WorldWideScience

Sample records for stone-webster reference pwr

  1. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    1982-04-01

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

  2. Elaboration and qualification of a reference calculation routes for the absorbers in the PWR reactors

    International Nuclear Information System (INIS)

    Blanc-Tranchant, P.

    1999-11-01

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B 4 C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  3. Radionuclide release from PWR fuels in a reference tuff repository groundwater

    International Nuclear Information System (INIS)

    Wilson, C.N.; Oversby, V.M.

    1985-03-01

    The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high-level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: fuel rod sections split open to expose bare fuel particles; rod sections with water-tight end fittings with a 2.5-cm long by 150-μm wide slit through the cladding; rod sections with water-tight end fittings and two 200-μm-diameter holes through the cladding; and undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested on deionized water. Selected initial results are also given for Turkey Point fuel specimens tested on J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO 2 spent fuel matrix dissolves. Fractional release of 137 Cs and 99 Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water. 8 references, 7 figures, 9 tables

  4. Development of a reference spent fuel library of 17x17 PWR fuel assemblies

    International Nuclear Information System (INIS)

    Rossa, Riccardo; Borella, Alessandro; Van der Meer, Klaas

    2013-01-01

    One of the most common ways to investigate new Non-Destructive Assays (NDA) for the spent fuel assemblies are Monte Carlo simulations. In order to build realistic models the user must define in an accurate way the material compositions and the source terms in the system. This information can be obtained using burnup codes such as ORIGEN-ARP and ALEPH2.2, developed at SCK-CEN. These software applications allow the user to select the irradiation history of the fuel assembly and to calculate the corresponding isotopic composition and neutron/gamma emissions as a function of time. In the framework of the development of an innovative NDA for spent fuel verifications, SCK•CEN built an extensive fuel library for 17x17 PWR assemblies, using both ORIGEN-ARP and ALEPH2.2. The parameters considered in the calculations were initial enrichment, discharge burnup, and cooling time. The combination of these variables allows to obtain more than 1500 test cases. Considering the broad range of the parameters, the fuel library can be used for other purposes apart from spent fuel verifications, for instance for the direct disposal in geological repositories. In addition to the isotopic composition of the spent fuel, the neutron and photon emissions were also calculated and compared between the two codes. The comparison of the isotopic composition showed a good agreement between the codes for most of the relevant isotopes in the spent fuel. However, specific isotopes as well as neutron and gamma spectra still need to be investigated in detail.

  5. Adaptive control for a PWR using a self-tuning reference model concept

    International Nuclear Information System (INIS)

    Miley, G.H.; Park, G.T.; Kim, B.S.

    1992-01-01

    Possible applications of an adaptive control method to a pressurized-water reactor nuclear power plant are investigated. The self-tuning technique with a reference model concept is employed. This control algorithm is developed by combining the self-tuning controller with the model reference adaptive control. This approach overcomes the difficulties in choosing the appropriate weighting polynomials in the cost function of the self-tuning control

  6. Use of a probabilistic safety study in the design of the Italian reference PWR

    International Nuclear Information System (INIS)

    Richardson, D.C.; Russino, G.; Valentini, V.

    1985-01-01

    The intent of this paper is to provide a description of the experience gained in having performed a Probabilistic Safety Study (PSS) on the proposed Italian reference pressurized water reactor. The experience revealed that through careful application of probabilistic techniques, Probabilistic Risk Assessment (PRA) can be used as a tool to develop an optimum plant design in terms of safety and cost. Furthermore, the PSS can also be maintained as a living document and a tool to assess additional regulatory requirements that may be imposed during the construction and operational life of the plant. Through the use of flexible probabilistic techniques, the probabilistic safety model can provide a living safety assessment starting from the conceptual design and continuing through the construction, testing and operational phases. Moreover, the probabilistic safety model can be used during the operational phase of the plant as a method to evaluate the operational experience and identify potential problems before they occur. The experience, overall, provided additional insights into the various aspects of the plants design and operation that would not have been identified through the use of traditional safety evaluation techniques

  7. Elaboration and qualification of a reference calculation routes for the absorbers in the PWR reactors; Elaboration et qualification des schemas de calcul de reference pour les absorbants dans les reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Blanc-Tranchant, P

    1999-11-01

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B{sub 4}C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  8. Enlarged level-1 PSA in regard to assessment of cross-cutting effects of hazards and consideration of their uncertainties for a KONVOI type PWR reference plant

    International Nuclear Information System (INIS)

    Haider, C.; Hofer, E.; Kloos, M.; Kuntze, W.; Liemersdorf, H.; Roewekamp, M.; Schwinges, B.; Tuerschmann, M.; Brenig, H.W.; Sommerfeld, H.

    2001-01-01

    In the frame of supporting BMU regarding generic questions concerning probabilistic safety analyses for nuclear power plants as well as regarding evaluation of nuclear specific standards and guidelines the significant contributions to damage states resulting from plant internal and external hazards had to be estimated for a German Konvoi type PWR reference plant. Furthermore, the suitability of the available methods for assessing these hazards should be checked. In the report presented hereafter, only the plant internal hazard Fire out of all the hazards to be considered was probabilistically analysed in detail First of all, screening analyses were carried out for identifying relevant plant areas and to assess their respective efficiency for a proper selection procedure. For a selected, plant area identified to be relevant (area of the cable distributions inside the reactor containment) an indepth analysis was performed. This analysis included all the steps of the probabilistic assessment, starting from the estimation of the fire occurrence frequency, followed by investigations on the fire effects and fire propagation, up to the determination of the fire induced failure probabilities of safety related equipment including the consequences on systems. In addition, the analyses contained particular uncertainty and sensitivity studies, for which aleatoric and epistemic uncertainties were distinguished. As a result of the screening analyses as well as of the in-depth investigations regarding the fire hazard, no significant contributions for the total frequencies for system, core, or plant damage states have been found. In this context, it has to be noticed that the study presented hereafter does not cover a complete fire PSA. With respect to assessing the available methods it has been found that improvements concerning the screening process as well as concerning the probabilistic fire event tree analyses are necessary. With regard to further hazards, a site specific

  9. ROX PWR

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  10. ROX PWR

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

  11. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  12. PWR system reliability improvement activities

    International Nuclear Information System (INIS)

    Yoshikawa, Yuichiro

    1985-01-01

    In Japan lacking in energy resources, it is our basic energy policy to accelerate the development program of nuclear power, thereby reducing our dependence. As referred to in the foregoing, every effort has been exerted on our part to improve the PWR system reliability by dint of the so-called 'HOMEMADE' TQC activities, which is our brain-child as a result of applying to the energy industry the quality control philosophy developed in the field of manufacturing industry

  13. Maturity of the PWR

    International Nuclear Information System (INIS)

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

    1983-03-01

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

  14. 75 FR 57532 - In the Matter of: Stone & Webster Construction, Inc.; Confirmatory Order (Effective Immediately)

    Science.gov (United States)

    2010-09-21

    ... Any person adversely affected by this Confirmatory Order, other than Shaw, may request a hearing... Safety Conscious Work Environment (SCWE) and Safety Culture: At the parent company, The Shaw Group Inc... performance evaluations; Periodic independent culture surveys. For nuclear maintenance sites: A SCWE Procedure...

  15. French PWR Safety Philosophy

    International Nuclear Information System (INIS)

    Conte, M. M.

    1986-01-01

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

  16. The PWR cores management

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  17. Technology, safety, and costs of decommissioning reference light-water reactors following postulated accidents. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, E S; Holter, G M

    1982-11-01

    Appendices contain information concerning the reference site description; reference PWR facility description; details of reference accident scenarios and resultant contamination levels; generic cleanup and decommissioning information; details of activities and manpower requirements for accident cleanup at a reference PWR; activities and manpower requirements for decommissioning at a reference PWR; costs of decommissioning at a reference PWR; cost estimating bases; safety assessment details; and details of post-accident cleanup and decommissioning at a reference BWR.

  18. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    Tabata, Hiroaki

    1996-01-01

    The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

  19. PWR core design calculations

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  20. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

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

  1. Scaling studies - PWR

    International Nuclear Information System (INIS)

    Sonneck, G.

    1983-05-01

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

  2. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  3. The integrated PWR

    International Nuclear Information System (INIS)

    Gautier, G.M.

    2002-01-01

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

  4. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

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

  5. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

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

  6. Water chemistry in PWR

    International Nuclear Information System (INIS)

    Abe, Kenji

    1987-01-01

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

  7. PWR: 10 years after and perspectives

    International Nuclear Information System (INIS)

    1990-01-01

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

  8. French PWR safety philosophy

    International Nuclear Information System (INIS)

    Conte, M.

    1986-05-01

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

  9. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

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

  10. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    Silliman, P.L.

    1978-01-01

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

  11. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-07-01

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

  12. Safety considerations of PWR's

    International Nuclear Information System (INIS)

    Arnold, W.H. Jr.

    1977-01-01

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

  13. PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

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

    1982-10-01

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

  14. PWR systems transient analysis

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  15. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

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

  16. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

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

  17. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  18. A Multi-Physics PWR Model for the Load Following

    OpenAIRE

    Muniglia , Mathieu; Do , Jean-Michel; Jean-Charles , Le Pallec; Grard , Hubert; Verel , Sébastien; David , S.

    2016-01-01

    International audience; In this paper, a new model of a Pressurized Water Reactor (PWR) is described. This model includes the description of the core as well as a simplified secondary loop: the goal is to reproduce a load-following type transient, where the output power of the plant is controlled by the electric grid. Consequently, the control systems are also modeled, as the control rods or the soluble boron. The reference power plant is a 1300MW electrical PWR, managed with the french G mode.

  19. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-01-01

    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% U 3 O 8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

  1. Evaluation model for PWR irradiated fuel

    International Nuclear Information System (INIS)

    Gomes, I.C.

    1983-01-01

    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) [pt

  2. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2017-04-15

    This paper presents the radiation shielding model of a typical PWR (CNPP-II) at Chashma, Pakistan. The model was developed using Monte Carlo N Particle code [2], equipped with ENDF/B-VI continuous energy cross section libraries. This model was applied to calculate the neutron and gamma flux and dose rates in the radial direction at core mid plane. The simulated results were compared with the reference results of Shanghai Nuclear Engineering Research and Design Institute (SNERDI).

  3. PWR-to-PWR fuel cycle model using dry process

    International Nuclear Information System (INIS)

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  4. PWR plant construction in Japan

    International Nuclear Information System (INIS)

    Tamura, Toshifumi

    2002-01-01

    The construction methods based on the experiences on the Nuclear Island, which is a critical path in the total construction schedule, have been studied and reconsidered in order to construct by more reliable and economical method. So various improved construction method are being applied and the duration of construction is being reduced continuously. So various improved construction method are being applied and the duration of construction is being reduced continuously. In this paper, the history of construction of twenty-three (23) PWR Plant, the actual construction methods and schedule of Ohi-3/4, to which the many improved methods were applied during their construction, are introduced mainly with the improved points for previously constructed plants. And also the situation of construction method for the next PWR Plant is simply explained

  5. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  6. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  7. PWR: nuclear islands

    International Nuclear Information System (INIS)

    1989-01-01

    Framatome and its partners have produced this glossary of technical terms that can be used in writing English language documents relating to power plants (nuclear islands, individual components, nuclear services, etc.) with the hope of improving the quality of the documents intended for their clients, suppliers and partners and for others. This glossary will be particularly useful to the translators and authors of technical proposals, design documents, manufacturing documents, construction and operating documents concerning Pressurized Water Reactors written in English or French. It can also be useful as a reference document for students, researchers, journalists, etc., having to write on this subject. We would like to thank all those individuals working at the Ministere de la Recherche et de la Technologie, Electricite de France, Jeumont Schneider and Framatome who have contributed to this glossary. We would also appreciate any comments or sugestions intended to improve subsequent editions of this glossary [fr

  8. 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.

  9. Surveillance of vibrations in PWR

    International Nuclear Information System (INIS)

    Espefaelt, R.; Lorenzen, J.; Aakerhielm, F.

    1980-07-01

    The core of a PWR - including fuel elements, internal structure, control rods and core support structure inside the pressure vessel - is subjected to forces which can cause vibrations. One sensitive means to detect and analyse such vibrations is by means of the noise from incore and excore neutron detector signals. In this project noise recordings have been made on two occasions in the Ringhals 2 plant and the obtained data been analysed using the Studsvik Noise Analysis Program System (SNAPS). The results have been intepreted and a detailed description of the vibrational status of the core and pressure vessel internals has been produced. On the basis of the obtained results it is proposed that neutron signal noise analysis should be performed at each PWR plant in the beginning, middle and end of each fuel cycle and an analysis be made using the methods developed in the project. It would also provide a contribution to a higher degree of preparedness for diagnostic tasks in case of unexpected and abnormal events. (author)

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

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-02-01

    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%U 3 O 8 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, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

  11. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Young; Chang, M H; Yu, K J; Lee, D J; Cho, B H; Kim, H Y; Yoon, J H; Lee, Y J; Kim, J P; Park, C T; Seo, J K; Kang, H S; Kim, J I; Kim, Y W; Kim, Y H

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  12. Conceptual study on advanced PWR system

    International Nuclear Information System (INIS)

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. 1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. 2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. 3) Control rod drive mechanism for fine control : type and function were surveyed. 4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. 5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. 6) Steam injector concepts: analysis and experiment were conducted. 7) Fluidic diode concepts : analysis and experiment were conducted. 8) Wet thermal insulator : tests for thin steel layers and assessment of materials. 9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs

  13. Friends of the Earth case against the PWR

    International Nuclear Information System (INIS)

    Boyle, S.

    1987-01-01

    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)

  14. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    International Nuclear Information System (INIS)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures

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

    International Nuclear Information System (INIS)

    Dionne, B.J.; Baum, J.W.

    1984-01-01

    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

  16. PWR thermocouple mechanical sealing structure

    International Nuclear Information System (INIS)

    Shen Qiuping; He Youguang

    1991-08-01

    The PWR in-core temperature detection device, which is one of measures to insure reactor safety operation, is to monitor and diagnose reactor thermal power output and in-core power distribution. The temperature detection device system uses thermocouples as measuring elements with stainless steel protecting sleeves. The thermocouple has a limited service time and should be replaced after its service time has reached. A new sealing device for the thermocouples of reactor in-core temperature detection system has been developed to facilitate replacement. The structure is complete tight under high temperature and pressure without any leakage and seepage, and easy to be assembled or disassembled in radioactive environment. The device is designed to make it possible to replace the thermocouple one by one if necessary. This is a new, simple and practical structure

  17. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    Bacher, P.E.

    1987-01-01

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

  18. Technology, safety, and costs of decommissioning a reference pressurized water reactor power station. Appendices

    International Nuclear Information System (INIS)

    Smith, R.I.; Konzek, G.J.; Kennedy, W.E. Jr.

    1978-05-01

    Detailed appendices are presented under the following headings: reference PWR facility description, reference PWR site description, estimates of residual radioactivity, alternative methods for financing decommissioning, radiation dose methodology, generic decommissioning activities, intermediate dismantlement activities, safe storage and deferred dismantlement activities, compilation of unit cost factors, and safety assessment details

  19. Neutronic feasibility of PWR core with mixed oxide fuels in the Republic of Korea

    International Nuclear Information System (INIS)

    Kim, Y.J.; Joo, H.K.; Jung, H.G.; Sohn, D.S.

    1997-01-01

    Neutronic feasibility of a PWR core with mixed oxide (MOX) fuels has been investigated as part of the feasibility study for recycling spent fuels in Korea. A typical 3-loop PWR with 900 MWe capacity is selected as reference plant to develop equilibrium core designs with low-leakage fuel management scheme, while incorporating various MOX loading. The fuel management analyses and limited safety analyses show that, safely stated, MOX recycling with 1/3 reload fraction can be accommodated for both annual and 18 month fuel cycle schemes in Korean PWRs, without major design modifications on the reactor systems. (author). 12 refs, 4 figs, 3 tabs

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

    International Nuclear Information System (INIS)

    Martins Filho, J.R.

    1980-01-01

    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) [pt

  1. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  2. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

    The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

  3. Feasibility of using gadolinium as a burnable poison in PWR cores. Final report

    International Nuclear Information System (INIS)

    Rothleder, B.M.

    1981-02-01

    As an alternative to the use of lumped burnable absorbers in PWR cores, distributed burnable absorbers are being considered for generic application. These burnable absorbers take the form of Gd 2 O 3 mixed with UO 2 in selected fuel rods (as is currently done in BWR cores). The work discussed herein concerns a three-dimensional feasibility study of the use of such distributed burnable absorbers in PWR cores. This study of distributed burnable absorbers was performed for the first cycle of a typical current design PWR using the following steps: analysis of a generic reference core design; determination of gadolinium assembly designs; determination of a generic gadolinium core design; evaluation of feasibility by examining selected parameters; and redesign of the generic gadolinium core, using axial zoning

  4. 21-PWR Waste Package Side and End Impacts

    International Nuclear Information System (INIS)

    V. Delabrosse

    2003-01-01

    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

  5. 21-PWR Waste Package Side and End Impacts

    International Nuclear Information System (INIS)

    T. Schmitt

    2005-01-01

    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

  6. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  7. Upgrading of PWR plant simulators

    International Nuclear Information System (INIS)

    Wada, Tomonori; Sasaki, Kazunori; Nakaishi, Hirokazu.

    1989-01-01

    For the education and training of operators in electric power plants, simulators have been employed, and it is well known that their effect is great. There are operation training simulators which simulate the dynamic characteristics of plants and all the machinery and equipment that operators handle, and train the procedure of restoration at the time of abnormality in plants, education simulators which can analyze the dynamic characteristics of plants efficiently in a short time, and offer information by visualizing phenomena with three-dimensional display and others so as to be easily understandable, and forecast simulators which do the analysis forecasting plant behavior at the time of abnormality in plants, and investigate the necessity of the guide for operation procedure and the countermeasures at the time of emergency. In this explanation, the upgrading of operation training simulators which have been put already in training is discussed. The constitution of simulator system and the instructor function, the outline of PWR plant simulation models comprising thermal flow model, pump model, leak model and so on, the techniques of increasing simulator speed, and the example of analysis using the NUPAC code are reported. (K.I.)

  8. PWR secondary water chemistry study

    International Nuclear Information System (INIS)

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

    1977-02-01

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

  9. 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).

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

    International Nuclear Information System (INIS)

    Kakubo, Masao

    1983-01-01

    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)

  11. Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

    International Nuclear Information System (INIS)

    Sayyaadi, Hoseyn; Sabzaligol, Tooraj

    2010-01-01

    A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

  12. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  13. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  14. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  15. Enlarged level-1 PSA in regard to assessment of cross-cutting effects of hazards and consideration of their uncertainties for a KONVOI type PWR reference plant; Erweiterte PSA der Stufe 1 im Hinblick auf die Behandlung uebergreifender Einwirkungen und die Beruecksichtigung ihrer Unsicherheiten am Beispiel einer Anlage vom Typ Konvoi

    Energy Technology Data Exchange (ETDEWEB)

    Haider, C.; Hofer, E.; Kloos, M.; Kuntze, W.; Liemersdorf, H.; Roewekamp, M.; Schwinges, B.; Tuerschmann, M.; Brenig, H.W.; Sommerfeld, H.

    2001-06-01

    In the frame of supporting BMU regarding generic questions concerning probabilistic safety analyses for nuclear power plants as well as regarding evaluation of nuclear specific standards and guidelines the significant contributions to damage states resulting from plant internal and external hazards had to be estimated for a German Konvoi type PWR reference plant. Furthermore, the suitability of the available methods for assessing these hazards should be checked. In the report presented hereafter, only the plant internal hazard Fire out of all the hazards to be considered was probabilistically analysed in detail First of all, screening analyses were carried out for identifying relevant plant areas and to assess their respective efficiency for a proper selection procedure. For a selected, plant area identified to be relevant (area of the cable distributions inside the reactor containment) an indepth analysis was performed. This analysis included all the steps of the probabilistic assessment, starting from the estimation of the fire occurrence frequency, followed by investigations on the fire effects and fire propagation, up to the determination of the fire induced failure probabilities of safety related equipment including the consequences on systems. In addition, the analyses contained particular uncertainty and sensitivity studies, for which aleatoric and epistemic uncertainties were distinguished. As a result of the screening analyses as well as of the in-depth investigations regarding the fire hazard, no significant contributions for the total frequencies for system, core, or plant damage states have been found. In this context, it has to be noticed that the study presented hereafter does not cover a complete fire PSA. With respect to assessing the available methods it has been found that improvements concerning the screening process as well as concerning the probabilistic fire event tree analyses are necessary. With regard to further hazards, a site specific

  16. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

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

  17. Thermohydraulic calculations of PWR primary circuits

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1984-01-01

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

  18. Reliability of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Ribeiro, A.A.T.; Muniz, A.A.

    1978-12-01

    Results of the analysis of factors influencing the reliability of international nuclear power plants of the PWR type are presented. The reliability factor is estimated and the probability of its having lower values than a certain specified value is discussed. (Author) [pt

  19. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  20. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

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

  1. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    Larsen, N.

    1987-03-01

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  2. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  3. Improvement of PWR reliability by corrosion prevention

    International Nuclear Information System (INIS)

    Takamatsu, Hiroshi

    1999-01-01

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

  4. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  5. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.

    1980-01-01

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.) [pt

  6. CFD simulation of a four-loop PWR at asymmetric operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jian-Ping; Yan, Li-Ming; Li, Feng-Chen, E-mail: lifch@hit.edu.cn

    2016-04-15

    Highlights: • A CFD numerical simulation procedure was established for simulating RPV of VVER-1000. • The established CFD approach was validated by comparing with available data. • Thermal hydraulic characteristics under asymmetric operation condition were investigated. • Apparent influences of the shutdown loop on its neighboring loops were obtained. - Abstract: The pressurized water reactor (PWR) with multiple loops may have abnormal working conditions with coolant pumps out of running in some loops. In this paper, a computational fluid dynamics (CFD) numerical study of the four-loop VVER-1000 PWR pressure vessel model was presented. Numerical simulations of the thermohydrodynamic characteristics in the pressure vessel were carried out at different inlet conditions with four and three loops running, respectively. At normal stead-state condition (four-loop running), different parameters were obtained for the full fluid domain, including pressure losses across different parts, pressure, velocity and temperature distributions in the reactor pressure vessel (RPV) and mass flow distribution of the coolant at the inlet of reactor core. The obtained results for pressure losses matched with the experimental reference values of the VVER-1000 PWR at Tianwan nuclear power plant (NPP). For most fuel assemblies (FAs), the inlet flow rates presented a symmetrical distribution about the center under full-loop operation conditions, which accorded with the practical distribution. These results indicate that it is now possible to study the dynamic transition process between different asymmetric operation conditions in a multi-loop PWR using the established CFD method.

  7. Reliability assessment of the containment of a PWR

    International Nuclear Information System (INIS)

    Schueller, G.I.; Wellein, R.; Wittmann, F.H.; Boulahdour, T.; Mihashi, M.; Zorn, N.F.; Bauer, J.

    1981-09-01

    The aim of this research effort was to contribute to the development of methods to quantify the risk involved with nuclear power plants. Using a large component, i.e. the containment of the reference plant BIBLIS B (PWR) as sample structure a reliability analysis was performed which is based on realistic assumptions of loads and material properties. For this purpose in many fields it was necessary to develop new methods, collect data, and where not available, obtain data in tests. This effort concentrated on partial aspects and on the other hand on the development of a methodology for an overall reliability concept. According to the results of the previous project, the keypoints of this effort are the treatment of loss of coolant accidents (small leak), earthquake loading, the possibly resulting crackpropagation in the steel hull, and the structural mechanics and material strength aspects of the reinforced concrete hull subjected to impact loading (aircraft impact). (orig./HP) [de

  8. French practice for assessing the fission product releases from the containment during a PWR severe accident

    International Nuclear Information System (INIS)

    Duco, J.; Dufresne, J.; L'homme, A.

