity effects and the neutron emission rates of uranium oxide and mixed oxide burnt fuel samples having a wide range of burnup values and coming from a Pressurised Water Reactor (PWR). The reactivity measurements have been made in a PWR lattice moderated in turn with: water, a water and heavy water mixture, and water containing boron. An interesting relationship has been found between the neutron emission rate and the measured reactivity. (authors)

  19. 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. PMID:21853349

  20. 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