    1988-10-01

    French safety philosophy as concerns severe PWR accidents has already been outlined by the Director of CEA/IPSN in an article published in ''Nuclear Safety''. Therefore the present paper will focus on: a) the French reference source terms, as used for elaborating ultimate emergency procedures on PWRs and for emergency planning; b) the methods currently developed for more realistic assessments of the release of fission products during a severe accident

  9. Benchmark calculations on resonance absorption by 238U in a PWR pin-cell geometry

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Janssen, A.J.

    1993-12-01

    Very accurate Monte Carlo calculations with MCNP have been performed to serve as a reference for benchmark calculations on resonance absorption by 238 U in a typical PWR pin-cell geometry. Calculations with the energy-pointwise slowing down code ROLAIDS-CPM show that this code calculates the resonance absorption accurately. Calculations with the multigroup discrete ordinates code XSDRN show that accurate results can only be achieved with a very fine energy mesh. (orig.)

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

    International Nuclear Information System (INIS)

    Illy, H.

    1981-12-01

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

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

    International Nuclear Information System (INIS)

    Illy, H.

    1986-09-01

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

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

    International Nuclear Information System (INIS)

    Aragones, J.M.; Corella, M.R.; Martinez-Val, J.M.

    1976-01-01

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

  13. Thermal-hydraulic analysis for wire-wrapped PWR cores

    Energy Technology Data Exchange (ETDEWEB)

    Diller, P. [General Electric Company, 3901 Castle Hayne Rd., Wilmington, NC 28401 (United States)], E-mail: pdiller@gmail.com; Todreas, N. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)], E-mail: todreas@mit.edu; Hejzlar, P. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2009-08-15

    This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH{sub 1.6} (referred to as U-ZrH{sub 1.6}) or UO{sub 2} fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA). The thermal-hydraulic performance of U-ZrH{sub 1.6} and UO{sub 2} were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.

  14. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  15. Design study of a PWR of 1.300 MWe of Angra-2 type operating in the thorium cycle

    International Nuclear Information System (INIS)

    Andrade, E.P.; Carneiro, F.A.N.; Schlosser, G.J.

    1984-01-01

    The utilization of the thorium-highly enriched uranium and thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. The PWR of 1300 MWe from KWU (Angra-2 type) is taken as the reference reactor for the study. Reactor core design calculations for both types of fuels considering once-through and recycle fuels. The calculations were performed with the KWU design codes FASER-3 and MEDIUM 2.2 after introduction of the thorium chain and some addition of nuclide data in FASER-3. A two-energy group scheme and a two-dimensional (XY) representation of the reactor core were utilized. (Author) [pt

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

    International Nuclear Information System (INIS)

    Zhang Min; Jue Ji; Liu Tianshu

    2013-01-01

    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)

  17. The PWR spectral code GELS. Pt. 1

    International Nuclear Information System (INIS)

    Penndorf, K.; Schult, F.; Schulz, G.

    1976-01-01

    The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

  18. Fuel management optimization for a PWR

    International Nuclear Information System (INIS)

    Dumas, M.; Robeau, D.

    1981-04-01

    This study is aimed to optimize the refueling pattern of a PWR. Two methods are developed, they are based on a linearized form of the optimization problem. The first method determines a feasible solution in two steps; in the first one the original problem is replaced by a relaxed one which is solved by the Method of Approximation Programming. The second step is based on the Branch and Bound method to find the feasible solution closest to the solution obtained in the first step. The second method starts from a given refueling pattern and tries to improve this pattern by the calculation of the effects of 2 by 2, 3 by 3 and 4 by 4 permutations on the objective function. Numerical results are given for a typical PWR refueling using the two methods

  19. RSK-guidelines for PWR reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The RSK guidelines for PWA reactors of April 24, 1974, have been revised and amended in this edition. The RSK presents a summary of safety requirements to be observed in the design, construction, and operation of PWR reactors in the form of guidelines. From January 1979 onwards these guidelines will be the basis of siting and safety considerations for new PWR reactors, and newly built nuclear power plants will have to form these guidelines. They are not binding for existing nuclear power plants under construction or in operation. It will be a matter of individual discussion whether or not the guidelines will be applied in these plants. The main purpose of the guidelines is to facilitate discussion among RSK members and to give early information on necessary safety requirements. If the guidelines are observed by producers and operators, the RSK will make statements on individual projects at short notice. (orig./HP) [de

  20. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  1. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  2. EDF/CIDEN - ONECTRA: PWR decontamination

    International Nuclear Information System (INIS)

    Fayolle, P.; Orcel, H.; Wertz, L.

    2010-01-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  3. EPRI PWR primary water chemistry guidelines revision

    International Nuclear Information System (INIS)

    McElrath, Joel; Fruzzetti, Keith

    2014-01-01

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

  4. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    Neubauer, W.

    1979-07-01

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

  5. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    Stutzmann, A.

    1997-01-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  6. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  7. Shielding design for PWR in France

    International Nuclear Information System (INIS)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983

  8. Organization patterns of PWR power plants

    International Nuclear Information System (INIS)

    Leicman, J.

    1980-01-01

    Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)

  9. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    Bruel, Renata Nunes

    1997-01-01

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

  10. T Plant removal of PWR Chiller Subsystem

    International Nuclear Information System (INIS)

    Dana, C.M.

    1994-01-01

    The PWR Pool Chiller System is not longer required for support of the Shippingport Blanket Fuel Assemblies Storage. The Engineering Work Plan will provide the overall coordination of the documentation and physical changes to deactivate the unneeded subsystem. The physical removal of all energy sources for the Chiller equipment will be covered under a one time work plan. The documentation changes will be covered using approved Engineering Change Notices and Procedure Change Authorizations as needed

  11. German risk study of PWR's

    International Nuclear Information System (INIS)

    Kafka, P.

    1983-01-01

    In this paper, first the status of German Risk Study is presented briefly. Specific reference is made to the investigations in Phase B of the study and related programs. Significant elements involved in the risk assessment for NPPs, mainly in the field of system and structural reliability analyses are mentioned. In particular, important outcomes and limiting facts in the process of a Probabilistic Risk Assessment (PRA) to evaluate the safety standard and above all the influence of individual components or subsystems on core melt frequency are discussed. (orig.)

  12. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    Spanner, J.

    2015-01-01

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

  13. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

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

    2012-09-01

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

  14. The Conceptual Design of Innovative Safe PWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  15. Materials performance in operating PWR steam generators

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

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

  16. Seismic behaviour of PWR fuel assemblies model and its validation

    International Nuclear Information System (INIS)

    Queval, J.C.; Gantenbein, F.; Brochard, D.; Benjedidia, A.

    1991-01-01

    The validity of the models simulating the seismic behaviour of PWR cores can only be exactly demonstrated by seismic testing on groups of fuel assemblies. Shake table seismic tests of rows of assembly mock-ups, conducted by the CEA in conjunction with FRAMATOME, are presented in reference /1/. This paper addresses the initial comparisons between model and test results for a row of five assemblies in air. Two models are used: a model with a single beam per assembly, used regularly in accident analyses, and described in reference /2/, and a more refined 2-beam per assembly model, geared mainly towards interpretation of test results. The 2-beam model is discussed first, together with parametric studies used to characterize it, and the study of the assembly row for a period limited to 2 seconds and for different excitation levels. For the 1-beam model assembly used in applications, the row is studied over the total test time, i.e twenty seconds, which covers the average duration of the core seismic behaviour studies, and for a peak exciting acceleration value at 0.4 g, which corresponds to the SSE level of the reference spectrum

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  18. The design of a compact integral medium size PWR

    International Nuclear Information System (INIS)

    Shirvan, Koroush; Hejzlar, Pavel; Kazimi, Mujid S.

    2012-01-01

    Highlights: ► We model the IRIS reactor in RELAP5 and VIPRE codes. ► We use Printed Circuit Heat Exchangers and internally and externally cooled fuel pins in IRIS. ► We increase the IRIS power by 50% and demonstrate adequate safety performance. ► We show significant potential gain in economics for any integral PWR reactor design. - Abstract: Integral reactors have been proposed in recent years as a means to eliminate loss of coolant events, and reduce the number of large vessels of a nuclear power plant. In this paper the focus on how to further increase the power that can be derived from a given vessel volume. The example is applied to the International Reactor Innovative and Secure (IRIS), a medium size, light water reactor rated at 1000 MWt. The IRIS is an integral design containing all pumps and steam generators along with a traditional PWR core inside the reactor vessel. IRIS was designed with 8 Once-Through Helically Coiled Steam Generators (OTHSG), located above the core, in an annular region between the riser and the pressure vessel wall. This work examines ideas to increase its power output in the same vessel size while maintaining or improving the safety margins. The combination of Printed Circuit Heat Exchangers (PCHE) and Internally and EXternally cooled Annular Fuel (IXAF) is proposed to implement such improvement in otherwise the reference IRIS design. Safety implications of such steam generator and fuel design changes for the same reactor size are examined, under both steady state and transients, using the RELAP5 and VIPRE codes. It is found that the IRIS reactor power can be increased by 50% by using the PCHE and IXAF. The proposed design is found to be less expensive per unit electric power produced, these improvements and analyses can be applied to any integral reactor design.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  20. Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

    2015-07-01

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  1. Study on quality control measures of static casting main pipe in PWR nuclear power plant

    International Nuclear Information System (INIS)

    Jiang Zhenbiao; Li Guanying; Liu Zhicheng

    2013-01-01

    This study analyzes the main reasons which impact the quality of primary pipe static casting elbows in PWR-M310 nuclear power plant. The quality control measures are developed from the election and inspection of material, improving sand production and casting process, improving lean management of personnel. The static casting defects of primary pipe elbows for Fuqing Unit 1 and 2 were down to less than 50% of the former project. The quality of static casting for the primary pipe elbows was significantly improved. Moreover, the implementation saves human resources and financing to repair casting defects, and also helps to win the delivery schedule. The quality control measures are good reference for improving primary pipe casting process. This study provides valuable experience for further study of improving the quality of static casting for the primary pipe of PWR nuclear power plant. (authors)

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

    International Nuclear Information System (INIS)

    Illy, H.

    1985-03-01

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

  3. Thermal Response of the 21-PWR Waste Package to a Fire Accident

    International Nuclear Information System (INIS)

    F.P. Faucher; H. Marr; M.J. Anderson

    2000-01-01

    The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation (Attachment IV) is that of the potential design of the type of waste package considered in this calculation. The procedure AP-3.12Q.Calculations (Reference 1), and the Development Plan (Reference 24) are used to develop this calculation

  4. Study of anticipated transient without scram for PWR

    International Nuclear Information System (INIS)

    Pu Jilong.

    1985-01-01

    Anticipated Transient Without Scram (ATWS) of PWR, the one of the 'Unresolved Safety Issue' with NRC, has been investigated for many years. The latest analysis done by the author considers the PWR's inherent stability and long-term performence under the condition of ATWS combined with SBLOCA and studies the sensitivity of several assumptions, which shows positive results

  5. Pushing back the boundaries of PWR fuel performance

    International Nuclear Information System (INIS)

    Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

    1985-01-01

    In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

  6. Highlights of the French program on PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pages, J P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1997-12-01

    The presentation reviews the French programme on PWR fuel including the overall results of the year 1996 for nuclear operation; fuel management and economy; French nuclear electricity generation sites; production of nuclear generated electricity; energy availability of the 900 and 1,300 Mw PWR units; average radioactive liquid releases excluding tritium per unit; plutonium recycling experience.

  7. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

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

    International Nuclear Information System (INIS)

    1980-10-01

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

  9. Oxygen control in PWR secondary coolant: Final report

    International Nuclear Information System (INIS)

    Oliker, I.; Shaikh, B.

    1988-12-01

    The objective of the study is to assess the technical aspects of utilizing direct contact heaters in the condensate train and a bubbling deaerating device in the condenser in order to improve the deaeration in the feedwater system, and develop cost estimates for such a plant retrofit. A reference PWR plant has been used to establish a basis for development of retrofit requirements and system costs. Emphasis has been placed on retrofitting the existing low pressure heaters in the condenser neck into direct contact heaters in order to improve the condensate deaeration at the very beginning of the condensate-feedwater cycle. Two basic designs of direct contact heater are discussed and their technical benefits are described. Required plant modifications have been developed and improvement in heat rates have been evaluated. Cost estimates have been developed for such a retrofit. Turbine protection against water induction has been addressed at length including recommendations on various protection measures. Incorporation of a bubbling deaerating device into the reference plant has been evaluated, and its cost estimate has been developed. 4 refs., 18 figs., 1 tab

  10. Chemical decontamination solutions: Effects on PWR equipment

    International Nuclear Information System (INIS)

    Pezze, C.M.; Colvin, E.R.; Aspden, R.G.

    1992-01-01

    A critical objective for the nuclear industry is the reduction of personnel exposure to radiation. Reductions have been achieved through industry's radiation management programs including training and radiation awareness concepts. Increased plant maintenance and higher radiation fields at many sites continue to raise concerns. To alleviate the radiation exposure problem, the sources of radiation which contribute to personnel exposure must be removed from the plant. A feasible was of significantly reducing these sources from a Pressurized Water Reactor (PWR) is to chemically decontaminate the entire reactor coolant system (RCS). A program was conducted to determine the technical acceptability of using certain dilute chemical solvent processes for full RCS chemical decontamination. The two processes evaluated were CAN-DEREM and LOMI. The purpose of the program was to define and complete a systematic evaluation of the major issues that need to be addressed for the successful decontamination of the entire RCS and affected portions of the auxiliary systems of a four-loop PWR system. A test program was designed to evaluate the corrosion effects of the two decontamination processes under expected plant conditions. Materials and sample configurations dictated by generic PWR components were evaluated. The testing also included many standard corrosion coupons. The test data were then used to assess the impact of chemical decontamination on the physical condition and operability of the components, equipment and mechanical systems that make up the RCS. An overview of the test program, sample configurations, data and engineering evaluations is presented. The data demonstrate that through detailed engineering evaluations of corrosion data and equipment function, the impact of full RCS chemical decontamination on plant equipment is established

  11. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

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

  12. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    Sotoma, H.

    1973-01-01

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

  13. Technical specifications for PWR secondary water chemistry

    International Nuclear Information System (INIS)

    Weeks, J.R.; van Rooyen, D.

    1977-08-01

    The bases for establishing Technical Specifications for PWR secondary water chemistry are reviewed. Whereas extremely stringent control of secondary water needs to be maintained to prevent denting in some units, sound bases for establishing limits that will prevent stress corrosion, wastage, and denting do not exist at the present time. This area is being examined very thoroughly by industry-sponsored research programs. Based on the evidence available to date, short term control limits are suggested; establishment of these or other limits as Technical Specifications is not recommended until the results of the research programs have been obtained and evaluated

  14. Technical basis for PWR emergency plans forming

    International Nuclear Information System (INIS)

    L'Homme, A.; Manesse, D.; Gauvain, J.; Crabol, B.

    1989-10-01

    Our speech first summarizes the french approach concerning the management of severe accidents which could occur on PWR stations. Then it defines the source-term which is being used as a general support for elaborating the emergency plans devoted to the protection of the population. It describes next the consequences of this source-term on the site and in the environment, which constitute the technical bases for defining actions of utilities and concerned authorities. It gives lastly information on the present status of the different emergency plans and the complementary work undertaken to improve them [fr

  15. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  16. Industrywide survey of PWR organics. Final report

    International Nuclear Information System (INIS)

    Richards, J.E.; Byers, W.A.

    1986-07-01

    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

  17. Microcomputer simulation of PWR power plant pressurizer

    International Nuclear Information System (INIS)

    Araujo, L.R.A. de; Calixto Neto, J.; Martinez, A.S.; Schirru, R.

    1990-01-01

    It is presented a method for the simulation of the pressurizer behavior of a PWR power plant. The method was implanted in a microcomputer, and it considers all the devices for the pressure control (spray and relief valves, heaters, controller, etc.). The physical phenomena and the PID (Proportional + Integral + Derivative) controller were mathematically represented by linear relations, uncoupled, discretized in the time. There are three different algorithms which take into account the non-linear effects introduced by the variation of the physical properties due to the temperature and pressure, and also the mutual effects between the physical phenomena and the PID controller. (author)

  18. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    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

  19. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

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

  20. PWR life time: the EDF project

    International Nuclear Information System (INIS)

    Noel, R.; Reynes, L.; Mercier, J.P.

    1987-01-01

    Operating a very large number of standardized PWR units which supply today 70% of French power generation, Electricite de France is highly interested in getting the best estimate of the safe and economical life of these plants. An extensive program of work has been undertaken in this respect. The studies have first to go through all available data on aging process, survey and maintenance of a limited number of major components. This review will lead to recommendation of complementary work in these fields. The first conclusions are that these units are able to perform a long service time, under provision of careful survey and maintenance [fr

  1. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

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

  2. Environmental surveillance of PWR power stations

    International Nuclear Information System (INIS)

    Conti, M.

    1980-01-01

    The action of Electricite de France with respect to the environment of PWR nuclear power stations is essentially centred on prevention. Controls are carried out at two levels: - before the power station goes on stream (radioecological study), - when the power station is operational. The purpose of the controls effected on the radioactive effluents and the environment is to check that the maximum discharge rate stipulated in the corresponding orders is complied with and to ensure that there are no anomalies in the environment [fr

  3. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  4. Recent development in PWR zinc injection

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Deleume, J.

    2007-11-01

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

  6. PWR reactor pressure vessel internals license renewal industry report; revision 1. Final report

    International Nuclear Information System (INIS)

    Schwirian, R.; Robison, G.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). Pressurized water reactor (PWR) reactor pressure vessel (RPV) internals designed by all three U.S. PWR nuclear steam supply system vendors have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits; inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these internals can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR reactor pressure vessel internals for license renewal

  7. Determination of welding parameters for execution of weld overlayer on PWR nuclear reactor nozzles

    International Nuclear Information System (INIS)

    Ribeiro, Gabriela M.; Lima, Luciana I.; Quinan, Marco A.; Schvartzman, Monica M.

    2009-01-01

    In the PWR reactors, nickel based dissimilar welds have been presented susceptibilities the stress corrosion (S C). For the mitigation the problem a deposition of weld layers on the external surface of the nozzle is an alternative, viewing to provoke the compression of the region subjected to S C. This paper presents a preliminary study on the determination of welding parameters to obtain these welding overlayers. Welding depositions were performed on a test piece welded with nickel 182 alloy, simulating the conditions of a nozzle used in a PWR nuclear power plant. The welding process was the GTAW (Gas Tungsten Arc Welding), and a nickel 52 alloy as addition material. The overlayers were performed on the base metals, carbon steel an stainless steel, changing the welding parameters and verifying the the time of each weld filet. After that, the samples were micro structurally characterized. The macro structures and the microstructures obtained through optical microscopy and Vickers microhardness are presented. The preliminary results make evident the good weld quality. However, a small weld parameters influence used in the base material microstructure (carbon steel and stainless steel). The obtained results in this study will be used as reference in the construction of a mock up which will simulate all the conditions of a pressurizer nozzle of PWR reactor

  8. Data assimilation and PWR primary measurement

    International Nuclear Information System (INIS)

    Mercier, Thibaud

    2015-01-01

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

  9. Advanced high conversion PWR: preliminary analysis

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  10. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    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. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  12. Workers doses in central European PWR NPPs

    International Nuclear Information System (INIS)

    Janzekovic, H.; Krizman, M.

    2003-01-01

    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. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    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. Comparison of Serpent and HELIOS-2 as applied for the PWR few-group cross section generation

    International Nuclear Information System (INIS)

    Fridman, E.; Leppaenen, J.; Wemple, C.

    2013-01-01

    This paper discusses recent modifications to the Serpent Monte Carlo code methodology and related to the calculation of few-group diffusion coefficients and reflector discontinuity factors The new methods were assessed in the following manner. First, few-group homogenized cross sections calculated by Serpent for a reference PWR core were compared with those generated 1 commercial deterministic lattice transport code HELIOS-2. Second, Serpent and HELIOS-2 fe group cross section sets were later employed by nodal diffusion code DYN3D for the modeling the reference PWR core. Finally, the nodal diffusion results obtained using the both cross section sets were compared with the full core Serpent Monte Carlo solution. The test calculations show that Serpent can calculate the parameters required for nodal analyses similar to conventional deterministic lattice codes. (authors)

  16. The simulation research for the dynamic performance of integrated PWR

    International Nuclear Information System (INIS)

    Yuan Jiandong; Xia Guoqing; Fu Mingyu

    2005-01-01

    The mathematical model of the reactor core of integrated PWR has been studied and simplified properly. With the lumped parameter method, authors have established the mathematical model of the reactor core, including the neutron dynamic equation, the feedback reactivities model and the thermo-hydraulic model of the reactor. Based on the above equations and models, the incremental transfer functions of the reactor core model have been built. By simulation experimentation, authors have compared the dynamic characteristics of the integrated PWR with the traditional dispersed PWR. The simulation results show that the mathematical models and equations are correct. (authors)

  17. Integrity of PWR pressure vessels during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

  18. Integrity of PWR pressure vessels during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-15

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

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

    Directory of Open Access Journals (Sweden)

    Huiun Ha

    2017-08-01

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

  1. Qualification tests for PWR control element drive mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Yong; Jin, Choon Eon; Choi Suhn [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    It is necessary to perform the qualification test for the magnetic jack type CEDM to show the design compatibility because the CEDM is composed of many mechanical and electrical components complicatedly. ABB-CE performed various qualification tests during the development of the System80 CEDM to which Korea Standard Nuclear Plant (KSNP) CEDM referred. The qualification test for the CEDM is classified into the performance test and the dynamic test. The performance test is to verify operability of the CEDM, and the dynamic test is to find dynamic characteristics and to verify the structural integrity if the CEDM for the seismic accidents. Described in this report are the test requirements, the test facilities and the test methods for the performance and the dynamic qualification tests of the PWR magnetic jack type CEDM. The impacts of the design changes in the Korea Next Generation Reactor (KNGR) on the KSNP CEDM were analyzed to present the necessity for the tests. This report also proposes the facilities to perform the tests in KAERI including reasonable schedule for the tests. Attached to this report is the summary of qualification tests of System 80 CEDM performed by ABB-CE. 20 figs., 16 tabs., 21 refs. (Author) .new.

  2. Qualification tests for PWR control element drive mechanism

    International Nuclear Information System (INIS)

    Kim, In Yong; Jin, Choon Eon; Choi Suhn

    1996-01-01

    It is necessary to perform the qualification test for the magnetic jack type CEDM to show the design compatibility because the CEDM is composed of many mechanical and electrical components complicatedly. ABB-CE performed various qualification tests during the development of the System80 CEDM to which Korea Standard Nuclear Plant (KSNP) CEDM referred. The qualification test for the CEDM is classified into the performance test and the dynamic test. The performance test is to verify operability of the CEDM, and the dynamic test is to find dynamic characteristics and to verify the structural integrity if the CEDM for the seismic accidents. Described in this report are the test requirements, the test facilities and the test methods for the performance and the dynamic qualification tests of the PWR magnetic jack type CEDM. The impacts of the design changes in the Korea Next Generation Reactor (KNGR) on the KSNP CEDM were analyzed to present the necessity for the tests. This report also proposes the facilities to perform the tests in KAERI including reasonable schedule for the tests. Attached to this report is the summary of qualification tests of System 80 CEDM performed by ABB-CE. 20 figs., 16 tabs., 21 refs. (Author) .new

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  4. Improvements of nuclear fuel management in pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1978-07-01

    The severe variations to which the different elements contributing to the determination of the fuel cycle cost are subjected have led to a reopening of the problem of ''optimization'' of nuclear fuel management. The increase in costs of uranium ore, isotope separation work units (swu), reprocessing, the political implications of proliferation associated with the employment of reprocessing operations have been at the origin of a reassessment of present-day management. It therefore appeared to be appropriate to study variants with respect to a reference mode represented by the management of the PWR 900 MWe systems, without burnable poison in the cycle at equilibrium (Case 3 of Table 1). In order to obtain a complete view of impacts of such modifications, computations were carried out as far as the appraisal of the cycle cost and with reprocessing. There has likewise been added to this the estimate of the gain anticipated from certain improvements in the neutron balance contributed at the level of the lattice

  5. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

    2016-10-15

    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

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

    International Nuclear Information System (INIS)

    Llambias, J.M.

    1993-01-01

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

  7. Radionuclide release calculations for selected severe accident scenarios. PWR, ice condenser design

    Energy Technology Data Exchange (ETDEWEB)

    Denning, R S; Gieseke, J A; Cybulskis, P; Lee, K W; Jordan, H; Curtis, L A; Kelly, R F; Kogan, V; Schumacher, P M

    1986-07-01

    This report presents results of analyses of the environmental releases of fission products (source terms) for severe accident scenarios in a pressurized water reactor with an ice-condenser containment. The analyses were performed to support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP) which is being undertaken for the U.S. Nuclear Regulatory Commission by Sandia National Laboratories. In the SARRP program, risk estimates are being generated for a number of reference plant designs. The Sequoyah Plant has been used in this study as an example of a PWR ice-condenser plant. (author)

  8. Optimization of deterministic based design of the PWR 1000 MW by aid of PSA

    International Nuclear Information System (INIS)

    Feigel, A.; Fabian, H.

    1987-01-01

    PSA was used to optimize the determinstic based design of the PWR 1000 MW. For this three reference accidents which are known to be the covering ones from previous valuations were investigated in detail. On basis of these accidents the integral core damage frequency has been estimated to be about 2 E-5/a. This result reflects a sufficient safety level and thus the quality of the requirement which has to be used for the design. Nevertheless the influence of some plant modifications was estimated in addition. It shows that especially the consideration of a modul with a diverse power generator results in a more balanced design on an increased safety level. (orig.)

  9. Level-1 seismic probabilistic risk assessment for a PWR plant

    International Nuclear Information System (INIS)

    Kondo, Keisuke; Nishio, Masahide; Fujimoto, Haruo; Ichitsuka, Akihiro

    2014-01-01

    In Japan, revised Seismic Design Guidelines for the domestic light water reactors was published on September 19, 2006. These new guidelines have introduced the purpose to confirm that residual risk resulting from earthquake that exceeds the design limit seismic ground motion (Ss) is sufficiently small, based on the probabilistic risk assessment (PRA) method, in addition to conventional deterministic design base methodology. In response to this situation, JNES had been working to improve seismic PRA (SPRA) models for individual domestic light water reactors. In case of PWR in Japan, total of 24 plants were grouped into 11 categories to develop individual SPRA model. The new regulatory rules against the Fukushima dai-ichi nuclear power plants' severe accidents occurred on March 11, 2011, are going to be enforced in July 2013 and utilities are necessary to implement additional safety measures to avoid and mitigate severe accident occurrence due to external events such as earthquake and tsunami, by referring to the results of severe accident study including SPRA. In this paper a SPRA model development for a domestic 3-loop PWR plant as part of the above-mentioned 11 categories is described. We paid special attention to how to categorize initiating events that are specific to seismic phenomena and how to confirm the effect of the simultaneous failure probability calculation model for the multiple components on the result of core damage frequency evaluation. Simultaneous failure probability for multiple components has been evaluated by power multiplier method. Then tentative level-1 seismic probabilistic risk assessment (SPRA) has been performed by the developed SPSA model with seismic hazard and fragility data. The base case was evaluated under the condition with calculated fragility data and conventional power multiplier. The difference in CDF between the case of conventional power multiplier and that of power multiplier=1 (complete dependence) was estimated to be

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

    International Nuclear Information System (INIS)

    Hursin, M.; Kochunas, B.; Downar, T. J.

    2008-01-01

    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

  11. A universal PWR spectral history correction

    International Nuclear Information System (INIS)

    Hutt, P.K.; Nunn, D.L.

    1989-01-01

    The accuracy of a form of universal correction for the difference between depletion conditions assumed in PWR assembly lattice calculations and those experienced in a reactor burn-up is investigated. The correction is based on lattice calculations in which only one such depletion history difference, depletion at two different water densities, is explicitly represented by lattice calculations. The assumption is made that other historical effects bear the same relationship to an appropriate time-average of the two-group neutron flux spectrum. The correction is shown to be accurate for the most important historical effects, depletion with burnable absorbers inserted, control rods inserted or at a different soluble boron level, in addition to density itself. The correction is less accurate for representing depletion at a different fuel or coolant temperature but even in these cases gives an improvement over no correction. In addition it is argued that these historic temperature effects are likely to be of minor importance. (author)

  12. Chemical cleaning of nuclear (PWR) steam generators

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Mundis, J.A.

    1982-01-01

    This paper reports on a significant research program sponsored by a group of utilities (the Steam Generator Owners Group), which was undertaken to develop a process to chemically remove corrosion product deposits from the secondary side of pressurized water reactor (PWR) power plant steam generators. Results of this work have defined a process (solvent system and application methods) that is capable of removing sludge and tube-to-tube support plate crevice corrosion products generated during operation with all-volatile treatment (AVT) water chemistry. Considers a plant-specific test program that includes all materials in the steam generator to be cleaned and accounts for the physical locations (proximity and contact) of those materials. Points out that prior to applying the process in an operational unit, the utility, with the participation of the NSSR vendor, must define allowable total corrosion to the materials of construction of the unit

  13. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    Smuc, T.; Pevec, D.

    1994-01-01

    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

  14. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.; Driscoll, M.J.; Lanning, D.D.

    1979-08-01

    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 - 235 U/UO 2 : Pu/ThO 2 : 233 U/ThO 2 - and the conventional recycle-mode uranium system - 235 U/UO 2 : Pu/UO 2 . 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

  15. 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)

    1999-12-31

    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)

  16. Zebra: An advanced PWR lattice code

    International Nuclear Information System (INIS)

    Cao, L.; Wu, H.; Zheng, Y.

    2012-01-01

    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)

  17. Simplified model of a PWR primary circuit

    International Nuclear Information System (INIS)

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

    1988-07-01

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

  18. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  19. 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)

  20. 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)

  1. Development of advanced PWR steam generator

    International Nuclear Information System (INIS)

    Saito, Itaru; Nakamura, Tomomichi

    1999-01-01

    In response to the increased power of the advanced PWR, it is necessary to develop a steam generator (SG) which has a large capacity with high performance and high reliability as well as being economical to produce. In this paper, the development of the design of a new SG for the advanced PWRs is described and compared with the design of a conventional SG. Moreover, an outline of a seismic verification test for the U-bend tube bundle which includes advanced anti-vibration bars (AVB) which are very important is described. As a result, it was verified that the bundle has sufficient strength and a relatively high attenuation to seismic loads. These results will be reflected in the detailed design of advanced AVBs. (author)

  2. Radiation embrittlement of PWR vessel supports

    International Nuclear Information System (INIS)

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

    1989-01-01

    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

  3. Interpretation of out of line control rod experiments for 1300 MWE PWR

    Energy Technology Data Exchange (ETDEWEB)

    Leroy, J.L.; Garcia-Fernandez, L.

    1988-01-01

    The present note summarizes the studies we performed recently in order to search a 2D reconstruction procedure for the 1300 MWE PWR power shape, starting from data coming out from thermocouples placed on several fuel assemblies. In classical PWR design, only a few assemblies are equipped with measurement devices, so that it is necessary to interpolate among measure points in order to obtain a complete coverage of the core. A mathematical approach based on the splitting of the power into a reference steady state nominal shape and some ''influence'' and harmonic functions was chosen. The reference steady state power shape, which corresponds to the full power operating mode, is obtained via direct mobile chamber measurements. The perturbations due to the control rod movements are accounted for by specific ''influence'' functions: moreover, harmonics are used to reconstruct the minor effects due to xenon tilts, rod out of line positions and all actual mechanical and thermohydraulic inhomogeneities. The weighting coefficients of the functions are evaluated by a least square method, starting from the distribution of the deviations among the measurements and the reference values.

  4. A comparison of the BUGLE-80, SAILOR, and ELXSIR neutron cross-section libraries for PWR pressure vessels surveillance dosimetry and shielding applications

    International Nuclear Information System (INIS)

    Basha, H.S.; Manahan, M.P.

    1992-01-01

    In this paper three multigroup neutron cross-section libraries are used in synthesized three-dimensional discrete ordinates transport analyses to investigate their similarities, differences, and results for pressurized water reactor (PWR) pressure vessel surveillance dosimetry and shielding applications. The calculated-to-experimental (C/E) rations and the calculated reaction rates of several fast reactions are compared for the BUGLE-80, SAILOR, and ELXSIR cross-section libraries at the 97-deg surveillance capsule of the San Onofre Nuclear Generation Station Unit 2 (SONGS-2) and at the 90- and 97-deg (C/E ratios only) cavity dosimetry locations for another PWR (referred to as Reactor X)

  5. Changes in 900 MW PWR alarm processing policy

    Energy Technology Data Exchange (ETDEWEB)

    Pont, M [Electricite de France, Generation and Transmission, Nuclear Power Plant Operations, Paris (France)

    1997-09-01

    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.

  6. Changes in 900 MW PWR alarm processing policy

    International Nuclear Information System (INIS)

    Pont, M.

    1997-01-01

    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

  7. Deboration in nuclear stations of the PWR type

    International Nuclear Information System (INIS)

    1978-01-01

    Reactivity control in nuclear power stations of the PWR type is realised with boric acid. A method to concentrate boric acid without an evaporator has been studied. A flow-sheet with reverse osmosis is proposed. (author)

  8. Severe accident considerations for modern KWU-PWR plants

    International Nuclear Information System (INIS)

    Eyink, J.

    1987-01-01

    In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

  9. Dose rate evaluation after accident in a PWR

    International Nuclear Information System (INIS)

    Cladel, C.; Duchemin, B.; Le Dieu de Ville, A.; Nimal, B.; Nimal, J.C.; Evrard, J.M.

    1983-05-01

    A calculation scheme for the gamma radiation dose rate after accident in a PWR is presented. These studies use a fine description of the geometry and of the fission product inventory. Some results are given and some improvements are planned

  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. Ultrasonic inspection for testing the PWR fuel rod endplug welds

    International Nuclear Information System (INIS)

    Pillet, C.; Destribats, M.T.; Papezyk, F.

    1976-01-01

    A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

  12. PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-04-01

    rinci. Kata kunci: pemodelan, Generasi III+, RELAP5.   Westinghouse’s AP1000 reactor design is the first Generation III+ nuclear power reactor to receive final design approval from the U.S. Nuclear Regulatory Commission (NRC. Currently, the China’s utilities are starting construction several units of AP1000 on two selected sites for scheduled operation in 2013–2015. The AP1000, based on proven technology of Westinghouse-designed PWR with enhancement on the passive safety system, could be considered to be built in Indonesia referring to the requirements of government regulation No. 43/2006 regarding the Nuclear Reactor Licensing. To be accepted by the regulation agency, the design needs to be verified by independent Technical Support Organization (TSO, which can be done using RELAP5 computer code as accident analyses. Currently, NPP safety accident analysis is performed for PWR 1000 MWe of generation II or conventional type. Considering that nowadays references about the technology of AP1000 that includes passive safety technology has been available and assessed, a modeling activity used for future accident analyzes is introduced. Method for developing the model refers to IAEA guide consisting of plant data collection, engineering data and input deck development, and verification and validation of input data. The model developed should be considered preliminary but has been generally representing the AP1000 systems as the basic model. The model has been verified and validated by comparing thermalhidraulic parameter responses with design data in references with ± 13% deviation except for core pressure drop with 13% lower than design. As a basic model, the input deck is ready for further development by integrating safety system, protection system and control system model specified for AP1000 for purposes of safety simulation in detailed way. Keywords: Modeling, Generation III+ , RELAP5.

  13. Model for transient simulation in a PWR steam circuit

    International Nuclear Information System (INIS)

    Mello, L.A. de.

    1982-11-01

    A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt

  14. GO evaluation of a PWR spray system. Final report

    International Nuclear Information System (INIS)

    Long, W.T.

    1975-08-01

    GO is a reliability analysis methodology developed over the years from 1960 to the present by Kaman Sciences Corporation, Colorado Springs, Colorado. In this report the GO methodology is presented and its application demonstrated by performing a reliability analysis of a conceptual PWR Containment Spray System. Certain numerical results obtained are compared with those of a prior fault tree analysis of the same system as documented in the 11 January 1973 draft report, A Fault Tree Evaluation of a PWR Spray System

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-01

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

  16. PWR fuel of high enrichment with erbia and enriched gadolinia

    International Nuclear Information System (INIS)

    Bejmer, Klaes-Håkan; Malm, Christian

    2011-01-01

    Today standard PWR fuel is licensed for operation up to 65-70 MWd/kgU, which in most cases corresponds to an enrichment of more than 5 w/o "2"3"5U. Due to criticality safety reason of storage and transportation, only fuel up to 5 w/o "2"3"5U enrichment is so far used. New fuel storage installations and transportation casks are necessary investments before the reactivity level of the fresh fuel can be significantly increased. These investments and corresponding licensing work takes time, and in the meantime a solution that requires burnable poisons in all pellets of the fresh high-enriched fuel might be used. By using very small amounts of burnable absorber in every pellet the initial reactivity can be reduced to today's levels. This study presents core calculations with fuel assemblies enriched to almost 6 w/o "2"3"5U mixed with a small amount of erbia. Some of the assemblies also contain gadolinia. The results are compared to a reference case containing assemblies with 4.95 w/o "2"3"5U without erbia, utilizing only gadolinia as burnable poison. The comparison shows that the number of fresh fuel assemblies can be reduced by 21% (which increases the batch burnup by 24%) by utilizing the erbia fuel concept. However, increased cost of uranium due to higher enrichment is not fully compensated for by the cost gain due to the reduction of the number assemblies. Hence, the fuel cycle cost becomes slightly higher for the high enrichment erbia case than for the reference case. (author)

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

    International Nuclear Information System (INIS)

    Osakabe, Masahiro

    1989-01-01

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

  18. Design of a PWR for long cycle and direct recycling of spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

    2015-12-15

    Highlights: • Single-batch loading PWR with a new fuel assembly for 36 calendar months cycle was designed. • The new fuel assembly is constructed from a number of CANDU fuel bundles. • This design enables to recycle the spent fuel directly in CANDU reactors for high burnup. • Around 56 MWd/kgU burnup is achieved from fuel that has average enrichment of 4.8 w/o U-235 using this strategy. • Safety parameters such as the power distribution and CANDU coolant void reactivity were considered. - Abstract: In a previous work, a new design was proposed for the Pressurized Water Reactor (PWR) fuel assembly for direct use of the PWR spent fuel without processing. The proposed assembly has four zircaloy-4 tubes contains a number of 61-element CANDU fuel bundles (8 bundles per tube) stacked end to end. The space between the tubes contains 44 lower enriched UO{sub 2} fuel rods and 12 guide tubes. In this paper, this assembly is used to build a single batch loading 36-month PWR and the spent CANDU bundles are recycled in the on power refueling CANDU reactors. The Advanced PWR (APWR) is considered as a reference design. The average enrichment in the core is 4.76%w U-235. IFBA and Gd{sub 2}O{sub 3} as burnable poisons are used for controlling the excess reactivity and to flatten the power distribution. The calculations using MCNPX showed that the PWR will discharge the fuel with average burnup of 31.8 MWd/kgU after 1000 effective full power days. Assuming a 95 days plant outage, 36 calendar months can be achieved with a capacity factor of 91.3%. Good power distribution in the core is obtained during the cycle and the required critical boron concentration is less than 1750 ppm. Recycling of the discharged CANDU fuel bundles that represents 85% of the fuel in the assembly, in CANDU-6 or in 700 MWe Advanced CANDU Reactor (ACR-700), an additional burnup of about 31 or 26 MWd/kgU burnup can be achieved, respectively. Averaging the fuel burnup on the all fuel in the PWR

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  1. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP

    International Nuclear Information System (INIS)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.

    2013-10-01

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

  2. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  3. 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)

  4. Maintenance technologies for SCC of PWR

    International Nuclear Information System (INIS)

    Okimura, Koji; Hori, Nobuyuki; Kanzaki, Hiroshi; Tokuhisa, Kiichi; Kamo, Kazuhiko; Kurokawa, Masaaki

    2007-01-01

    The recent technologies of test, relaxation of deterioration, repairing and change of materials are explained for safe and stable operation of pressurized water reactor (PWR). Stress corrosion cracking (SCC) is originated by three factors such as materials, stress and environment. The eddy current test (ECT) method for the stream generator pipe and the ultrasonic test method for welding part of pipe were developed as the test technologies. Primary water stress corrosion cracking (PWSCC) of Inconel 600 in the welding part is explained. The shot peening of instrument in the gas, the water jet peening of it in water, and laser irradiation on the surface are illustrated as some examples of improvement technology of stress. The cladding of Inconel 690 on Inconel 600 is carried out under the condition of environmental cut. Total or some parts of the upper part of reactor, stream generator and structure in the reactor are changed by the improvement technologies. Changing Inconel 600 joint in the exit pipe of reactor with Inconel 690 is illustrated. (S.Y.)

  5. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  6. MELCOR/VISOR PWR desktop simulator

    International Nuclear Information System (INIS)

    With, Anka de; Wakker, Pieter

    2010-01-01

    Increasingly, there is a need for a learning support and training tool for nuclear engineers, utilities and students in order to broaden their understanding of advanced nuclear plant characteristics, dynamics, transients and safety features. Nuclear system analysis codes like ASTEC, RELAP5, RETRAN and MELCOR provide calculation results of and visualization tools can be used to graphically represent these results. However, for an efficient education and training a more interactive tool such as a simulator is needed. The simulator connects the graphical tool with the calculation tool in an interactive manner. A small number of desktop simulators exist [1-3]. The existing simulators are capable of representing different types of power plants and various accident conditions. However, they were found to be too general to be used as a reliable plant-specific accident analysis or training tool. A desktop simulator of the Pressurized Water Reactor (PWR) has been created under contract of the Dutch nuclear regulatory body (KFD). The desktop simulator is a software package that provides a close to real simulation of the Dutch nuclear power plant Borssele (KCB) and is used for training of the accident response. The simulator includes the majority of the power plant systems, necessary for the successful simulation of the KCB plant during normal operation, malfunctions and accident situations, and it has been successfully validated against the results of the safety evaluations from the KCB safety report. (orig.)

  7. Summary of PWR leak detection studies

    International Nuclear Information System (INIS)

    Cho, J.H.; Elia, F.A. Jr.

    1986-01-01

    Thermal-hydraulic analysis can be used to determine the location and magnitude of leaks inside and location of leaks outside a pressurized water reactor (PWR) containment as required by plant technical specifications. The major advantage of this detection method is that it minimizes radiation exposure of maintenance personnel because most of the leak detection process is performed from the control room outside containment. Plant-specific analyses are utilized to predict change in parameters such as local dew point temperature, relative humidity, dry bulb temperature, and flow rate to sump for various leak rates and enthalpies. These parameter responses are then programmed into the plant computer and instrumentation is provided for area monitoring. The actual inputs are continuously monitored and compared to the predicted plant responses to identify the leak location and quantify the leak. This study concludes that a system that monitors dew point (or relative humidity) and dry bulb temperature changes together with the flow rate to the sump will provide the capability to both locate and quantify a leak inside a containment, while a system that monitors dew point temperature (or relative humidity) changes will provide the capability to locate a leak outside a containment

  8. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1984-01-01

    Hydrogen generation during a PWR LOCA has been estimated for design basis accident and for two more severe hypothetical accidents. Hydrogen production during design basis accident is a rather slow mechanism, allowing in the worst case, 15 days to connect a hydrogen recombining unit to the containment atmosphere monitoring system. Hydrogen generated by steam oxidation during more severe hypothetical accidents was found limited by steam availability and fuel melting phenomena. Uncertainty is, however, still remaining on corium-zirconium-steam interaction. In the worst case, calculations lead to the production of 500 kg of hydrogen, thus leading to a volume concentration of 15% in containment atmosphere, assuming homogeneous hydrogen distribution within the reactor building. This concentration is within flammability limits but not within detonation limits. However, hydrogen detonation due to local hydrogen accumulation cannot be discarded. A major uncertainty subsisting on hydrogen hazard is hydrogen distribution during the first hours of the accident. This point determines the effects and consequences of local detonation or deflagration which could possibly be harmful to safeguard systems, or induce missile generation in the reactor building. As electrical supply failures are identified as an important contributor to severe accident risk, corrective actions have been taken in France to improve their reliability, including the installation of a gas turbine on each site to supplement the existing sources. These actions are thus contributing to hydrogen hazard reduction

  9. Analysis of reactivity accidents in PWR'S

    International Nuclear Information System (INIS)

    Camous, F.; Chesnel, A.

    1989-12-01

    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

  10. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    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

  11. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  12. Multicriteria analysis of public protection in PWR's

    International Nuclear Information System (INIS)

    Lombard, J.

    1986-09-01

    In order to manage a risk efficiently and to reach the ALARA level of protection, the best possible protection options must be employed. As the available resources are limited, it is not always possible to choose those options that minimize the risk, therefore a compromise must be made between risks and safety expenses. When the choice is difficult or complex, finding such a compromise can be facilitated by resorting to a decision aiding method which allows the assessment of the respective advantages of the various protection options considered. The multicriteria methods employ successive comparisons. Instead of searching for a final indicator expressing the performance of each option they compare all option pairs in order to determine if the gap between their respective advantages and disadvantages is sufficient to estimate that one option of the each pair is better than the other. Instead of judging each option globally these methods evaluate the advantages and disadvantages associated with the eventual choice of an option as compared with the others. These differential and comparative approach gives more flexibility and allows the introduction of qualitative criteria. The method presented here (Electre 3), one of the most recent ones, allows a multicriteria comparison of a set of options keeping into account the uncertainties associated with the options or the preferences. In order to illustrate this method a simple example (4 options, 4 criteria) dealing with a PWR liquid releases treatment system, is taken up

  13. Modeling of PWR fuel at extended burnup

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael M.; Silva, Antonio Teixeira, E-mail: rmdias@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    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)

  14. Operation and maintenance in Genkai PWR Plant

    International Nuclear Information System (INIS)

    Ohta, Shojiro

    1984-01-01

    The No.1 PWR plant with 559 MW capacity in the Genkai Nuclear Power Station, Kyushu Electric Power Co., Inc., required about 115 days for the regular inspection in fiscal 1982 and thereafter, although more maintenance work was done. But No.2 plant of the same type required not more than 80 days. In most cases, the period of one operation cycle was from 10 to 12 months, but in the third operation cycle of No.2 plant, it is expected to be 13 months. The capacity ratio of the whole power station was 75.2% at the end of fiscal 1983. These operational records all exceeded the Japanese average. The plants are two-loop Westinghouse type PWRs, and No.1 plant started the commercial operation of anti h and the increment of P 0 + . (author) apacity ratio of No.1 plant was 71.6%, and that of No.2 plant was 85.5%. The intergranular attack on steam generator tubes was found first in the fifth regular inspection, and also in the sixth and seventh inspections, and the faulty tubes were plugged. The prevention of its spread is the largest problem. The in-service quality assurance activity, the personnel training program and the effort of upgrading the plant availability are reported. (Kako, I.)

  15. Method of starting up PWR type reactor

    International Nuclear Information System (INIS)

    Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

    1988-01-01

    Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

  16. Computer aided information system for a PWR

    International Nuclear Information System (INIS)

    Vaidian, T.A.; Karmakar, G.; Rajagopal, R.; Shankar, V.; Patil, R.K.

    1994-01-01

    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

  17. Economic optimization of PWR cores with ROSA

    International Nuclear Information System (INIS)

    Verhagen, F.C.M.; Wakker, P.H.

    2005-01-01

    The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG's loading pattern optimization code system for PWRs, has proven for over a decade to be a valuable tool to reactor operators for improving their fuel management economics. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and rules. This paper outlines recent developments of the ROSA code system and discusses results of PWR specific applications of ROSA. Core designs with a large variety of challenging constraints have been realized with ROSA. As a typical example, for the 193 assembly, Vantage 5H/RFA-2 fueled TVA's Watts Bar unit 1, a cycle 4 core with 76 feed assemblies was designed. This was followed by a high-energy cycle 5 with only 77 feed assemblies and approximately 535 days of natural cycle length. Subsequently, an economical core using 72 bundles was designed for cycle 6. This resulted in considerable savings in the cost of feed assemblies for reloads. The typical accuracy of ROSA compared to results of license codes in within ±0.02 for normalized assembly powers, ±0.03 for maximum enthalpy rise hot channel factor (F ΔH ), and ±3 days for natural cycle length. (author)

  18. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael M.; Silva, Antonio Teixeira

    2015-01-01

    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)

  19. ABB PWR fuel design for high burnup

    International Nuclear Information System (INIS)

    Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.

    1998-01-01

    Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)

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

    International Nuclear Information System (INIS)

    Le Calvar, Marc; Bretelle, J. L.; Cailleaux, J. P.; Lacroix, R.; Guivarch, M.; Gay, N.; Taunier, S.; Gressier, F.; Varry, P.; Corredera, G.; Alos-Ramos, O.; Dijoux, M.

    2012-09-01

    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)

  1. 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.

  2. Benefits of Low Boron Core Design Concept for PWR

    International Nuclear Information System (INIS)

    Daing, Aung Tharn; Kim, Myung Hyun

    2009-01-01

    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 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bezdikian, G. [Electricite de France (EDF), Div. Production Nucleaire, 93 - Saint-Denis (France); Quinot, P. [FRAMATOME, Dept. Bloc Reacteur et Boucles Primaires, 92 - Paris-La-Defence (France); Faidy, C.; Churier-Bossennec, H. [Electricite de France (EDF), Div. Ingenierie et Service, 69 - Villeurbanne (France)

    2001-07-01

    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)

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

    International Nuclear Information System (INIS)

    Bezdikian, G.; Quinot, P.; Faidy, C.; Churier-Bossennec, H.

    2001-01-01

    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)

  5. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    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

  6. Development of Cost Estimation Methodology of Decommissioning for PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il; Yoo, Yeon Jae; Lim, Yong Kyu; Chang, Hyeon Sik; Song, Geun Ho

    2013-01-01

    The permanent closure of nuclear power plant should be conducted with the strict laws and the profound planning including the cost and schedule estimation because the plant is very contaminated with the radioactivity. In Korea, there are two types of the nuclear power plant. One is the pressurized light water reactor (PWR) and the other is the pressurized heavy water reactor (PHWR) called as CANDU reactor. Also, the 50% of the operating nuclear power plant in Korea is the PWRs which were originally designed by CE (Combustion Engineering). There have been experiences about the decommissioning of Westinghouse type PWR, but are few experiences on that of CE type PWR. Therefore, the purpose of this paper is to develop the cost estimation methodology and evaluate technical level of decommissioning for the application to CE type PWR based on the system engineering technology. The aim of present study is to develop the cost estimation methodology of decommissioning for application to PWR. Through the study, the following conclusions are obtained: · Based on the system engineering, the decommissioning work can be classified as Set, Subset, Task, Subtask and Work cost units. · The Set and Task structure are grouped as 29 Sets and 15 Task s, respectively. · The final result shows the cost and project schedule for the project control and risk management. · The present results are preliminary and should be refined and improved based on the modeling and cost data reflecting available technology and current costs like labor and waste data

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

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    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

  8. Lightweight submersed 'Walking' NDE manipulators for PWR and BWR vessel weld inspection

    International Nuclear Information System (INIS)

    Saernmark, Ivan; Lenz, Herbert

    2008-01-01

    Three new manipulators developed by WesDyne TRC in Sweden have under the year 2007 performed three very successful inspections in the PWR reactor Ringhals 3 and the BWR reactors Ringhals 1 and Oskarshamn 1. The manipulator systems can be used to perform inspection of circumferential and vertical welds on the reactor pressure vessel, the core shroud, core shroud support in BWR reactors or vessel and core barrel welds in PWR reactors. Most other flat or curved surfaces can be inspected using the new concept through relatively simple mechanical reconfigurations of system modules. The first inspection was performed on the R3 PWR core barrel in June 2007 with a very good result. This Manipulator is designed for access in very narrow gaps and for the type of core barrels with a shield covering the whole area of the perimeter. The manipulator is attached to the inspection area by means of a new unique suction cup system. The current manipulators consist of a curved horizontal beam, with radius similar to the reactor vessel, and a straight vertical beam, forming a T-shaped structure. By alternating the application of suction cup pairs on the horizontal beam and the vertical beam and by driving the scanning motors, the manipulator performs an incremental translational movement upwards/downwards or from side to side. The principles of this system give a well defined and stable platform for global and local positioning accuracy. A combination of advanced sensor solutions provides accurate position information in the absence of other physical reference objects. The system is controlled by the new WesDyne TRC Motor Control Panel and software, the MCP is specifically designed for remote control of submersed manipulators using techniques for cable reduction

  9. Lightweight submersed 'Walking' NDE manipulators for PWR and BWR vessel weld inspection

    Energy Technology Data Exchange (ETDEWEB)

    Saernmark, Ivan; Lenz, Herbert [WesDyne TRC AB, Stockholm (Sweden)

    2008-04-15

    Three new manipulators developed by WesDyne TRC in Sweden have under the year 2007 performed three very successful inspections in the PWR reactor Ringhals 3 and the BWR reactors Ringhals 1 and Oskarshamn 1. The manipulator systems can be used to perform inspection of circumferential and vertical welds on the reactor pressure vessel, the core shroud, core shroud support in BWR reactors or vessel and core barrel welds in PWR reactors. Most other flat or curved surfaces can be inspected using the new concept through relatively simple mechanical reconfigurations of system modules. The first inspection was performed on the R3 PWR core barrel in June 2007 with a very good result. This Manipulator is designed for access in very narrow gaps and for the type of core barrels with a shield covering the whole area of the perimeter. The manipulator is attached to the inspection area by means of a new unique suction cup system. The current manipulators consist of a curved horizontal beam, with radius similar to the reactor vessel, and a straight vertical beam, forming a T-shaped structure. By alternating the application of suction cup pairs on the horizontal beam and the vertical beam and by driving the scanning motors, the manipulator performs an incremental translational movement upwards/downwards or from side to side. The principles of this system give a well defined and stable platform for global and local positioning accuracy. A combination of advanced sensor solutions provides accurate position information in the absence of other physical reference objects. The system is controlled by the new WesDyne TRC Motor Control Panel and software, the MCP is specifically designed for remote control of submersed manipulators using techniques for cable reduction.

  10. Fatigue Life of Stainless Steel in PWR Environments with Strain Holding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taesoon; Kim, Kyuhyung [KHNP CRI, Daejeon (Korea, Republic of); Seo, Myeonggyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Many components and structures of nuclear power plants are exposed to the water chemistry conditions during the operation. Recently, as design life of nuclear power plant is expanded over 60 years, the environmentally assisted fatigue (EAF) due to these water chemistry conditions has been considered as one of the important damage mechanisms of the safety class 1 components. Therefore, many studies to evaluate the effect of light water reactor (LWR) coolant environments on fatigue life of materials have been conducted. Many EAF test results including Argonne National Laboratory’s consistently indicated the substantial reduction of fatigue life in the light water reactor environments. However, there is a discrepancy between laboratory test data and plant operating experience regarding the effects of environment on fatigue: while laboratory test data suggest huge accumulation of fatigue damage, very limited experience of cracking caused by the low cycle fatigue in light water reactor. These hold-time effect tests are preformed to characterize the effects of strain holding on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 stainless steel in 310℃ air and PWR environments with triangular strain. In agreement with the previous reports, the LCF life was reduced in PWR environments. Also for the slower strain rate, the reduction of LCF life was greater than the faster strain rate. The LCF test conditions for the hold-time effects were determined by the references and consideration of actual plant transient. To simulate the heat-up and cooldown transient, sub-peak strain holding during the down-hill of strain amplitude was chosen instead of peak strain holding which used in the previous researches.

  11. Development of system design in recent Siemens/KWU PWR influenced by PSA results

    International Nuclear Information System (INIS)

    Feigel, A.; Fabian, H.

    1989-01-01

    This paper reports on the design of the latest Siemens/KWU PWRs (Convoy plants) which is checked by a PSA, performed as a Δ-analysis to the German Risk Study (GPRA), which used a PWR 1300 MW in commercial operation since 1977 as a reference plant. The 10 years difference in the design between the reference plant and the Convoy plants, led to design changes, due to operational experience and findings from GPRA, Phase A. These are evaluated quantitatively by a PSA with respect to plat safety level and balance of the safety concept. The results gained from the Convoy PSA showed the importance and appropriateness of these modifications. Even if the latest results from GPRA, Phase B are considered with respect to additional accident sequences, it can be demonstrated that the new design is balanced with respect to these additional sequences. So no need exists to improve the new design any more

  12. Korean Reference HLW Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Lee, J. Y.; Kim, S. S. (and others)

    2008-03-15

    This report outlines the results related to the development of Korean Reference Disposal System for High-level radioactive wastes. The research has been supported around for 10 years through a long-term research plan by MOST. The reference disposal method was selected via the first stage of the research during which the technical guidelines for the geological disposal of HLW were determined too. At the second stage of the research, the conceptual design of the reference disposal system was made. For this purpose the characteristics of the reference spent fuels from PWR and CANDU reactors were specified, and the material and specifications of the canisters were determined in term of structural analysis and manufacturing capability in Korea. Also, the mechanical and chemical characteristics of the domestic Ca-bentonite were analyzed in order to supply the basic design parameters of the buffer. Based on these parameters the thermal and mechanical analysis of the near-field was carried out. Thermal-Hydraulic-Mechanical behavior of the disposal system was analyzed. The reference disposal system was proposed through the second year research. At the final third stage of the research, the Korean Reference disposal System including the engineered barrier, surface facilities, and underground facilities was proposed through the performance analysis of the disposal system.

  13. Activity incorporation into zinc doped PWR oxides

    International Nuclear Information System (INIS)

    Maekelae, Kari

    1998-01-01

    Activity incorporation into the oxide layers of PWR primary circuit constructional materials has been studied in Halden since 1993. The first zinc injection tests showed that zinc addition resulted in thinner oxide layers on new metal surfaces and reduced further incorporation of activity into already existing oxides. These tests were continued to find out the effects of previous zinc additions on the pickup of activity onto the surface oxides which were subsequently exposed to zinc-free coolant. The results showed that previous zinc addition will continue to reduce the rate of Co-60 build-up on out-of-core surfaces in subsequent exposure to zinc-free coolants. However, the previous Zn free test was performed for a relatively short period of time and the water chemistry programme was continued to find out the long term effects for extended periods without zinc. The activity incorporation into the stainless steel oxides started to increase as soon as zinc dosing to the coolant was stopped. The Co-60 concentration was lowest on all of the coupons which were first oxidised in Zn containing primary coolant. After the zinc injection period the thickness of the oxides increased, but activity in the oxide films did not increase at the same rate. This could indicate that zinc in the oxide blocks the adsorption sites for Co-60 incorporation. The Co-60 incorporation rate into the oxides on Inconel 600 seemed to be linear whether the oxide was pre-oxidised with or without Zn. The results indicate that zinc can either replace or prevent cobalt transport in the oxides. The results show that for zinc injection to be effective it should be carried out continuously. Furthermore the actual mechanism by which Zn inhibits the activity incorporation into the oxides is still not clear. Therefore, additional work has to follow with specified materials to verify the conclusions drawn in this work. (author)

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

    International Nuclear Information System (INIS)

    Gao Rui; Yang Yanhua; Lin Meng

    2007-01-01

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

  15. Basic information about development and construction of a PWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1977-01-01

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

  16. Improved emergency elevated air release for simplified PWR

    International Nuclear Information System (INIS)

    Naitoh, T.; Bruce, R.A.; Hirota, K.; Tajiri, Y.

    1992-01-01

    In developing the application of the simplified PWR in Japan, one of the most important areas is to limit post-accident site boundary whole body dose. In addressing this, the concept of Emergency Passive Air Filtration System (EPAFS) and it's feasibility is developed. The efficiency of charcoal filtering and the atmospheric diffusion effect of an elevated air release are important for dose reduction. The performance of these functions was evaluated by confirmatory testing. The test results confirmed a 99 percent efficiency of charcoal filter and an atmospheric diffusion effect higher than that of a conventional plant. The Emergency Passive Air Filtration System (EPAFS) and the atmospheric diffusion effect of elevated air release contribute to making the calculated post-accident site boundary whole body dose of simplified PWR as low as that of the conventional Japanese PWR plant. (author)

  17. Swing-Down of 21-PWR Waste Package

    International Nuclear Information System (INIS)

    A.K. Scheider

    2001-01-01

    The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended 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. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design

  18. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  19. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

    1980-01-01

    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

  20. The advanced main control console for next japanese PWR plants

    International Nuclear Information System (INIS)

    Tsuchiya, A.; Ito, K.; Yokoyama, M.

    2001-01-01

    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  1. Cylindrization of a PWR core for neutronic calculations

    International Nuclear Information System (INIS)

    Santos, Rubens Souza dos

    2005-01-01

    In this work we propose a core cylindrization, starting from a PWR core configuration, through the use of an algorithm that becomes the process automated in the program, independent of the discretization. This approach overcomes the problem stemmed from the use of the neutron transport theory on the core boundary, in addition with the singularities associated with the presence of corners on the outer fuel element core of, existents in the light water reactors (LWR). The algorithm was implemented in a computational program used to identification of the control rod drop accident in a typical PWR core. The results showed that the algorithm presented consistent results comparing with an production code, for a problem with uniform properties. In our conclusions, we suggest, for future works, for analyzing the effect on mesh sizes for the Cylindrical geometry, and to compare the transport theory calculations versus diffusion theory, for the boundary conditions with corners, for typical PWR cores. (author)

  2. PWR fuel performance and future trend in Japan

    International Nuclear Information System (INIS)

    Kondo, Y.

    1987-01-01

    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 R and D on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important R and D targets are the burnup extension, Gd contained fuel, Pu utilizatoin and the load follow capacility. (author)

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

    International Nuclear Information System (INIS)

    Fairclough, M.P.; Tymons, B.J.

    1985-06-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

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

    International Nuclear Information System (INIS)

    Bayley, B.; Stilwell, W.E.; Kent, N.A.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2012-01-01

    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)

  8. Steam Generator Owners Group PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Green, S.J.

    1985-01-01

    In 1981 the Steam Generator Owners Group (SGOG), a group of domestic and foreign pressurized water reactor (PWR) owners, developed and issued the PWR secondary water chemistry guidelines. The guidelines were prepared in response to the growing recognition that a majority of the problems causing reduced steam generator reliability (e.g., denting, wasteage, pitting, etc.) were related to secondary (steam) side water purity. The guidelines were subsequently issued as an Electric Power Research Institute (EPRI) report. In 1984 they were revised to reflect industry experience in adopting the original issuance and to incorporate new information on causes of corrosion damage. The guidelines have been endorsed and their adoption recommended by the SGOG

  9. Sensitivity of risk parameters to human errors for a PWR

    International Nuclear Information System (INIS)

    Samanta, P.; Hall, R.E.; Kerr, W.

    1980-01-01

    Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study

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

    International Nuclear Information System (INIS)

    Souza, A.L. de

    1989-01-01

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

  11. The latest full-scale PWR simulator in Japan

    International Nuclear Information System (INIS)

    Nishimuru, Y.; Tagi, H.; Nakabayashi, T.

    2004-01-01

    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)

  12. Advanced ion exchange resins for PWR condensate polishing

    International Nuclear Information System (INIS)

    Hoffman, B.; Tsuzuki, S.

    2002-01-01

    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)

  13. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

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

    1979-11-01

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

  14. Two optimal control methods for PWR core control

    International Nuclear Information System (INIS)

    Karppinen, J.; Blomsnes, B.; Versluis, R.M.

    1976-01-01

    The Multistage Mathematical Programming (MMP) and State Variable Feedback (SVF) methods for PWR core control are presented in this paper. The MMP method is primarily intended for optimization of the core behaviour with respect to xenon induced power distribution effects in load cycle operation. The SVF method is most suited for xenon oscillation damping in situations where the core load is unpredictable or expected to stay constant. Results from simulation studies in which the two methods have been applied for control of simple PWR core models are presented. (orig./RW) [de

  15. 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.

  16. Irradiation behavior of German PWR RPV steels under operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

    2011-07-01

    In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

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

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

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

  18. Transient performance of flow in PWR reactor circuits

    International Nuclear Information System (INIS)

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

    1988-12-01

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

  19. Qualification according to PDI's techniques UT EPRI methodology Phased Array for the inspection of vessels of PWR reactor with advanced robotic equipment; Cualificacion segun metodologia PDI de EPRI de te cnicas UT Phased Array para la inspeccion de vasijas de reactor PWR con eq uipos roboticos avanzados

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J. R.; Gonzalez, P.; Fernandez, F.

    2014-07-01

    The techniques and procedures qualified in the program EPRI PDI are directly applicable in plants whose reference code is ASME XI - specifically the Appendix VIII-, mainly USA and countries in which it is established American PWR technology. While countries with reactors in operation technology ABB (Sweden) or type VVER (Finland and Eastern countries) requires a qualification of specific technical type ENIQ, PDI qualification is a valuable reference since it allows to deal with such qualifications with guarantees. (Author)

  20. Study on dynamic characteristics of reduced analytical model for PWR reactor internal structures

    International Nuclear Information System (INIS)

    Yoo, Bong; Lee, Jae Han; Kim, Jong Bum; Koo, Kyeong Hoe

    1993-01-01

    The objective of this study is to establish the procedure of the reduced analytical modeling technique for the PWR reactor internal(RI) structures and to carry out the sensitivity study of the dynamic characteristics of the structures by varying the structural parameters such as the stiffness, the mass and the damping. Modeling techniques for the PWR reactor internal structures and computer programs used for the dynamic analysis of the reactor internal structures are briefly investigated. Among the many components of RI structures, the dynamic characteristics for CSB was performed. The sensitivity analysis of the dynamic characteristics for the reduced analytical model considering the variations of the stiffnesses for the lower and upper flanges of the CSB and for the RV Snubber were performed to improve the dynamic characteristics of the RI structures against the external loadings given. In order to enhance the structural design margin of the RI components, the nonlinear time history analyses were attempted for the RI reduced models to compare the structural responses between the reference model and the modified one. (Author)

  1. Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)

    International Nuclear Information System (INIS)

    Peng Hong Liem; Surian Pinem; Tagor Malem Sembiring; Tran Hoai Nam

    2015-01-01

    A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA). The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducting to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed. (author)

  2. A survey of available margin in a PWR RIA with statistical methods and 3D kinetics

    International Nuclear Information System (INIS)

    Riverola Gurruchaga, J.; Nunez Rodriguez, T.

    2010-01-01

    This paper investigates the recovery of margin in a PWR RIA simulation with 3D kinetics, due to statistical techniques. The chosen reference core is a typical 12 feet, 17*17 PWR, with very low leakage loading pattern strategy and gadolinium oxide as burnable poison. The PARCS calculated average nuclear power and nodal power are transferred to a hot spot model for a sequential calculation of fuel temperature and enthalpy responses allowing for independent hypothesis in both calculations. The hot spot analysis is done with a pellet type model with RELAP. The analysis is done at HZP and EOC, since this state is the most limiting one respect to the enthalpy rise criterion, compared to other burn-up condition or initial power cases. In this work, the enthalpy increase is estimated with several statistical methods of propagation of uncertainties: order statistics, parametric statistics, surface response and sensitivities. A discussion on the advantages and disadvantages of each method is also presented. This statistical analysis is also useful to confirm a previous classification of parameters and assumptions according to their importance for the simulation, and found to be consistent with the state of the art in the published literature. These parameters include ejected rod worth and ejection time, delayed neutron fraction and yields, nuclear power peaking factor, and Doppler. (authors)

  3. Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

    International Nuclear Information System (INIS)

    Polidoro, Franco; Corsetti, Edoardo; Vimercati, Giuliano

    2011-01-01

    The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO 2 - PuO 2 ) fuel assemblies up to 50% of the core, together with UO 2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO 2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO 2 . (author)

  4. Analysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems

    International Nuclear Information System (INIS)

    Baiocco, G.; Petruzzi, A.; Bznuni, S.; Kozlowski, T.

    2017-01-01

    Highlights: • The consistency between Helios and Serpent few-group cross sections is shown. • The PARCS model is validated against a Monte Carlo 3D model. • The fission and capture rates are compared. • The influence of the spacer grids on the axial power distribution is shown. - Abstract: Lattice physics codes are primarily used to generate cross-section data for nodal codes. In this work the methodology of homogenized constant generation was applied to a small Pressurized Water Reactor (PWR) core, using the deterministic code Helios and the Monte Carlo code Serpent. Subsequently, a 3D analysis of the PWR core was performed with the nodal diffusion code PARCS using the two-group cross section data sets generated by Helios and Serpent. Moreover, a full 3D model of the PWR core was developed using Serpent in order to obtain a reference solution. Several parameters, such as k eff , axial and radial power, fission and capture rates were compared and found to be in good agreement.

  5. The pseudo-harmonics method: an application involving perturbations caused by control rod insertion in PWR reactors

    International Nuclear Information System (INIS)

    Claro, L.H.; Alvim, A.C.M.; Thome, Z.D.

    1988-08-01

    The objective of this work is to stydy the effect of intense perturbations, such as control rod insertion in the core of PWR reactors, through a perturbation approach consisting of a modified version of the pseudo-harmonics method. A typical one-dimensional PWR reactor model was used as a reference state, from which two perturbations were imposed, simulation gray and black control rod insertion. In the first case, eigenvalue convergence was achieved with the eighth order of approximation approximation and perturbed fluxes and eigenvalue estimates agreed very well with direct calculation results. The second case tested represents a very intense localized perturbation. Oscillation in keff were observed er of approximation increased and the method failed to converge. Results obtained indicate that the pseudo-harmonics method can be used to compute 2 group fluxes and fundamental eigenvalue of perturbated states resulting from gray control rod insertion in PWR reactors. The method is limited, however, by perturbation intensity, as other perturbation methods are. (author) [pt

  6. Growth references

    NARCIS (Netherlands)

    Buuren, S. van

    2007-01-01

    A growth reference describes the variation of an anthropometric measurement within a group of individuals. A reference is a tool for grouping and analyzing data and provides a common basis for comparing populations.1 A well known type of reference is the age-conditional growth diagram. The

  7. 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

  8. 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

  9. An integrated PWR for marine propulsion

    International Nuclear Information System (INIS)

    Letouze, A.; Marecaux, A.; Rollason, J.; Heap, S.; Foster, A.; Jewer, S.; Thompson, A. C.; Williams, A. M.; Beeley, P. A.

    2008-01-01

    Results from a design study for a nuclear propulsion plant utilising a small integrated PWR using many of the inherent safety features of the IRIS design. The design consists of a single pass, low enrichment core housed, together with all associated primary circuit components, within a reactor pressure vessel 10.3 m high and 4.1 m in diameter. Reactor physics calculations were conducted with the codes WIMS9a and MONK8b. The core design contains 21 fuel assemblies each containing 264 UO 2 fuel pins. Each fuel module has a cluster of 24 boron carbide control rods and a central instrumentation channel. The fuel enrichment was 9% in order to achieve the core lifetime requirement of 3000 EFPD at a reactor power of 120 MWth. This gives a discharge burnup of 51,000 MWd/t. To control excess reactivity, two forms of burnable poison are employed: a zirconium dibromide (ZrB 2 ) coating on the fuel compacts, and gadolinium oxide homogeneously mixed in the fuel. Thermal hydraulic calculations were performed using TRAC-P(ND) for steady-state operation and for a number of fault transients. The helical once through steam generators were modelled using heat structure and pipe components and their performance compared to independent calculations including heat transfer correlations for the helical coiled geometry. Intact circuit calculations for steady state were followed by a small break LOCA calculation including the effect of a containment volume which reproduced the gain of coolant effect reported for IRIS. It was demonstrated that the thermal limits were not exceeded for the identified key transients. The dynamic response of the reactor plant to typical power demands was modelled using AcslXtreme software. Several schemes for limiting the power overshoot that was found on rapid increase to full power were examined. It was concluded that the SG must be operated with variable secondary pressure and the best means of reducing power overshoot is to step back the throttle opening

  10. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Doyen, J.J.; Lebey, J.

    1979-01-01

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 280 0 C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 280 0 C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  11. Methodology for the LABIHS PWR simulator modernization

    International Nuclear Information System (INIS)

    Jaime, Guilherme D.G.; Oliveira, Mauro V.

    2011-01-01

    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

  12. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J.; Hoogenboom, J.E.; Leege, P.F.A. de; Voet, J. van der; Verhagen, F.C.M.

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs

  13. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

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

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  16. Method of characteristics - Based sensitivity calculations for international PWR benchmark

    International Nuclear Information System (INIS)

    Suslov, I. R.; Tormyshev, I. V.; Komlev, O. G.

    2013-01-01

    Method to calculate sensitivity of fractional-linear neutron flux functionals to transport equation coefficients is proposed. Implementation of the method on the basis of MOC code MCCG3D is developed. Sensitivity calculations for fission intensity for international PWR benchmark are performed. (authors)

  17. Studies of a small PWR for onsite industrial power

    International Nuclear Information System (INIS)

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

    1977-01-01

    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

  18. Make use of EDF orientations in PWR fuel management

    International Nuclear Information System (INIS)

    Gloaguen, A.

    1989-01-01

    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 [fr

  19. Parameterized representation of macroscopic cross section for PWR reactor

    International Nuclear Information System (INIS)

    Fiel, João Cláudio Batista; Carvalho da Silva, Fernando; Senra Martinez, Aquilino; Leal, Luiz C.

    2015-01-01

    Highlights: • This work describes a parameterized representation of the homogenized macroscopic cross section for PWR reactor. • Parameterization enables a quick determination of problem-dependent cross-sections to be used in few group calculations. • This work allows generating group cross-section data to perform PWR core calculations without computer code calculations. - Abstract: The purpose of this work is to describe, by means of Chebyshev polynomials, 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 235 92 U enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement

  20. Dissolution process for advanced-PWR-type fuels

    International Nuclear Information System (INIS)

    Black, D.E.; Decker, L.A.; Pearson, L.G.

    1979-01-01

    The new Fluorinel Dissolution Process and Fuel Storage (FAST) Facility at ICPP will provide underwater storage of spent PWR fuel and a new head-end process for fuel dissolution. The dissolution will be two-stage, using HF and HNO 3 , with an intermittent H 2 SO 4 dissolution for removing stainless steel components. Equipment operation is described

  1. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

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

  2. Is it possible to improve regulation system of PWR

    International Nuclear Information System (INIS)

    Bonnemay, A.; Martinez, J.M.

    1983-03-01

    This paper deals with two problems: first of all, it presents the critical analysis of usually implemented general regulation systems, on PWR plants, and derives from it same possibilities to improve the transient behavior of reactor, the second part is a proposition from an automatic control system for spatial distribution of flux

  3. Coolant flow monitoring in a PWR core using noise analysis

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1992-01-01

    Experimental investigations of the neutron and temperature noise field have been performed in the 1350 MW PWR nuclear power plant. Evaluation in the low frequency range, where both feedback effects and different thermohydraulics phenomena are dominant, succeeded in measuring the coolant velocity. This is important for determination and localization of essential deviations and possible anomalies. (author)

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

    International Nuclear Information System (INIS)

    Delourme, Didier.

    1980-11-01

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

  5. Performance of PWR Nuclear power plants, up to 1985

    International Nuclear Information System (INIS)

    Muniz, A.A.

    1987-01-01

    The performance of PWR nuclear power plants is studied, based on operational data up to 1985. The availability analysis was made with 793 unit-year and the reliability analysis was made with 5851 unit x month. The results were discussed and the availability of those nuclear power plants were estimated. (E.G.) [pt

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

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1976-01-01

    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) [de

  7. The new French code for PWR in service inspection

    Energy Technology Data Exchange (ETDEWEB)

    Noel, R; Hutin, J P [Electricite de France (EDF), 75 - Paris (France)

    1988-12-31

    This document presents the new french code for pressured water reactor in service inspection. The historic regulatory basis is presented, together with the new regulatory act (dating back to the 26 february 1974) and the major guidelines of the french practice for in service inspection of PWR components. (TEC).

  8. Influence of boron reduction strategies on PWR accident management flexibility

    International Nuclear Information System (INIS)

    Papukchiev, Angel Aleksandrov; Liu, Yubo; Schaefer, Anselm

    2007-01-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. Design changes to reduce boron concentration in the reactor coolant are of general interest regarding three aspects - improved reactivity feedback properties, lower impact of boron dilution scenarios on PWR safety and eventually more flexible accident management procedures. In order to assess the potential advantages through the introduction of boron reduction 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 ppm and 805 ppm. For the assessment of the potential safety advantages of these cores a hypothetical beyond design basis accident has been simulated with the system code ATHLET. The analyses showed improved inherent safety and increased accident management flexibility of the low boron cores in comparison with the standard PWR. (author)

  9. Directives and general design requirements for a small PWR

    International Nuclear Information System (INIS)

    Arrieta, L.A.

    1992-08-01

    A proposal of directives and general requirements for the development of a small PWR conceptual design is presented. These directives address the main safety, performance and economic design aspects. The purpose is to use this work as a base for a wide discussion, involving all project participants, culminating with the definition of the final directives and general requirements. (author)

  10. Post irradiation examination on test fuel pins for PWR

    International Nuclear Information System (INIS)

    Fogaca Filho, N.; Ambrozio Filho, F.

    1981-01-01

    Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

  11. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    O'Leary, P.M.; Pitts, M.L.

    2000-01-01

    Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers

  12. Zinc injection in German PWR plants

    International Nuclear Information System (INIS)

    Streit, K.

    2004-01-01

    Operating experience acquired at PWR NNPs shows that zinc injection at low concentrations of 5 ppb is a very effective source term reduction measure. This method does not lead to any operating restrictions or other negative effects on plant systems and components. The nuclear industry has been very successful in reducing radiation exposures within the past two decades. Annual exposures could be significantly decreased and are now at a level of around 1 man-Sv per plant and year. This great success can mainly be attributed to the general commitment of plant operators to maintaining radiation exposures of workers in the controlled access area as low as reasonably achievable (ALARA principle). The ALARA principle, of course, also implies evaluation of the economic benefit of radiation protection measures. Radiation source term reduction has drawn increasing attention of plant operators in recent years. For the new PWRs cobalt-based alloys in the primary system have successively been eliminated already at the design and construction phase within the last decade. Use of wear-resistant cobalt-free substitute materials in combination with the general use of advanced alloys for the steam generator tubing of PWRs resulted in low values for the two most common sources of plant radiation fields, namely 58 Co and 60 Co. Investigations showed that the beneficial effect of zinc can be related to its high affinity for mixed spinel oxide phases, resulting in the following two basic effects: -Zinc is incorporated preferentially into the oxide layer on primary system surfaces and thus reduces pickup of 58 Co and 60 Co and - Zinc can displace cobalt isotopes from existing oxide layers. In German PWRs with Incoloy 800 steam generator tubing material (Ni-content -32%), the observed reductions correspond to a decrease in dose rates of around 10 to 15% per year and thus follow, as predicted, the half-life time of 60 Co. Overall reductions in high radiation areas are now in the range of

  13. [Reference citation].

    Science.gov (United States)

    Brkić, Silvija

    2013-01-01

    Scientific and professional papers represent the information basis for scientific research and professional work. References important for the paper should be cited within the text, and listed at the end of the paper. This paper deals with different styles of reference citation. Special emphasis was placed on the Vancouver Style for reference citation in biomedical journals established by the International Committee of Medical Journal Editors. It includes original samples for citing various types of articles, both printed and electronic, as well as recommendations related to reference citation in accordance with the methodology and ethics of scientific research and guidelines for preparing manuscripts for publication.

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

    International Nuclear Information System (INIS)

    Hermann, O.W.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  16. Construction of PWR nuclear cross sections for transient calculations. Test of the ANTI program against TWODIM

    International Nuclear Information System (INIS)

    Thorlaksen, B.

    1981-05-01

    Nuclear cross sections for fuel assemblies of the more recent Westinghouse designs, representing two different PWR reactor cores, are calculated as functions of average fuel temperature, moderator density, and moderator poison concentration. The cross-section functions are verified by referring to Westinghouse power-shape calculations and other analysis. Computations on the side reflector resulted in significantly higher albedo values than used previously for BWR's in similar nodal codes. This led to an investigation of the influence of the internodal coupling coefficients on the power shape. It is concluded that the calculated power shape is strongly dependent, on the choise of coupling coefficients. However, it is shown that ''the correct'' set of coupling coefficients depends mostly on the nodal configuration, and that it is fairly independent of the power condition. (author)

  17. Reentrainment of droplet from grid spacer in mist flow portion of LOCA reflood of PWR

    International Nuclear Information System (INIS)

    Lee, S.L.; Cho, S.K.; Sheen, H.J.

    1983-01-01

    An investigation is made on the influence of a quenched grid spacer on the greatly enhanced heat transfer from heated fuel rods during the mist flow phase of emergency reflood of loss of coolant accident (LOCA) of a pressurized water reactor (PWR). The situation for the case of a dry grid spacer before its quenching has not been covered in this study. The experimental technique used is a relatively simple optical scheme which combines the reference-mode laser-Doppler anemometry making use of the scattering of a light beam from a droplet. The results reveal that the large droplets in the mist flow, which are intercepted by the grid spacer, are responsible for the creation of a large number of smaller droplets. These small droplets, due to their large surface area to mass ratios, can serve as superb evaporative cooling agents to heat transfer downstream of the grid spacer

  18. Full vessel CFD analysis on thermal-hydraulic characteristics of CPR1000 PWR

    International Nuclear Information System (INIS)

    Chao Yanmeng; Yang Lixin; Zhang Mingqian

    2014-01-01

    To obtain flow distributions and thermal-hydraulic properties in a full vessel PWR under limited computation ability and time, a full vessel simulation model of CPR1000 was built based on two simplification methods. One simplified the inner geometry of the control rod guide tubes using equivalent flow area. Another substituted the core by a porous domain to maintain the pressure drop and temperature rise. After the computation, global and localized flow distributions, hydraulic loads of some main assemblies were obtained, as well as other thermal-hydraulic properties. The results indicate the flow distribution in the full vessel is asymmetrical. Therefore it is essential to use the full vessel model to simulate. The calculated thermal-hydraulic characteristics agree well with the operation statistics, providing the reference data for the reactor safety operation. (authors)

  19. Assessment of non-backfittable concepts to improve PWR uranium utilization

    International Nuclear Information System (INIS)

    LaBelle, D.W.; Sankovich, M.F.; Spetz, S.W.; Uotinen, V.O.

    1980-12-01

    Seven non-backfittable improvements to light water reactors were assessed for Batelle/Pacific Northwest Laboratories in support of the Department of Energy's program on Advanced Reactor Studies. The objective was to provide industrial perspective as to which concepts have the best potential for development to improve fuel utilization. The concepts were rated against the assessment criteria while considering the key questions identified for each concept, and recommendations were made for further action on unresolved key questions. The concepts were subjectively ranked against each other in terms of relative investment potential. The ranking considered all criteria but, for example, weighted fuel utilization savings more heavily than development costs. Finally, conclusions and recommendations for future action were determined. The reference design for this study was the NASAP Composite Improved PWR

  20. Analysis of the NEACRP PWR rod ejection benchmark problems with DIF3D-K

    International Nuclear Information System (INIS)

    Kim, M.H.

    1994-01-01

    Analyses of the NEACRP PWR rod ejection transient benchmark problems with the DIF3D-K nodal kinetics code are presented. The DIF3D-K results are shown to be in generally good agreement with results obtained using other codes, in particular reference results previously generated with the PANTHER code. The sensitivity of the transient results to the DIF3D-K input parameters (such as time step size, radial and axial node sizes, and the mesh structure employed for fuel pin heat conduction calculation) are evaluated and discussed. In addition, the potential in reducing computational effort by application of the improved quasistatic scheme (IQS) to these rod ejection transients, which involve very significant flux shape changes and thermal-hydraulic feedback is evaluated

  1. Reference Assessment

    Science.gov (United States)

    Bivens-Tatum, Wayne

    2006-01-01

    This article presents interesting articles that explore several different areas of reference assessment, including practical case studies and theoretical articles that address a range of issues such as librarian behavior, patron satisfaction, virtual reference, or evaluation design. They include: (1) "Evaluating the Quality of a Chat Service"…

  2. VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2015-10-01

    Full Text Available ABSTRACT VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS. The coupled neutronic and thermal-hydraulic (T/H code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR ejection at peripheral core using a full core geometry model, the C1 and C2 cases.  By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM and the improved quasistatic method (IQS. All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16% occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4% for C2 case.  All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. Keywords: nodal method, coupled neutronic and thermal-hydraulic code, PWR, transient case, control rod ejection.   ABSTRAK VALIDASI MODEL GEOMETRI TERAS PENUH PAKET PROGRAM NODAL3 DALAM PROBLEM BENCHMARK GAYUT WAKTU PWR. Paket program kopel neutronik dan termohidraulika (T/H, NODAL3, telah divalidasi dengan beberapa kasus benchmark statis PWR dan kasus benchmark gayut waktu PWR NEACRP.  Akan tetapi, paket program NODAL3 belum divalidasi dalam kasus benchmark gayut waktu akibat penarikan sebuah perangkat batang kendali (CR di tepi teras menggunakan model geometri teras penuh, yaitu kasus C1 dan C2. Dengan penelitian ini, akurasi paket program

  3. Analysis of size effect applicable to evaluation of fracture toughness of base metal for PWR vessel

    International Nuclear Information System (INIS)

    Benhamou, C.; Joly, P.; Andrieu, A.; Parrot, A.; Vidard, S.

    2015-01-01

    The objective of the present paper is to review the specimen size effect (also called crack front length effect) on Fracture Toughness of PWR Reactor Pressure Vessel Steel base metal. The analysis of the reality and amplitude of this effect is conducted in a first step on a database (the so-called GKSS database) including fracture toughness test results on a single representative material using specimens of different thicknesses, tested in the same temperature range. A realistic analytical form for describing the size effect observed in this data set is thus derived from statistical analyses and proposed for engineering application. In a second step, this size effect formulation is then applied to a large number of fracture toughness data, obtained in Irradiation Surveillance Programs, and also to the numerous data used for the definition of the ASME (and RCC-M) fracture toughness reference curves. This analysis allows normalizing all the available fracture toughness data with a single specimen width of 100 mm and defining the fracture toughness reference curve as the lower bound of this normalized set of data points. It is thus demonstrated that the fracture toughness reference curve is associated with a reference crack length of 100 mm, and can be used in RPV integrity analyses for other crack front length in association with the crack front length correction formula defined in the first step. (authors)

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

    International Nuclear Information System (INIS)

    Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

    1988-01-01

    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

  5. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

  6. Recent references

    International Nuclear Information System (INIS)

    Ramavataram, S.

    1991-01-01

    In support of a continuing program of systematic evaluation of nuclear structure data, the National Nuclear Data Center maintains a complete computer file of references to the nuclear physics literature. Each reference is tagged by a keyword string, which indicates the kinds of data contained in the article. This master file of Nuclear Structure References (NSR) contains complete keyword indexes to literature published since 1969, with partial indexing of older references. Any reader who finds errors in the keyword descriptions is urged to report them to the National Nuclear Data Center so that the master NSR file can be corrected. In 1966, the first collection of Recent References was published as a separate issue of Nuclear Data Sheets. Every four months since 1970, a similar indexed bibliography to new nuclear experiments has been prepared from additions to the NSR file and published. Beginning in 1978, Recent References was cumulated annually, with the third issue completely superseding the two issues previously published during a given year. Due to publication policy changes, cumulation of Recent Reference was discontinued in 1986. The volume and issue number of all the cumulative issues published to date are given. NNDC will continue to respond to individual requests for special bibliographies on nuclear physics topics, in addition to those easily obtained from Recent References. If the required information is available from the keyword string, a reference list can be prepared automatically from the computer files. This service can be provided on request, in exchange for the timely communication of new nuclear physics results (e.g., preprints). A current copy of the NSR file may also be obtained in a standard format on magnetic tape from NNDC. Requests for special searches of the NSR file may also be directed to the National Nuclear Data Center

  7. MVP utilization for PWR design code

    International Nuclear Information System (INIS)

    Matsumoto, Hideki; Tahara, Yoshihisa

    2001-01-01

    MHI studies the method of the spatially dependent resonance cross sections so as to predict the power distribution in a fuel pellet accurately. For this purpose, the multiband method and the Stoker/Weiss method were implemented to the 2 dimensional transport code PHOENIX-P, and the methods were validated by comparing them with MVP code. Although the appropriate reference was not obtain from the deterministic codes on the resonance cross section study, now the Monte Carlo code MVP result is available and useful as reference. It is shown here how MVP is used to develop the multiband method and the Stoker/Weiss method, and how effective the result of MVP is on the study of the resonance cross sections. (author)

  8. French approach on the definition of reference defects to be considered for fracture mechanics analyses at design state

    Energy Technology Data Exchange (ETDEWEB)

    Grandemange, J M; Pellissier-Tanon, A [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-La-Defense (France)

    1988-12-31

    This document describes the french approach for verifying fracture resistance of PWR primary components. Three reference defects have been defined, namely the envelope defect, the exceptional defect and the conventional defect. It appears that a precise estimation of the available margins may be obtained by analyzing a set of reference defects representative of the flaws likely to exist in the components. (TEC). 5 refs.

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

    International Nuclear Information System (INIS)

    Fiel, Joao Claudio B.

    2015-01-01

    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 235 U 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 inf , generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)

  10. Seismic proving test of PWR reactor containment vessel

    International Nuclear Information System (INIS)

    Akiyama, H.; Yoshikawa, T.; Tokumaru, Y.

    1987-01-01

    The seismic reliability proving tests of nuclear power plant facilities are carried out by Nuclear Power Engineering Test Center (NUPEC), using the large-scale, high-performance vibration of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry (MITI). In 1982, the seismic reliability proving test of PWR containment vessel started using the test component of reduced scale 1/3.7 and the test component proved to have structural soundness against earthquakes. Subsequently, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. Whereupon, the seismic analysis and evaluation on the actual containment vessel were performed by these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed

  11. PWR plant transient analyses using TRAC-PF1

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  12. Material property changes of stainless steels under PWR irradiation

    International Nuclear Information System (INIS)

    Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

    2009-01-01

    Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

  13. Application of digital control in Japanese PWR Plants

    International Nuclear Information System (INIS)

    Taguchi, S.; Kondo, Y.; Teranishi, S.; Matsumiya, M.; Takashima, M.; Nagai, T.

    1986-01-01

    More reliable and flexible control system to improve the plant availability and operability is constantly demanded. In order to answer the demands, digital control systems are being applied to Japanese PWR plants. Microprocessor-based digital control systems are widely used in other industries and show good performance. The digital control system has been already applied to the chemical and volume control system and the radioactive waste disposal system in the operating plants. These systems have been working as expected and demonstrating good performances. The digital control system for the reactor control system, which is the main control system of the PWR plants, is being developed. The design of the system has been already finished and the verification/validation process is now in progress

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

    International Nuclear Information System (INIS)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de

    2015-01-01

    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)

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

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

    1988-05-01

    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

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

    International Nuclear Information System (INIS)

    Matzie, R.A.; Daleas, R.S.; Miller, D.D.

    1980-11-01

    As part of the US Department of Energy's Advanced Reactor Design Study, various nonbackfittable PWR design modifications were evaluated to determine their potential for improved uranium utilization and commercial viability. Combustion Engineering, Inc. contributed to this effort through participation in the Battelle Pacific Northwest Laboratory industrial assessment of such design modifications. Seven modifications, including the use of higher primary system temperatures and pressures, rapid-frequent refueling, end-of-cycle stretchout, core periphery modifications, radial blankets, low power density cores, and small PWR assemblies, were evaluated with respect to uranium utilization, economics, technical and operational complexity, and several other subjective considerations. Rapid-frequent refueling was judged to have the highest potential although it would probably not be economical for the majority of reactors with the design assumptions used in this assessment

  17. Electrical and control aspects of the Sizewell B PWR

    International Nuclear Information System (INIS)

    1992-01-01

    The pressurized water reactor, Sizewell-B, which is being built in Suffolk is well on in its construction schedule. This conference looked at the electrical and control aspects of the first PWR to be built in the United Kingdom. Although based on the standard Westinghouse PWR design, modifications have been made to meet the particular requirements of the site and the UK licensing regulations. There are 11 papers on all aspects of the electrical systems, 5 papers on the cables and cable installation, 5 on the main control rooms and auxiliary shutdown room, 5 on the integrated system and centralised operation, 6 on the monitoring and protection systems and 9 on the reactor protection systems. All 41 are indexed separately. (UK)

  18. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    2008-12-01

    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. PWR accident management realated tests: some Bethsy results

    International Nuclear Information System (INIS)

    Clement, P.; Chataing, T.; Deruaz, R.

    1993-01-01

    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

  20. A nodal model for the simulation of a PWR core

    International Nuclear Information System (INIS)

    Souza Pinto, R. de.

    1981-06-01

    A computer program FORTRAN language was developed to simulate the neutronic and thermal-hydraulic transient behaviour of a PWR reactor core. The reator power is calculated using a point kinectics model with six groups of delayed neutron precursors. The fission product decay heat was considered assuming three effective decay heat groups. A nodal model was employed for the treatment of heat transfer in the fuel rod, with integration of the heat equation by the lumped parameter technique. Axial conduction was neglected. A single-channel nodal model was developed for the thermo-hydrodynamic simulation using mass and energy conservation equations for the control volumes. The effect of the axial pressure variation was neglected. The computer program was tested, with good results, through the simulation of the transient behaviour of postulated accidents in a typical PWR. (Author) [pt

  1. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

    1993-03-01

    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

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

    International Nuclear Information System (INIS)

    Zhou Xiaojia; Mao Fei; Min Peng; Lin Shaoxuan

    2013-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Malone, D.J. [Consumers Power Co., Covert, MI (United States)

    1995-03-01

    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.

  4. Assessment of PWR plutonium burners for nuclear energy centers

    International Nuclear Information System (INIS)

    Frankel, A.J.; Shapiro, N.L.

    1976-06-01

    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible

  5. 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)

  6. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  7. Surveillance systems (PWR) - loose parts monitoring - vibration monitoring - leakage detection

    International Nuclear Information System (INIS)

    Schuette, A.; Blaesig, H.

    1982-01-01

    The contribution is engaged in the task and the results of the loose parts monitoring and the vibration monitoring following from the practice at the PWR of Biblis. First a description of both systems - location and type of the sensors used, the treatment of the measurements and the indications - is given. The results of the analysis of some events picked up by the surveillance systems are presented showing applicabilty and benefit of such systems. (orig.)

  8. A model to calculate the burn of gadolinium in PWR

    International Nuclear Information System (INIS)

    Sannazzaro, L.R.

    1983-01-01

    A cell model to calculate the burnup of a PWR fuel element with gadolinium as a poison, projected by KWU, is presented. With the model proposed, the burn of the gadolinium isotopes is analyzed, as well as the effect of these isotopes in the fuel element behaviour. The results obtained with this cell model are compared with those obtained by a conventional cell model. (E.G.) [pt

  9. Conversion rate for PWR reactors in thorium cycle

    International Nuclear Information System (INIS)

    Angelkorte, G.M.

    1980-01-01

    This work concerns to the determination of the conversion-rate for a PWR reactor with an enrichment of 7.47%, considering a cell, geometrically equal to Angra I, composed by Thorium and U-238 in a 1:1 relation. The study was performed considering neutrons of one and two groups of energy, according to the suggestion from other authors sup(1,2). It was also performed a study about the production and consumption of fissile material. (author)

  10. Measured performance of four PWR liquid radioactive waste treatment systems

    International Nuclear Information System (INIS)

    McIsaac, C.V.; Mandler, J.W.; Stalker, A.C.

    1980-01-01

    This paper presents results of a study of the liquid radwaste treatment and boron recovery systems of four operating PWR power plants. The performance of a given system was determined from measurements of radionuclide inventories in samples drawn from demineralizers, evaporators, filters, and gaseous cleanup systems. The plants at which measurements were made are Fort Calhoun, Zion 1 and 2, Turkey Point 3 and 4, and Rancho Seco

  11. Uranium savings on a once through PWR fuel cycle

    International Nuclear Information System (INIS)

    Cupo, J.V.

    1980-01-01

    A number of alternatives which have the greatest potential for near term savings with minimum plant and fuel modifications have been examined at Westinghouse as part of continued internal assessment and part of NASAP study conducted for DOE pertaining to uranium utilization in a once through PWR fuel cycle. The alternatives which could be retrofitted to existing reactors were examined in more detail in the evaluation since they would have the greater near term impact on U savings

  12. Vibration behavior of PWR reactor internals Model experiments and analysis

    International Nuclear Information System (INIS)

    Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

    1975-01-01

    In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

  13. Stress analysis on a PWR pressure vessel support structure

    International Nuclear Information System (INIS)

    Cruz, J.R.B.; Mattar Neto, M.; Jesus Miranda, C.A. de.

    1992-01-01

    The paper presents the stress analysis of a research PWR vessel support structure. Different geometries and thermal boundary conditions are evaluated. The finite element analysis is performed using ANSYS program. The ASME Section III criteria are applied for the stress verification and the following points are discussed: stress classification and linearization; jurisdictional boundary between ASME Subsection NB (Class 1 Components) and Subsection NF (Component Supports). (author)

  14. ORNL: PWR-BDHT analysis procedure, a preliminary overview

    International Nuclear Information System (INIS)

    Cliff, S.B.

    1978-01-01

    The computer programs currently used in the analysis of the ORNL-PWR Blowdown Heat Transfer Separate-Effects Program are overviewed. The current linkages and relationships among the programs are given along with general comments about the future directions of some of these programs. The overview is strictly from the computer science point of view with only minimal information concerning the engineering aspects of the analysis procedure

  15. Improvement on main control room for Japanese PWR plants

    International Nuclear Information System (INIS)

    Matsumiya, Masayuki

    1996-01-01

    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

  16. A comparative study of fuel management in PWR reactors

    International Nuclear Information System (INIS)

    Barroso, D.E.G.; Nair, R.P.K.; Vellozo, S.O.

    1981-01-01

    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 U 3 O 8 demand and SWU and the approximate costs of the principal stages of the fuel cycle, comparing with the stardand cycle without recycling. (E.G.) [pt

  17. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Andrade, G.G. de

    1982-01-01

    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.) [pt

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

    International Nuclear Information System (INIS)

    Aragones, J.M.

    1977-01-01

    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) [es

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

    International Nuclear Information System (INIS)

    Schoemacker, M.

    1984-01-01

    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 [fr

  20. Natural circulation in a scaled PWR integral test facility

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  1. 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.

  2. ORNL-PWR BDHT analysis procedure: an overview

    International Nuclear Information System (INIS)

    Cliff, S.B.

    1978-01-01

    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

  3. PWR surveillance based on correspondence between empirical models and physical

    International Nuclear Information System (INIS)

    Zwingelstein, G.; Upadhyaya, B.R.; Kerlin, T.W.

    1976-01-01

    An on line surveillance method based on the correspondence between empirical models and physicals models is proposed for pressurized water reactors. Two types of empirical models are considered as well as the mathematical models defining the correspondence between the physical and empirical parameters. The efficiency of this method is illustrated for the surveillance of the Doppler coefficient for Oconee I (an 886 MWe PWR) [fr

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

    International Nuclear Information System (INIS)

    Kodaira, Hideki

    1990-01-01

    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. Core power capability verification for PWR NPP

    International Nuclear Information System (INIS)

    Xian Chunyu; Liu Changwen; Zhang Hong; Liang Wei

    2002-01-01

    The Principle and methodology of pressurized water reactor nuclear power plant core power capability verification for reload are introduced. The radial and axial power distributions of normal operation (category I or condition I) and abnormal operation (category II or condition II) are simulated by using neutronics calculation code. The linear power density margin and DNBR margin for both categories, which reflect core safety, are analyzed from the point view of reactor physics and T/H, and thus category I operating domain and category II protection set point are verified. Besides, the verification results of reference NPP are also given

  6. Acceptance test for 900 MWe PWR unit replacement steam generators

    International Nuclear Information System (INIS)

    Gourguechon, B.

    1993-01-01

    During the first half of 1994, the Gravelines 1 steam generators will be replaced (SG replacement procedure). The new SG's differ from the former components notably by the alloy used for the tube bundle, in this case, the high chromium content Inconel 690. So, from this standpoint, they are to be considered as PWR 900 replacement SG first models and their thermal efficiency has consequently to be assessed. This will provide an opportunity of ensuring that the performance of the components delivered is in compliance with requirements and of making the necessary provisions if significant deviations are observed. The EFMT branch, which has been in charge of the instrumentation and acceptance of the different SG first models since the first PWR plants were commissioned, will be responsible for the acceptance tests and the ultimate validation of a performance assessment procedure applicable to the future replacement steam generators. The methods and tests proposed for SG expert appraisal are based on consideration of the importance of primary measurement quality for satisfactory SG assessment and of the new test facilities with which the 900 and 1 300 PWR plants are gradually being equipped. These facilities provide an on-site computer environment for tests compatible with the tools (PATTERN, etc.) used at EFMT and in other departments. This test is the first of this kind performed by EFMT and the test facility of a nuclear power plant. (author). 6 figs

  7. Performance of high burned PWR fuel during transient

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio

    1992-01-01

    In a majority of Japanese light water type commercial powder reactors (LWRs), UO 2 pellet sheathed by zircaloy cladding is used. Licensed discharged burn-up of the PWR fuel rod is going to be increased from 39 MWd/kgU to 48 MWd/kgU. This requests the increased reliability of cladding material as a strong barrier against fission product (FP). A long time usage in the neutron field and in the high temperature coolant will cause the zircaloy hardening and embrittlement. The cladding material is also degraded by waterside corrosion. These degradations are enhanced much by increased burn-up. A increased magnitude of the pellet-cladding mechanical interaction (PCMI) is of importance for increasing the stress of cladding material. In addition, aggressive FPs released from the fuel tends to attack the cladding material to cause stress corrosion cracking (SCC). At the Nuclear Safety Research Reactor (NSRR) in JAERI, 14 x 14 PWR type fuel rods preirradiation up to 42 MWd/kgU was prepared for the transient pulse irradiation under the simulated reactivity initiated accident (RIA) conditions. This will cause a prompt increase of the fuel temperature and stress on the highly burned cladding material. In the present paper, steady-state and transient behavior observed from the tested PWR fuel rod and calculational results obtained from the computer code FPRETAIN will be described. (author)

  8. Benchmarking Computational Fluid Dynamics for Application to PWR Fuel

    International Nuclear Information System (INIS)

    Smith, L.D. III; Conner, M.E.; Liu, B.; Dzodzo, B.; Paramonov, D.V.; Beasley, D.E.; Langford, H.M.; Holloway, M.V.

    2002-01-01

    The present study demonstrates a process used to develop confidence in Computational Fluid Dynamics (CFD) as a tool to investigate flow and temperature distributions in a PWR fuel bundle. The velocity and temperature fields produced by a mixing spacer grid of a PWR fuel assembly are quite complex. Before using CFD to evaluate these flow fields, a rigorous benchmarking effort should be performed to ensure that reasonable results are obtained. Westinghouse has developed a method to quantitatively benchmark CFD tools against data at conditions representative of the PWR. Several measurements in a 5 x 5 rod bundle were performed. Lateral flow-field testing employed visualization techniques and Particle Image Velocimetry (PIV). Heat transfer testing involved measurements of the single-phase heat transfer coefficient downstream of the spacer grid. These test results were used to compare with CFD predictions. Among the parameters optimized in the CFD models based on this comparison with data include computational mesh, turbulence model, and boundary conditions. As an outcome of this effort, a methodology was developed for CFD modeling that provides confidence in the numerical results. (authors)

  9. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    Ciftcioglu, Oe.

    1996-10-01

    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.)

  10. 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.

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

    International Nuclear Information System (INIS)

    Claro, Luiz H.

    2009-01-01

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

  12. A scheme of better utilization of PWR spent fuels

    International Nuclear Information System (INIS)

    Chung, Bum Jin; Kang, Chang Soon

    1991-01-01

    The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is investigated in this study. This scheme of utilizing PWR spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories: refabrication and reconfiguration. For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burn up and distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The results show that most tandem fuel cycle option considered in this study are technically feasible as well as economically viable. (Author)

  13. RCC-C: Design and construction rules for fuel assemblies of PWR nuclear power plants

    International Nuclear Information System (INIS)

    2015-01-01

    The RCC-C code contains all the requirements for the design, fabrication and inspection of nuclear fuel assemblies and the different types of core components (rod cluster control assemblies, burnable poison rod assemblies, primary and secondary source assemblies and thimble plug assemblies). The design, fabrication and inspection rules defined in RCC-C leverage the results of the research and development work pioneered in France, Europe and worldwide, and which have been successfully used by industry to design and build nuclear fuel assemblies and incorporate the resulting feedback. The code's scope covers: fuel system design, especially for assemblies, the fuel rod and associated core components, the characteristics to be checked for products and parts, fabrication methods and associated inspection methods. The RCC-C code is used by the operator of the PWR nuclear power plants in France as a reference when sourcing fuel from the world's top two suppliers in the PWR market, given that the French operator is the world's largest buyer of PWR fuel. Fuel for EPR projects is manufactured according to the provisions of the RCC-C code. The code is available in French and English. The 2005 edition has been translated into Chinese. Contents of the 2015 edition of the RCC-C code: Chapter 1 - General provisions: 1.1 Purpose of the RCC-C, 1.2 Definitions, 1.3 Applicable standards, 1.4 Equipment subject to the RCC-C, 1.5 Management system, 1.6 Processing of non-conformances; Chapter 2 - Description of the equipment subject to the RCC-C: 2.1 Fuel assembly, 2.2 Core components; Chapter 3 - Design: Safety functions, operating functions and environment of fuel assemblies and core components, design and safety principles; Chapter 4 - Manufacturing: 4.1 Materials and part characteristics, 4.2 Assembly requirements, 4.3 Manufacturing and inspection processes, 4.4 Inspection methods, 4.5 Certification of NDT inspectors, 4.6 Characteristics to be inspected for the

  14. Analysis of differences in fuel safety criteria for WWER and western PWR nuclear power plants

    International Nuclear Information System (INIS)

    2003-11-01

    In 2001 the OECD issued a report of the NEA/CSNI (Committee on the Safety of Nuclear Installations) Task Force on the existing safety criteria for reactor fuel for western LWR nuclear power plants (both for PWRs and BWRs) under new design elements. Likewise in 2001, the IAEA released a report by a Working Group on the existing safety criteria for reactor fuel for WWER nuclear power plants under new design requirements. However, it was found that it was not possible to compare the two sets of criteria on the basis upon which they had been established. Therefore, the IAEA initiated an assessment of the common features and differences in fuel safety criteria between plants of eastern and western design, focusing on western PWRs and eastern WWER reactors. Between October 2000 and November 2001, the IAEA organized several workshops with representatives from eastern and western European countries in which the current fuel safety related criteria for PWR and WWER reactors were reviewed and compared. The workshops brought together expert representatives from the Russian Federation, from the Ukraine and from western countries that operate PWRs. The first workshop focused on a general overview of the fuel safety criteria in order for all representatives to appreciate the various criteria and their respective bases. The second workshop (which involved one western and one eastern expert) concentrated on addressing and explaining the differences observed, and documenting all these results in preparation for a panel discussion. This panel discussion took place during the third workshop, where the previously obtained results were reviewed in detail and final recommendations were made. This report documents the findings of the workshops. It highlights the common features and differences between PWR and WWER fuel, and may serve as a general basis for the safety evaluation of these fuels. Therefore, it will be very beneficial for licensing activities for PWR and WWER plants, as it

  15. THYDE-P, PWR LOCA Thermohydraulic Transient Analysis

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2001-01-01

    1 - Description of problem or function: THYDE-P1 analyzes the behaviour of LWR plants in response to various disturbances, including the thermal hydraulic transient following a break of the primary coolant pipe system, generally referred to as a loss-of-coolant-accident (LOCA). LOCA can be considered as the most critical condition for testing the methods and models for plant dynamics, since thermal hydraulic conditions in the system change drastically during the transient. THYDE-P is capable of a complete LOCA calculation from start to complete reflooding of the core by subcooled water. The program performs steady-state adjustment, which is complete in the sense that the steady state obtained is a set of exact solutions of all the transient equations without time derivatives, not only for plant hydraulics but also for all the other phenomena in the simulation of a PWR plant. THYDE-P2 contains among others the following improvements over THYDE-P1: (1) not only the mass and momentum equations but also the energy equation are included in the non-linear implicit scheme; (2) the valve model is implemented; (3) the relaxation equation for void fraction is theoretically derived; (4) vectorized programming is implemented; (5) both EM (evaluation mode) and BE (best estimate) calculations are possible. THYDE-W is an improved version of THYDE-P2 and contains the following additional features: (a) analysis of multiple number of disjoint loops is possible; (b) a control system simulation model is included; (c) the trip model has been improved; (d) heavy water is allowed as coolant; (e) the effect of drift flux is accounted for in the steady state calculation; (f) boron transport is included; (g) to obtain steady state loop heat balance, the option of adjusting the enthalpy distribution is prepared included in addition to that of adjusting heat exchanger areas; (h) to obtain steady state pressure distribution, three other options are prepared in addition to the original ones

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

    International Nuclear Information System (INIS)

    Fang Baoguo; Zhang Dafa; Lin Yajun

    2006-01-01

    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)

  17. Validation of full core geometry model of the NODAL3 code in the PWR transient Benchmark problems

    International Nuclear Information System (INIS)

    T-M Sembiring; S-Pinem; P-H Liem

    2015-01-01

    The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16 % occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4 % for C2 case. All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. (author)

  18. 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)

  19. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  20. PWR pressure vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.

    1981-01-01

    Pressurized water reactors are susceptible to certain types of hypothetical accidents that under some circumstances, including operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of propagation of crack-like defects in the vessel wall. The accidents of concern are those that result in thermal shock to the vessel while the vessel is subjected to internal pressure. Such accidents, referred to as pressurized thermal shock or overcooling accidents (OCA), include a steamline break, small-break LOCA, turbine trip followed by stuck-open bypass valves, the 1978 Rancho Seco and the TMI accidents and many other postulated and actual accidents. The source of cold water for the thermal shock is either emergency core coolant or the normal primary-system coolant. ORNL performed fracture-mechanics calculations for a steamline break in 1978 and for a turbine-trip case in 1980 and concluded on the basis of the results that many more such calculations would be required. To meet the expected demand in a realistic way a computer code, OCA-I, was developed that accepts primary-system temperature and pressure transients as input and then performs one-dimensional thermal and stress analyses for the wall and a corresponding fracture-mechanics analysis for a long axial flaw. The code is briefly described, and its use in both generic and specific plant analyses is discussed

  1. Abnormality diagnosis device for PWR type reactor

    International Nuclear Information System (INIS)

    Kurosawa, Masahiko.

    1993-01-01

    The device of the present invention rapidly detects a small amount of primary coolants leaked from a heat transfer pipes in a steam generator to estimate the scale of the rupture and forecast the transition of the amount of the leakage. That is, a gamma-ray spectrum and dose rate measuring system having a bypass line capable of decaying short half-time nuclides disposed to each of second main steam pipelines and primary coolant pipelines. Data obtained from the measuring systems are compared, to judge the presence of abnormal symptom due to leakage and further, radioactivity concentration in main steams is calculated based on the dose rate at each of the measuring points. Further, radioactivity concentration of the leakage from the secondary main steam pipelines is calculated by change with passage of time. With such procedures, when primary coolants are leaked to the secondary main steam pipelines, the dose rate of the measured system is changed, thereby enabling to recognize the state at the abnormal point. The transition of abnormalities can be forecast with reference to the dose rate by change with passage of time. Further, state of the abnormal point can be recognized based on the gamma-ray spectrum, which is inherent upon occurrence of fuel rupture. (T.S.)

  2. Safety aspects of the using Gd as burnable poison in PWR's

    International Nuclear Information System (INIS)

    Vandenberg, C.; Bonet, H.; Charlier, A.

    1978-01-01

    The experience of BELGONUCLEAIRE in using Gd in LWR's has indicated the safety related advantages of this burnable poison. The successfully operation of the BR3 PWR power plant with 5% of Gd rods is presented and extrapolated to large PWR's. (authro)

  3. Extended blackout mitigation strategy for PWR

    International Nuclear Information System (INIS)

    Prošek, Andrej; Volkanovski, Andrija

    2015-01-01

    to a pressurized water reactor on reference model. Several case scenarios are developed and analysed considering different sources of reactor coolant system inventory loss and procedures to limit this loss. The obtained results show effectiveness of the proposed strategy for mitigation of the event of loss of all electric power sources at the nuclear power plant. The obtained results demonstrate the need for assessment of the pump injection flow rates before the utilization of the pump for mitigation of the event. The applicability of the developed method on operational power plant is validated.

  4. Extended blackout mitigation strategy for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Prošek, Andrej, E-mail: andrej.prosek@ijs.si; Volkanovski, Andrija, E-mail: andrija.volkanovski@ijs.si

    2015-12-15

    to a pressurized water reactor on reference model. Several case scenarios are developed and analysed considering different sources of reactor coolant system inventory loss and procedures to limit this loss. The obtained results show effectiveness of the proposed strategy for mitigation of the event of loss of all electric power sources at the nuclear power plant. The obtained results demonstrate the need for assessment of the pump injection flow rates before the utilization of the pump for mitigation of the event. The applicability of the developed method on operational power plant is validated.

  5. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Provost, Jean-Luc; Debes, Michel

    2005-01-01

    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)

  6. Integral type small PWR with stand-alone safety

    International Nuclear Information System (INIS)

    Makihara, Yoshiaki

    2001-01-01

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

  7. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Garcia, S.; Lynch, N.; Reid, R.

    2014-01-01

    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)

  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)

    Ball, G.

    1991-06-01

    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. 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)

    Hughes, G.; Mueller, R.

    1980-03-01

    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) [de

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

    International Nuclear Information System (INIS)

    S. Mastilovic; A. Scheider; S.M. Bennett

    2001-01-01

    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

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

    International Nuclear Information System (INIS)

    Chung, Chin Man; Kim, Cheol Jung; Kim, Min Suk

    1998-06-01

    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

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

    International Nuclear Information System (INIS)

    Naitoh, T.; Nakahara, Y.

    1987-01-01

    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

  13. Explicit treatment of spectral history effects in PWR design

    International Nuclear Information System (INIS)

    Gavin, P.H.

    1995-01-01

    Spectral history effects in pressurized water reactors (PWRs) are a consequence of spatially distributed and/or time-dependent quantities such as power, moderator temperature, soluble boron concentration, control rod position, etc., defining open-quotes operating conditions.close quotes Operating conditions, global and local, affect neutron spectrum and isotopic reaction rates and thus the evolution of the fuel composition. Any effect that hardens the neutron spectrum, such as elevated temperature or high soluble boron concentration, will increase the fuel conversion ratio and result in more reactive fuel. This paper describes history effects for an 18-month equilibruim cycle of an ABB CE system 80 PWR

  14. Model for calculating the boron concentration in PWR type reactors

    International Nuclear Information System (INIS)

    Reis Martins Junior, L.L. dos; Vanni, E.A.

    1986-01-01

    A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

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

    International Nuclear Information System (INIS)

    Boulot, F.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    1983-05-01

    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. Axial gap formation in P.W.R. fuel pins

    International Nuclear Information System (INIS)

    Roberts, G.; Jones, K.W.

    1978-07-01

    The potential mechanisms of axial gap formation in PWR fuel pins are examined analytically and also using evidence from post-irradiation examination (p.i.e.) investigation. It is concluded that fuel and cladding cannot remain in contact during densification and so the settling of of the fuel stack, which forms the gaps, must be prevented by such things as asperities in the cladding, fuel chips or tilted pellets. Examples from the p.i.e. examination programme are used to support this conclusion. (author)

  18. Fatigue crack growth analysis of a 450 PWR - lateral

    International Nuclear Information System (INIS)

    Taupin, P.; Flamand, F.

    1988-01-01

    Fatigue Crack Growth analysis of a 5 mm deep surface crack in the crotch region of a 45 0 Lateral (12 inch diameter) was performed on a 3-Loop 900 MWe PWR Plant under Normal and upset loading conditions. Stress Intensity factors were computed using the weight-function technique. The latter were obtained for a polynomial stress distribution at the corner of the lateral under contract with the Pressure Vessel Research Committee of the WRC. The study shows that after 40 years of normal operation the size of the end of life crack is limited to about 25 mm for the chosen lateral with a thickness of 300 mm

  19. 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)

  20. PWR core follow calculations using the ELCOS code system

    International Nuclear Information System (INIS)

    Grimm, P.; Paratte, J.M.

    1990-01-01

    The ELCOS code system developed at PSI is used to simulate a cycle of a PWR in which one fifth of the assemblies are MOX fuel. The reactor and the calculational methods are briefly described. The calculated critical boron concentrations and power distributions are compared with the measurements at the plant. Although the critical boron concentration is somewhat overpredicted and the computed power distributions are slightly flatter than the measured ones the results of the calculations agree generally well with the measured data. (author) 1 tab., 8 figs., 6 refs

  1. 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)

    1998-12-31

    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)

  2. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gorman, J.A.; Turner, A.P.L.

    1997-01-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses

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

    International Nuclear Information System (INIS)

    Esteves, A.M.; Silva, A.T. e.

    1992-01-01

    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)

  4. Study of the noise propagation in PWR with coupled codes

    International Nuclear Information System (INIS)

    Verdu, G.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Barrachina, T.

    2011-01-01

    The in-core detectors provide signals of the power distribution monitoring for the Reactor Protection System (RPS). The advanced fuel management strategies (high exposure) and the power upratings for PWR reactor types have led to an increase in the noise amplitude in detectors signals. In the present work a study of the propagation along the reactor core and the effects on the core power evolution of a small perturbation on the moderator density, using the coupled code RELAP5-MOD3.3/PARCSv2.7 is presented. The purpose of these studies is to be able to reproduce and analyze the in-core detector simulated signals. (author)

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

    International Nuclear Information System (INIS)

    Silva Galetti, M.R. da.

    1984-01-01

    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) [pt

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  7. Life management plants at nuclear power plants PWR

    International Nuclear Information System (INIS)

    Esteban, G.

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Parry, A.; Petetrot, J.F.; Vivier, M.J.

    1985-10-01

    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

  9. Upper internals of PWR with coolant flow separator

    International Nuclear Information System (INIS)

    Chevereau, G.; Heuze, A.

    1989-01-01

    The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

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

    International Nuclear Information System (INIS)

    Barroso, D.E.G.

    1982-01-01

    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.) [pt

  11. Conversion ratio and consumption of fissile material in PWR reactors

    International Nuclear Information System (INIS)

    Tiba, C.

    1977-01-01

    It has been shown that the uranium resources will be insufficient for future projected demand. The many solutions to this problem are considered and, in particular, the effect of enrichment on the conversion ratio and hence total uranium comsumption is studied. The developed computacional method employs the one-group neutron diffusion theory. The model is verified by calculating typical burn-up, conversion ratio, U-235 comsumption and plutonium production values in PWR's, and comparing results with those in the published literature. The associated costs of U and U-Pu fuel cycles are also studied for various enrichment values [pt

  12. Comparison of PWR-IMF and FR fuel cycles

    International Nuclear Information System (INIS)

    Darilek, Petr; Zajac, Radoslav; Breza, Juraj; Necas, Vladimir

    2007-01-01

    The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

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

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

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

  14. Underwater welding and repair technologies applied in PWR environment

    International Nuclear Information System (INIS)

    Scandella, Fabrice; Carpreau, Jean-Michel

    2012-01-01

    The authors describe several welding processes and technologies which have been used for underwater applications and which can be applied when repairing components of a PWR type reactor. They address, describe and discuss wet arc welding processes, the peculiarities of underwater welding, and the use of various processes such as 111, 114 and 135 processes, underwater welding with the hybrid plasma MIG-MAG process, underwater welding with the laser wire process, underwater welding with the FSW, FSP or UWFSW processes, underwater welding with variants of the friction welding process (friction surfacing, taper stitch welding, hydro-pillar processing

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

    International Nuclear Information System (INIS)

    Gautier, A.; Miossec, C.

    1985-12-01

    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

  16. Chemical cleaning of PWR steam generators: application at Nogent 1

    International Nuclear Information System (INIS)

    Fiquet, J.M.; Veysset, J.P.; Esteban, L.; Saurin, P.

    1990-01-01

    EDF has developed and patented a chemical cleaning process for PWR steam generators, based on the use of a mixture of organic acids in order to: - dissolve iron oxides and copper with a single solution; - clean dented crevices. Qualification tests have permitted to demonstrate effectiveness of the solution and its inocuousness related to steam generator materials. The process, the license of which belongs to SOMAFER R.A. and FRAMATOME, has been implemented in France at Nogent. The goal was to dissolve iron oxides allowing metallic particles, aggregated on the tubesheet, to be released and mechanically removed. The effectiveness was satisfactory and this treatment is to be extended to other units [fr

  17. Transfer of chemicals in PWR systems: secondary side

    International Nuclear Information System (INIS)

    Jonas, O.

    1978-01-01

    Transfer of chemicals in the secondary side of pressurized water reactor systems with recirculating and once-through steam generators is considered. Chemical data on water, steam and deposit chemistry of twenty-six operating units are given and major physical-chemical processes and differences between the two systems and between fossil and PWR systems are discussed. It is concluded that the limited available data show the average water and steam chemistry to be within recommended limits, but large variations of impurity concentrations and corrosion problems encountered indicate that our knowledge of the system chemistry and chemical thermodynamics, system design, sampling, analysis and operation need improvement. (author)

  18. Condensate polisher application for PWR steam generator corrosion control

    International Nuclear Information System (INIS)

    Sawochka, S.G.; Leibovitz, J.; Siegwarth, D.P.; Pearl, W.L.

    1981-01-01

    The evolution of corrosion attack modes particularly in recirculating U-tube PWR steam generators has dictated a thorough review of the advantages and disadvantages of condensate polishing. Analytical modeling techniques to qualitatively predict crevice chemistry variations resulting from steam generator bulk water variations have allowed valuable insights to be developed. Modeling results complemented by steam generator and laboratory corrosion data will be employed to set condensate demineralizer effluent specifications consistent with control of steam generator corrosion. Laboratory and plant studies are being performed to demonstrate achievability of necessary effluent specifications. (author)

  19. Structure-dynamic model verification calculation of PWR 5 tests

    International Nuclear Information System (INIS)

    Engel, R.

    1980-02-01

    Within reactor safety research project RS 16 B of the German Federal Ministry of Research and Technology (BMFT), blowdown experiments are conducted at Battelle Institut e.V. Frankfurt/Main using a model reactor pressure vessel with a height of 11,2 m and internals corresponding to those in a PWR. In the present report the dynamic loading on the pressure vessel internals (upper perforated plate and barrel suspension) during the DWR 5 experiment are calculated by means of a vertical and horizontal dynamic model using the CESHOCK code. The equations of motion are resolved by direct integration. (orig./RW) [de

  20. The N4 plant: culmination of French PWR experience

    International Nuclear Information System (INIS)

    Bellet, J.; Houyez, A.; Journet, J.; Pierrard, J.H.

    1985-01-01

    The model N4 series of 1400MWe class PWR plants has been developed in France from a unique base of technical and operating experience. It meets the French government's requirement for a reactor free of constraints due to licensing agreements with overseas companies, with enhanced safety features and incorporating the lessons of Three Mile Island. In particular, improvements have been made to the reactor vessel, the steam generators, the primary pumps and control systems. The units are capable of daily load following and extended operation between refuelling. The N4 plant includes a new design of turbine-generator. (author)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  2. 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)

  3. Chooz-B1, the new Electricite de France PWR: calculation scheme of neutron leakages from the reactor cavity

    International Nuclear Information System (INIS)

    Champion, G.; Thiriet, A.; Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.

    1987-04-01

    A new calculation scheme has been set up to assess the neutron field characteristics inside French PWR. In order to take into account multiple neutron scattering and the complexity of the reactor geometry, the use of Monte-Carlo methods have been heavily increased. They are coupled with classical SN.-methods. The main goal aimed at was to find out the neutron field characteristics at the level of the reactor pit openings. These radiation reference sources will be used to check the neutron shielding efficiencies. The new calculation scheme has been applied to CHOOZ-B1, the first unit of the new N4 program. The former results have been compared with the measurement results related to PALUEL-I and II PWR, two units of the previous P4 program. Although the core and the geometry are not entirely similar, it is possible to check with confidence the calculation results along the vessel and at the core midplane level with the measurement results at the same locations. It appears that they are in good agreement. Consequently, the new calculation scheme appears reliable

  4. The Influence of atmospheric conditions to probabilistic calculation of impact of radiology accident on PWR 1000 MWe

    International Nuclear Information System (INIS)

    Pande Made Udiyani; Sri Kuntjoro

    2015-01-01

    The calculation of the radiological impact of the fission products releases due to potential accidents that may occur in the PWR (Pressurized Water Reactor) is required in a probabilistic. The atmospheric conditions greatly contribute to the dispersion of radionuclides in the environment, so that in this study will be analyzed the influence of atmospheric conditions on probabilistic calculation of the reactor accidents consequences. The objective of this study is to conduct an analysis of the influence of atmospheric conditions based on meteorological input data models on the radiological consequences of PWR 1000 MWe accidents. Simulations using PC-Cosyma code with probabilistic calculations mode, the meteorological data input executed cyclic and stratified, the meteorological input data are executed in the cyclic and stratified, and simulated in Muria Peninsula and Serang Coastal. Meteorological data were taken every hour for the duration of the year. The result showed that the cumulative frequency for the same input models for Serang coastal is higher than the Muria Peninsula. For the same site, cumulative frequency on cyclic input models is higher than stratified models. The cyclic models provide flexibility in determining the level of accuracy of calculations and do not require reference data compared to stratified models. The use of cyclic and stratified models involving large amounts of data and calculation repetition will improve the accuracy of statistical calculation values. (author)

  5. Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2017-01-01

    Full Text Available The in-house coupled neutronic and thermal-hydraulic (N/T-H code of BATAN (National Nuclear Energy Agency of Indonesia, NODAL3, based on the few-group neutron diffusion equation in 3-dimensional geometry using the polynomial nodal method, has been verified with static and transient PWR benchmark cases. This paper reports the verification of NODAL3 code in the NEA-NSC PWR uncontrolled control rods withdrawal at zero power benchmark. The objective of this paper is to determine the accuracy of NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient. The benchmark is chosen since many organizations participated using various methods and approximations, so the calculation results of NODAL3 can be compared to other codes’ results. The calculated parameters are performed for the steady-state, transient core averaged, and transient hot pellet results. The influence of radial and axial nodes number was investigated for all cases. The results of NODAL3 code are in very good agreement with the reference solutions if the radial and axial nodes number is 2 × 2 and 2 × 18 (total axial layers, respectively.

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

    Energy Technology Data Exchange (ETDEWEB)

    Esteban, G.

    2014-10-01

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

  7. Results of a recent crud/corrosion fuel risk assessment at a U.S. PWR

    International Nuclear Information System (INIS)

    Lamanna, Larry; Pop, Mike; Gregorich, Carola; Harne, Richard; Jones, John

    2012-09-01

    In order to avoid potential fuel reliability issues, specifically crud-related issues, it is necessary to achieve and maintain a crud safe environment. Therefore, the ability to confidently predict risks associated with crud deposition on fuel becomes critically important. AREVA is applying its cutting-edge PWR Fuel Crud (Primary System corrosion products)/Corrosion Tools, i.e. COBRA-FLX (subchannel-by-subchannel T/H tool) coupled with FDIC (crud deposition tool) to subsequently perform PWR Fuel Crud /Corrosion risk assessments for operating plants in the US. After describing the method, the result of one of these assessments is presented for an operating plant in the US that has experienced recent crud observations/concerns. Both Crud Induced Localized Corrosion (CILC) and Crud Induced Power Shift (CIPS) risk assessment methods, as applied to the upcoming cycle (Cycle N), were compared to the current/on-going cycle (Cycle N-1) and to the previous cycle (Cycle N-2). The results allowed the Utility to consider crud risk management changes associated with the upcoming cycle (Cycle-N). Benchmarking of the AREVA tools, using the plant-specific crud information gained from the crud sampling/characterization for the Unit will be presented. The CIPS analysis references boron loading and the amount of insoluble iron-nickel-borates predicted for Cycles N-2, N-1, and N. The results of the CILC evaluation reference FDIC-predicted crud thickness, cladding temperature under deposit, evolution of CILC bearing species and lithium concentration in the zirconium oxide layer. The approach taken by AREVA during the evaluation was to consider both 'risk' and 'margin' to fuel performance impact caused by crud deposits. The conclusion of the assessment, illustrated by the results presented in this paper, is that the example Plant has sufficient margin in worst case conditions for CIPS and CILC risk in Cycle N, based on Cycle N-1 and Cycle N-2 conditions and behavior

  8. 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)

  9. Fracture toughness behavior of irradiated stainless steel in PWR systems

    Energy Technology Data Exchange (ETDEWEB)

    Xu, H.; Fyfitch, S. [AREVA NP Inc., Lynchburg, Pennsylvania (United States); Tang, H.T. [Electric Power Research Inst., Palo Alto, California (United States)

    2007-07-01

    Data from available research programs were collected and evaluated by the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) to determine the relationship between fracture toughness and neutron fluence for conditions representative of pressurized water reactor (PWR) conditions. It is shown that the reduction of fracture toughness with increasing neutron dose in both boiling water reactors (BWRs) and PWRs is consistent with that observed in fast reactors. The lower bound fracture toughness observed for irradiated stainless steels in PWRs is 38 MPa{radical}m (34.6 ksi{radical}in) at neutron exposures greater than 6.7 X 10{sup 21} n/cm{sup 2} (E > 1.0 MeV) or approximately 10 dpa. For such levels of fracture toughness, it is recommended that linear-elastic fracture mechanics (LEFM) analyses be considered for design and operational analyses. The results from this study can be used by the nuclear industry to assess the effects of irradiation on stainless steels in PWR systems. (author)

  10. Improved liquid waste processing system of PWR plant

    International Nuclear Information System (INIS)

    Suehiro, Kazuyasu

    1977-01-01

    Mitsubishi Heavy Industries, Ltd. has engaged in the improvement and enhancement of waste-processing facilities for PWR power stations, and recently established the improved processing system. With this system, it becomes possible to contain radioactive waste gas semi-permanently within plants and to recycle waste liquid after the treatment, thus to make the release of radioactive wastes practically zero. The improved system has the following features, namely the recycling system is adopted, drain is separated and each separated drain is treated by specialized process, the reboiler type evaporator and the reverse osmosis equipment are used, and the leakless construction is adopted for the equipments. The radioactive liquid wastes in PWR power stations are classified into coolant drain, drain from general equipments, chemical drain and cleaning water. The outline of the improved processing system and the newly developed equipments such as the reboiler type evaporator and the reverse osmosis equipment are explained. With the evaporator, the concentration rate of waste liquid can be raised to about three times, and foaming waste can be treated efficiently. The decontamination performance is excellent. The reverse osmosis treatment is stable and reliable method, and is useful for the treatment of cleaning water. It is also effective for concentrating treatment. The unmanned automatic operation is possible. (Kako, I.)

  11. Application of burnup credit for PWR spent fuel storage pool

    International Nuclear Information System (INIS)

    Shin, Hee Sung; Ro, Seung-Gy; Bae, Kang Mok; Kim, Ik Soo; Shin, Young Joon

    1999-01-01

    A study on the application of burnup credit for a PWR spent fuel storage pool has been investigated using a computer code system such as CSAS6 module of SCALE 4.3 in association with 44-group SCALE cross-section library. The calculation bias of the code system at a 95% probability with a 95% confidence level seems to be 0.00951 by benchmarking the system for forty six experimental data. With the aid of this computer code system, criticality analysis has been performed for the PWR spent fuel storage pool. Uncertainties due to postulated abnormal and accidental conditions, and manufacturing tolerance such as stainless steel thickness of storage rack, fuel enrichment, fuel density and box size have statistically been combined and resulted in 0.00674. Also, isotopic correction factor which was based on the calculated and measured concentration of 43 isotopes for both selected actinides and fission products important in burnup credit application has been taken into account in the criticality analysis. It is revealed that the minimum burnup with the corrected isotopic concentrations as required for the safe storage is 5,730 MWd/tU in enriched fuel of 5.0 wt%. (author)

  12. Assessment of environmentally assisted cracking in PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Tice, D.R.

    1991-01-01

    There is a possibility that extension of pre-existing flaws in the reactor pressure vessel of a pressurised water reactor (PWR) may occur by environmentally assisted cracking, in particular by corrosion fatigue under cyclic transient loading. Crack growth predictions have usually been carried out using cyclic crack growth rate (da/dN) versus stress intensity range (δK) curves, such as those given in Section XI, Appendix A of the ASME Boiler and Pressure Vessel Code. However, the inherent time dependent nature of environmental cracking processes renders such an approach unrealistic. The present paper describes the development of an alternative time based assessment methodology. Illustrative calculations of expected crack growth of assumed defects made using the cyclic (ASME XIA) and time-based approaches are compared. The results illustrate that crack growth predicted by the time-based approach can be greater or less than that calculated by the traditional method. For a PWR operated with good control of water chemistry, actual crack growth rates are expected to be well below those predicted by the ASME code. (Author)

  13. Design and Development of Virtual Reactivity System for PWR

    International Nuclear Information System (INIS)

    Anwar, M. I.

    2012-01-01

    The reactivity monitoring and investigation is an important mean to ensure the safety operation of a nuclear power plant. But the reactivity of the nuclear reactor usually cannot be directly measured. It should be computed with certain estimation method. In this thesis, an effort has been made using an artificial neural network and highly fluctuating experimental data for predicting the total reactivity of the nuclear reactor based on all components of net reactivity. This virtual reactivity system is designed by taking advantage of neural network's nonlinear mapping capability. Based on analysis of the reactivity contributing factors, several neural network models are built separately for control rod, boron, poisons, fuel Doppler Effect and moderator effect. Extensive simulation and validation tests for PWR show that satisfied results have been obtained with the proposed approach. It presents a new idea to estimate the PWR's reactivity using artificial intelligence. All the design and simulation work is carried out in MATLAB and a real time programming environment is chosen for the computation and prediction of reactivity. (author)

  14. A comprehensive in-pile test of PWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang Rixin; Zhang Shucheng; Chen Dianshan (Academia Sinica, Beijing (China). Inst. of Atomic Energy)

    1991-02-01

    An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3x3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 {mu}m. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation. (orig.).

  15. Beta and gamma dose calculations for PWR and BWR containments

    International Nuclear Information System (INIS)

    King, D.B.

    1989-07-01

    Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 x 10 8 rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 x 10 8 rad equipment qualification test region. 8 refs., 23 figs., 12 tabs

  16. Development of MHI PWR fuel assembly with high thermal performance

    International Nuclear Information System (INIS)

    Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

    2005-01-01

    Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

  17. Reverse depletion method for PWR core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.; Kim, Y.J.

    1985-01-01

    Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. Prospects for even greater use of in-board fresh fuel loading are good as utilities seek to reduce core vessel fluence, mitigate pressurized thermal shock concerns, and extend vessel lifetime. Consequently, large numbers of burnable poison (BP) pins are being used to control the power peaking at the in-board fresh fuel positions. This has presented an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of BP pins in the fresh fuel. A procedure was developed that utilizes the well-known Haling depletion to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and BP distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provide for the maximum utility to PWR core reload design

  18. Fiber optic components compatibility with the PWR containment radiation field

    International Nuclear Information System (INIS)

    Breuze, G.; Serre, J.

    1990-01-01

    Present and future applications of fiber optics transmission in the nuclear industrial field are emphasized. Nuclear acceptance criteria for relevant electronic equipments in terms of radiation dose rate, integrated dose and required reliability are given. Ambient conditions of PWR containment are especially considered in the present paper. Experimental results of optical fibers and end-components exposed to 60 Co gamma rays are successively shown. Main radiation response characteristics up to 10 4 Gy (with dose rates of about 100 Gy.h -1 ) of both multimodal fiber families (step index and gradient index fibers) are compared. Predominant features of pure silica core fibers are: * an efficient photobleaching with near IR light from LED and LD commonly used in transmission data links, * a radiation hardening reducing induced losses down to 10 dB.km -1 in fine fibers up to date with latest developments. Dose rate effect on induced losses is also outlined for these fibers. Optoelectronic fiber-end components radiation response is good only for special LED (AsGa) and PD (Si). Radiation behavior of complex pigtailed LDM (laser diode + photodiode + Peltier element + thermistor) is not fully acceptable and technological improvements were made. Preliminary results are given. Two applications of fiber links transmitting data in a PWR containment and a hot cell are described. Hardening levels obtained and means required are given

  19. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni

    1982-10-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    J.S. Tang

    2001-05-03

    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&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.

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

    International Nuclear Information System (INIS)

    J.S. Tang

    2001-01-01

    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

  2. A new approach to PWR power control using intelligent techniques

    International Nuclear Information System (INIS)

    Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.; Sadati, N.

    2004-01-01

    Improved load following capability is one of the main technical performances of advanced PWR(APWR). Controlling the nuclear reactor core during load following operation encounters some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking, while the core is subject to large and sharp variation of local power density during transients. Axial offset (A.O) is the parameter usually used to represent of core power peaking, in form of a practical parameter. This paper, proposes a new intelligent approach to A.o control of PWR nuclear reactors core during load following operation. This method uses a neural network model of the core to predict the dynamic behavior of the core and a fuzzy critic based on the operator knowledge and experience for the purpose of decision-making during load following operations. Simulation results show that this method can use optimum control rod groups maneuver with variable overlapping and may improve the reactor load following capability

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

    International Nuclear Information System (INIS)

    Stainman, J.P.

    1985-01-01

    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

  4. An analysis of transients in the PWR downcomer

    International Nuclear Information System (INIS)

    Jovanovic, A.

    1981-01-01

    The paper deals with the problem of determining non-stationary temperature field in the downcomer of a PWR type reactor. For this purpose, an analytical model has been developed. The model covers five components of (PWR - Krsko) downcomer: the core-barrel, floor between the core-barrel and the thermal shield, the thermal shield, flow between the thermal shield and the reactor vessel wall, the reactor vessel wall. The model includes internal heat generation in metal structures. The governing equations of the model have been written in the finite difference explicit form. The system of resulting algebraic equations was solved bu Gauss-Seidel method, using a modular computer code. Several characteristic transients were examined (step and continuous change of fluid temperature at the inlet nozzle). Also, an analysis of main parameters (heat transfer coefficient and flow rate) has been performed. The model is intended to be used as basics for further development of a more realistic model that could be used for practical safety analysis. (author)

  5. Liquid radioactive waste processing improvement of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Nery, Renata Wolter dos Reis; Martinez, Aquilino Senra; Monteiro, Jose Luiz Fontes

    2005-01-01

    The study evaluate an inorganic ion exchange to process the low level liquid radwaste of PWR nuclear plants, so that the level of the radioactivity in the effluents and the solid waste produced during the treatment of these liquid radwaste can be reduced. The work compares two types of ion exchange materials, a strong acid cation exchange resin, that is the material typically used to remove radionuclides from PWR nuclear plants wastes, and a mordenite zeolite. These exchange material were used to remove cesium from a synthetic effluent containing only this ion and another effluent containing cesium and cobalt. The breakthrough curves of the zeolite and resin using a fix bed reactor were compared. The results demonstrated that the zeolite is more efficient than the resin in removing cesium from a solution containing cesium and cobalt. The results also showed that a bed combining zeolite and resin can process more volume of an effluent containing cesium and cobalt than a bed resin alone. (author)

  6. Four-fluid model of PWR degraded cores

    International Nuclear Information System (INIS)

    Dearing, J.F.

    1985-01-01

    This paper describes the new two-dimensional, four-fluid fluid dynamics and heat transfer (FLUIDS) module of the MELPROG code. MELPROG is designed to give an integrated, mechanistic treatment of pressurized water reactor (PWR) core meltdown accidents from accident initiation to vessel melt-through. The code has a modular data storage and transfer structure, with each module providing the others with boundary conditions at each computational time step. Thus the FLUIDS module receives mass and energy source terms from the fuel pin module, the structures module, and the debris bed module, and radiation energy source terms from the radiation module. MELPROG, which models the reactor vessel, is also designed to model the vessel as a component in the TRAC/PF1 networking solution of a PWR reactor coolant system (RCS). The coupling between TRAC and MELPROG is implicit in the fluid dynamics of the reactor coolant (liquid water and steam) allowing an accurate simulation of the coupling between the vessel and the rest of the RCS during an accident. This paper deals specifically with the numerical model of fluid dynamics and heat transfer within the reactor vessel, which allows a much more realistic simulation (with less restrictive assumptions on physical behavior) of the accident than has been possible before

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

    International Nuclear Information System (INIS)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

    2006-01-01

    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high requirements of

  8. SCC of Alloy 600 components in PWR primary loop

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  9. Proposal for a advanced PWR core with adequate characteristics for passive safety concept

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto

    1999-01-01

    This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

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

    International Nuclear Information System (INIS)

    Yaguchi, Seiji; Yonezawa, Toshio

    2012-01-01

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

  11. Interface tracking simulations of bubbly flows in PWR relevant geometries

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Jun, E-mail: jfang3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Rasquin, Michel, E-mail: michel.rasquin@colorado.edu [Aerospace Engineering Department, University of Colorado, Boulder, CO 80309 (United States); Bolotnov, Igor A., E-mail: igor_bolotnov@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

    2017-02-15

    Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D{sub h}) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in

  12. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

    International Nuclear Information System (INIS)

    MAJI, A. K.; MARSHALL, B.

    2000-01-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation FR-om nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  13. Status and future perspectives of PWR and comparing views on WWER fuel technology

    International Nuclear Information System (INIS)

    Weidinger, H.

    2003-01-01

    The main purpose of this paper is to give an overview on status and future perspectives of the Western PWR fuel technology. For easer understanding and correlating, some comparing views to the WWER fuel technology are provided. This overview of the PWR fuel technology of course can not go into the details of the today used designs of fuel, fuel rods and fuel assemblies. However, it tries to describe the today achieved capability of PWR fuel technology with regard to reliability, efficiency and safety

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

    International Nuclear Information System (INIS)

    Kelly, J.L.

    1978-09-01

    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

  15. The research on burnup characteristic of doping burnable poison in PWR

    International Nuclear Information System (INIS)

    Qiang Shenglong; Qin Dong; Chai Xiaoming; Yao Dong

    2014-01-01

    In PWR core design, burnable poisons are usually used for reactive compensation and power flatten. The choice of burnable poisons and how to match burnup would be the key-points for a long-life core design. We study the burnup character of doping burnable poisons (such as natural element, manual nuclide and soluble boron) in the PWR by the core burnup code MOI based on Monte Carlo method. The results show that Hf, Er and Eu doping burnable poison would be applicable for the nuclear design research on the long-life PWR core. (authors)

  16. On site PWR fuel inspection measurements for operational and design verification

    International Nuclear Information System (INIS)

    1996-01-01

    The on-site inspection of irradiated Pressurized Water Reactor (PWR) fuel and Non-Fuel Bearing Components (NFBC) is typically limited to visual inspections during refuelings using underwater TV cameras and is intended primarily to confirm whether the components will continue in operation. These inspections do not normally provide data for design verification nor information to benefit future fuel designs. Japanese PWR utilities and Nuclear Fuel Industries Ltd. designed, built, and performed demonstration tests of on-site inspection equipment that confirms operational readiness of PWR fuel and NFBC and also gathers data for design verification of these components. 4 figs, 3 tabs

  17. Test requirements for the integral effect test to simulate Korean PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  18. Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Sung-Min; Oh, Seung-Jong [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Ain Shams Univ., Cairo (Egypt). Mechanical Power Engineering Dept.

    2017-12-15

    This paper focuses on the in-vessel phase of Severe Accident Management (SAM) strategy for a Korean 1000 MWe Pressurized Water Reactor (PWR) with reference to ROAAM+ framework approach. To apply ROAAM+, it is needed to identify epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is RCS depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, a sensitivity analysis is carried out to assess the impact of the cut-off porosity below which the flow area of a core node is zero (EPSCUT), and the critical temperature for cladding rupture (TCLMAX) on the core melting and relocation process. In this paper, the SAM strategy for maintaining the integrity of RPV is derived after quantification of the scenario and phenomenological uncertainties.

  19. ASTEC-CATHARE2 benchmarks on French PWR 1300MWe reactors

    International Nuclear Information System (INIS)

    Tregoures, Nicolas; Philippot, Marc; Foucher, Laurent; Guillard, Gaetan; Fleurot, Joelle

    2009-01-01

    The French Institut de Radioprotection et de Surete Nucleaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe reactors. This PSA-2 is heavily relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, an important series of benchmarks with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe accident scenarios. The present paper details 2 out of the 14 studied scenarios: a 12 inches cold leg Loss of Coolant Accident (LOCA) and a 2 tubes Steam Generator Tube Rupture (SGTR). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and the ASTEC results of the core degradation phase are presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results. (author)

  20. Considerations of the manner of accounting for fast fracture risk in the design of PWR vessels

    International Nuclear Information System (INIS)

    Pellissier-Tanon, A.; Grandemange, J.M.

    1986-01-01

    The French approach to the prevention of fast fracture in PWR vessels is to consider it as a whole and to choose the most convenient way to meet this general goal from an economic and technical point of view. According to this approach, there are no specific limits imposed on such factors as end of life RTsub(NDT) or neutron fluence, which are taken as criteria in other countries. The RCCM design and construction code specifications on chemical content and RTsub(NDT) for beltline and non-irradiated parts establish a sound basis for safety. However, for the most critical parts, the existence of large margins with respect to fast fracture is demonstrated by analysis for all second, third and fourth category design transients. To this aim, the RCCM code needs to demonstrate specified safety margins, depending on the transient category, for reference defects defined in kind and size, in order to bound realistically any defects which have a chance of occurring in the part during manufacture. This approach, which enables the disclosure of the influence of all significant design factors on fracture risk, ensures the most consistent way to improve design safety. (author)

  1. Considerations of the manner of accounting for fast fracture risk in the design of PWR vessels

    Energy Technology Data Exchange (ETDEWEB)

    1986-01-01

    The French approach to the prevention of fast fracture in PWR vessels is to consider it as a whole and to choose the most convenient way to meet this general goal from an economic and technical point of view. According to this approach, there are no specific limits imposed on such factors as end of life RTsub(NDT) or neutron fluence, which are taken as criteria in other countries. The RCCM design and construction code specifications on chemical content and RTsub(NDT) for beltline and non-irradiated parts establish a sound basis for safety. However, for the most critical parts, the existence of large margins with respect to fast fracture is demonstrated by analysis for all second, third and fourth category design transients. To this aim, the RCCM code needs to demonstrate specified safety margins, depending on the transient category, for reference defects defined in kind and size, in order to bound realistically any defects which have a chance of occurring in the part during manufacture. This approach, which enables the disclosure of the influence of all significant design factors on fracture risk, ensures the most consistent way to improve design safety.

  2. Test requirements for the integral effect test to simulate Korean PWR plants

    International Nuclear Information System (INIS)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K.

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time

  3. Thermal hydraulic design of a hydride-fueled inverted PWR core

    International Nuclear Information System (INIS)

    Malen, J.A.; Todreas, N.E.; Hejzlar, P.; Ferroni, P.; Bergles, A.

    2009-01-01

    An inverted PWR core design utilizing U(45%, w/o)ZrH 1.6 fuel (here referred to as U-ZrH 1.6 ) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH 1.6 . The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW t , which is 135% of the optimally powered standard design (5080 MW t -determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.

  4. Comparative study T-type and I-type layout of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Eko Rudi Iswanto and Siti Alimah

    2010-01-01

    Determining plant layout is one of the five major stages during the life time of a nuclear power plant. Some important factors that affect in the selecting of plant layout are availability of infrastructure, economic aspects, social aspects, public and environment safety, and also easy to do. Another factor to be considered is requirements as seismic design, which refers to the principles of good security workers, communities and the environment of radiological risks. There are many layout types of nuclear power plant, two of them are T-type layout and I-type layout. Each type of the plant layout has advantage and disadvantage, therefore this study is to understand them. Good layout is able to provide a high level of security against earthquakes. In term of earthquake design, I-type layout has a higher security level than T-type layout. Therefore, I-type layout can be a good choice for PWR nuclear power plants 1000 MWe that will be built in Indonesia. (author)

  5. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  6. The database system for the management of technical documentations of PWR fuel design project using CD-ROM

    International Nuclear Information System (INIS)

    Park, Bong Sik; Lee, Won Jae; Ryu, Jae Kwon; Jo, In Hang; Chang, Jong Hwa.

    1996-12-01

    In this report, the database system developed for the management of technical documentation of PWR fuel design project using CD-ROM (compact disk - read only memory) is described. The database system, KIRDOCM (KAERI Initial and Reload Fuel project technical documentation management), is developed and installed on PC using Visual Foxpro 3.0. Descriptions are focused on the user interface of the KIRDOCM. Introduction addresses the background and concept of the development. The main chapter describes the user requirements, the analysis of computing environment, the design of KIRDOCM, the implementation of the KIRDOCM, user's manual of KIRDOCM and the maintenance of the KIRDOCM for future improvement. The implementation of KIRDOCM system provides the efficiency in the management, maintenance and indexing of the technical documents. And, it is expected that KIRDOCM may be a good reference in applying Visual Foxpro for the development of information management system. (author). 2 tabs., 13 figs., 8 refs

  7. Probes for inspections of heat exchanges installed at nuclear power plants type PWR by eddy current method

    International Nuclear Information System (INIS)

    Silva, Alonso F.O.

    2007-01-01

    From all non destructive examination methods usable to perform integrity evaluation of critical equipment installed at nuclear power plants (NPP), eddy current test (ET) may be considered the most important one, when examining heat exchangers. For its application, special probes and reference calibration standards are employed. In pressurized water reactor (PWR) NPPs, a particularly critical equipment is the steam generator (SG), a huge heat exchanger that contains thousands of U-bend thin wall tubes. Due to its severe working conditions (pressure and temperature), that component is periodically examined by means of ET. In this paper a revision of the operating fundamentals of the main ET probes, used to perform SG inspections is presented. (author)

  8. Generic Crystalline Disposal Reference Case

    Energy Technology Data Exchange (ETDEWEB)

    Painter, Scott Leroy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chu, Shaoping [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Harp, Dylan Robert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Perry, Frank Vinton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wang, Yifeng [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-02-20

    A generic reference case for disposal of spent nuclear fuel and high-level radioactive waste in crystalline rock is outlined. The generic cases are intended to support development of disposal system modeling capability by establishing relevant baseline conditions and parameters. Establishment of a generic reference case requires that the emplacement concept, waste inventory, waste form, waste package, backfill/buffer properties, EBS failure scenarios, host rock properties, and biosphere be specified. The focus in this report is on those elements that are unique to crystalline disposal, especially the geosphere representation. Three emplacement concepts are suggested for further analyses: a waste packages containing 4 PWR assemblies emplaced in boreholes in the floors of tunnels (KBS-3 concept), a 12-assembly waste package emplaced in tunnels, and a 32-assembly dual purpose canister emplaced in tunnels. In addition, three failure scenarios were suggested for future use: a nominal scenario involving corrosion of the waste package in the tunnel emplacement concepts, a manufacturing defect scenario applicable to the KBS-3 concept, and a disruptive glaciation scenario applicable to both emplacement concepts. The computational approaches required to analyze EBS failure and transport processes in a crystalline rock repository are similar to those of argillite/shale, with the most significant difference being that the EBS in a crystalline rock repository will likely experience highly heterogeneous flow rates, which should be represented in the model. The computational approaches required to analyze radionuclide transport in the natural system are very different because of the highly channelized nature of fracture flow. Computational workflows tailored to crystalline rock based on discrete transport pathways extracted from discrete fracture network models are recommended.

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

    International Nuclear Information System (INIS)

    Bilodid, Y.; Mittag, S.

    2011-01-01

    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)

    Huang Fuduan; Yu Degui; Lu Jingju; Ding Dejun; Zhao Yukun

    1991-02-01

    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. Electrochemical measurements in PWR steam generators to follow crevice chemistry

    International Nuclear Information System (INIS)

    Feron, D.

    1991-01-01

    In PWR steam generator crevices, the evolution of chemistry is important for the understanding of corrosion phenomena. Electrochemical measurements have been performed in high temperature simulated crevice environments in order to follow hideout processes and remedial actions (on-line addition of boric acid). Reported tests have been conducted with model boilers of AJAX facilities. Eccentric and concentric tube support plate crevices have been instrumented with platinum electrodes. Electrochemical measurements have been collected when model boiler was under nominal conditions (primary temperature: 335 deg C, secondary temperature: 280 deg C). They include Electrochemical Impedance Spectroscopy (EIS) and potential measurements: with EIS, sodium and boric acid hideouts have been detected and followed. Potential measurements have been performed in an attempt to measure crevice PH evolution

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

    International Nuclear Information System (INIS)

    Wu, C.H.; Lin, T.J.; Kao, T.M.

    2001-01-01

    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)

  13. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  14. 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.

  15. Characteristics of several equilibrium fuel cycles of PWR

    International Nuclear Information System (INIS)

    Waris, Abdul; Sekimoto, Hiroshi

    2001-01-01

    This paper evaluated the influence of neutron spectrum on characteristics of several equilibrium fuel cycles of pressurized water reactor (PWR). In this study, five kinds of fuel cycles were investigated. Required uranium enrichment, required natural uranium amount, and toxicity of heavy metals (HMs) in spent fuel were presented for comparison. The results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined heavy nuclides when uranium is discharged from the reactor. On the other hand, when uranium is totally confined, the enrichment becomes extremely high. The confinement of plutonium and minor actinides (MA) seems effective in reducing radio-toxicity of discharged wastes. By confining all heavy nuclides except uranium those three characteristics could be reduced considerably. For this fuel cycle the toxicity of HMs in spent fuel become nearly equal to or less than that of loaded uranium. (author)

  16. Reduced scale PWR passive safety system designing by genetic algorithms

    International Nuclear Information System (INIS)

    Cunha, Joao J. da; Alvim, Antonio Carlos M.; Lapa, Celso Marcelo Franklin

    2007-01-01

    This paper presents the concept of 'Design by Genetic Algorithms (DbyGA)', applied to a new reduced scale system problem. The design problem of a passive thermal-hydraulic safety system, considering dimensional and operational constraints, has been solved. Taking into account the passive safety characteristics of the last nuclear reactor generation, a PWR core under natural circulation is used in order to demonstrate the methodology applicability. The results revealed that some solutions (reduced scale system DbyGA) are capable of reproducing, both accurately and simultaneously, much of the physical phenomena that occur in real scale and operating conditions. However, some aspects, revealed by studies of cases, pointed important possibilities to DbyGA methodological performance improvement

  17. B ampersand W PWR advanced control system algorithm development

    International Nuclear Information System (INIS)

    Winks, R.W.; Wilson, T.L.; Amick, M.

    1992-01-01

    This paper discusses algorithm development of an Advanced Control System for the B ampersand W Pressurized Water Reactor (PWR) nuclear power plant. The paper summarizes the history of the project, describes the operation of the algorithm, and presents transient results from a simulation of the plant and control system. The history discusses the steps in the development process and the roles played by the utility owners, B ampersand W Nuclear Service Company (BWNS), Oak Ridge National Laboratory (ORNL), and the Foxboro Company. The algorithm description is a brief overview of the features of the control system. The transient results show that operation of the algorithm in a normal power maneuvering mode and in a moderately large upset following a feedwater pump trip

  18. Experiments on the risk of sump plugging on French PWR

    International Nuclear Information System (INIS)

    Armand, Y.; Mattei, J.M.; Vicena, I.; Gubco, V.; Batalik, J.; Murani, J.; Davydov, M.; Melikhov, O.I.; Blinkov, V.N.

    2003-01-01

    The 'Institut de Radioprotection et de Surete Nucleaire' has decided to perform an experimental program of studies on the risk of sump plugging in a 900 MWe pressurized water reactor (PWR). A general overview of the literature has been conducted between October 1999 and November 2000, ending by the definition of an approach methodology and the written of technical specifications. The problem identified gave rise to an European call for tenders leading to four contracts signed in november 2001. Three contracts are managed by VUEZ (Slovakia), the last by EREC (Russia). The methodology and the facilities devoted to these studies are presented in this paper, the preliminary results observed are also presented. The objectives for 2003 to finalize the program and the safety assessment are presented. (authors)

  19. Contribution to the experimental qualification of PWR fuel storage calculations

    International Nuclear Information System (INIS)

    Marsault, Philippe.

    1980-12-01

    Experiments were carried out on assemblies representative of those used in PWR reactors in a configuration made critical with a driver zone. In this way, certain parameters were able to be measured using current classical techniques. As the multiplication factor for a group of assemblies cannot be determined directly, substitutions were made with an equivalent homogeneous lattice in which Laplacian measurements could be made. The k(infinite) factor was obtained by introducing a migration area which can only be obtained from calculations. Experimental storage studies realized during the CRISTO 1 campaign utilize: 1) a lattice with 4 14x14 pin assemblies immersed in ordinary water; 2) a lattice with 4 14x14 pin assemblies and 3) a regular lattice. The CRISTO experiment enabled criticality calculations to be qualified with these lattices for storage under accidental conditions [fr

  20. Abnormal transient analysis by using PWR plant simulator, (2)

    International Nuclear Information System (INIS)

    Naitoh, Akira; Murakami, Yoshimitsu; Yokobayashi, Masao.

    1983-06-01

    This report describes results of abnormal transient analysis by using a PWR plant simulator. The simulator is based on an existing 822MWe power plant with 3 loops, and designed to cover wide range of plant operation from cold shutdown to full power at EOL. In the simulator, malfunctions are provided for abnormal conditions of equipment failures, and in this report, 17 malfunctions for secondary system and 4 malfunctions for nuclear instrumentation systems were simulated. The abnormal conditions are turbine and generator trip, failure of condenser, feedwater system and valve and detector failures of pressure and water level. Fathermore, failure of nuclear instrumentations are involved such as source range channel, intermediate range channel and audio counter. Transient behaviors caused by added malfunctions were reasonable and detail information of dynamic characteristics for turbine-condenser system were obtained. (author)

  1. Hydrodynamical tests with an original PWR heat removal pump

    International Nuclear Information System (INIS)

    Wietstock, P.

    1984-01-01

    GKSS-Forschungszentrum performes hydrodynamical tests with an original PWR heat removal pump to analyse the influences of fluid parameters on the capacity and cavitation behavior of the pump in order to get further improvements of the quantification of the reached safety-level. It can be concluded, that in case of the tested heat removal pump the additional loads during transition from cavitation free operation into fully cavitation for the investigated operation point with 980 m 3 /h will be smaller than the alteration of loads during passing through the total characteristic. The results from cavitation tests for other operation points indicate, that this very important consequence especially for accident operation will be valid for the total specified pump flow area. (orig.)

  2. Assessment of environmentally assisted cracking in PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Tice, D.R.

    1987-01-01

    1) Since environmentally assisted cracking (EAC) is a time dependent process, assessment should be based on time rather than cycle dependent parameters. Thus an a/sub e/ vs a/sub i/ (or strain rate) basis for assessment should be used in preference to da/dN vs ΔK. 2) The threshold strain rate or velocity for the onset of EAC is controlled by material and environmental factors (e.g. steel sulphur content and water chemistry), and possibly by mechanical loading factors such as R ratio and load interaction effects. Above the threshold, crack growth rates are usually unacceptably rapid. 3) Sample calculations show that predicted crack growth rates using a time based model can be below or above those calculated using ASME XI depending on the value of the EAC threshold velocity but that for normal PWR operating conditions rates are likely to be below those predicted by the ASME code

  3. Experimental study on reflooding in advanced tight lattice PWR

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    1982-04-01

    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.)

  5. Tritium target performance during an LBLOCA in a PWR

    International Nuclear Information System (INIS)

    Reid, B.D.

    1996-01-01

    In December 1995, the U.S. Department of Energy (DOE) announced a preferred strategy for acquiring a new supply of tritium. That strategy is based on pursuing the two most promising production alternatives. These alternatives include either constructing an accelerator-produced tritium system for tritium production or procuring an existing commercial light water reactor or irradiation services from such a reactor to irradiate tritium targets. This paper discusses the safety performance of a tritium target in a commercial pressurized water reactor (PWR). The current conceptual design for the light water tritium targets is quite similar, in terms of external dimensions and materials, to early designs for stainless steel clad discrete burnable absorbers used in PWRs. The tritium targets nominally consist of an annular lithium aluminate pellet wrapped in a Zircaloy-4 getter and clad with Type 316 stainless steel

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

    International Nuclear Information System (INIS)

    Nordmann, F.

    1980-03-01

    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 [fr

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Blazquez, J.; Vallejo, I.

    1998-01-01

    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)

  10. Non linear identification applied to PWR steam generators

    International Nuclear Information System (INIS)

    Poncet, B.

    1982-11-01

    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 [fr

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

    International Nuclear Information System (INIS)

    Reparaz, A.; Smith, M.H.; Stephens, L.G.

    1992-01-01

    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

  12. PWR pressurizer discharge piping system on-site testing

    International Nuclear Information System (INIS)

    Anglaret, G.; Lasne, M.

    1983-08-01

    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

  13. Delayed phenomena analysis from French PWR containment instrumentation system

    International Nuclear Information System (INIS)

    Costaz, J.L.

    1987-01-01

    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.)

  14. Optimization of the decontamination in EDF PWR power plants

    International Nuclear Information System (INIS)

    Gosset, P.; Dupin, M.; Buisine, D.; Buet, J.F.; Brunel, V.

    2002-01-01

    The optimisation of decontamination in EDF PWR power plants is the result of a permanent collaborative work between the plant operators, the subcontractors, central services of nuclear power division of EDF. This collaborative work enables the saving of all the feedback experience. The main operations carried out on nuclear sites like mechanical decontamination of valves, use of the ''EMMAC'' process on big components (replacement of steam generator, hydraulic parts of the reactor coolant pumps), use of foam on pools walls and divers in highly contaminated pools have been discussed. This paper shows that the choice of decontamination processes is very dependant on the components, on the dose rate reduction to be aimed and on the possibility to treat the waste on site. (authors)

  15. Confirmation tests of PWR surveillance capsule shipping container

    International Nuclear Information System (INIS)

    Tomita, N.; Ue, K.; Ohashi, M.; Asada, K.; Yoneda, Y.

    1980-01-01

    Mitsubishi Heavy Industries, Ltd. carried out the confirmation tests to confirm the reliability of the PWR surveillance capsule shipping container and to collect cask design data using a 10-ton weight full scale model at Kobe Shipyard and Engine Works. This report presents the outline of these tests. The B Type container was a cylinder 3289 mm long, 1080 mm in diameter and designed in accordance with the new modified Japanese regulations similar to IAEA regulation. These tests consist of four 9 m drop tests, two 1 m puncture tests, a fire test and an immersion test. In conclusion, safetyness of this container has been proved and various technical data for cask design were also collected through these tests. (author)

  16. Experimental studies of PWR primary piping under loca

    International Nuclear Information System (INIS)

    Caumette, Pierre; Garcia, J.L.

    1980-07-01

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

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

    International Nuclear Information System (INIS)

    Kumamaru, H.; Kukita, Y.

    1991-01-01

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

  18. Waste management aspects of entire PWR LOOP decontamination

    International Nuclear Information System (INIS)

    Murray, A.P.; Roesmer, J.

    1988-01-01

    The waste management parameters for decontamination of an entire PWR primary circuit have been determined for dilute alkaline-permanganate/citric acid (APCA), LOMI, ozone and cerium acid process variations. APCA processes generate the largest waste volumes; over 140 m 3 (5000 ft 3 ) in some cases. This represents a potential disposal cost of one million dollars. The cation regeneration column makes the greatest contribution to the disposal volume. In contrast, the LOMI process generates approximately half as much waste, but it is expected to contain relatively high metal concentrations (200-800 ppm). The ozone and cerium acid processes product the least waste, usually under 45 m 3 . These waste volume estimates represent considerable fractions of a utility's annual disposal volume. Consequently, improved waste processing technology is required, and several approaches are suggested

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

    International Nuclear Information System (INIS)

    Zauq, M.H.

    1982-02-01

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

  20. Neutronal aspects of PWR control for transient load following

    International Nuclear Information System (INIS)

    Cossic, A.

    1985-01-01

    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 [fr

  1. Overview PWR-Blowdown Heat Transfer Separate-Effects Program

    International Nuclear Information System (INIS)

    White, J.D.

    1978-01-01

    The ORNL Pressurized Water Reactor Blowdown Heat Transfer Program (PWR-BDHT) is a separate-effects experimental study of thermal-hydraulic phenomena occurring during the first 20 sec of a hypothetical LOCA. Specific objectives include the determination, for a wide range of parameters, of time to CHF and the following variables for both pre- and post-CHF: heat fluxes, ΔT (temperature difference between pin surface and fluid), heat transfer coefficients, and local fluid properties. A summary of the most interesting results from the program obtained during the past year is presented. These results are in the area of: (1) RELAP verification, (2) electric pin calibration, (3) time to critical heat flux (CHF), (4) heat transfer coefficient comparisons, and (5) nuclear fuel pin simulation

  2. PWR-blowdown heat transfer separate effects program

    International Nuclear Information System (INIS)

    Thomas, D.G.

    1976-01-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results are obtained from the Thermal-Hydraulic Test Facility (THTF). Supporting experiments are carried out in several additional test loops - the Forced Convection Test Facility (FCTF), an air-water loop, a transient steam-water loop, and a low-temperature water mockup of the THTF heater rod bundle. The studies to date are described

  3. Data for use in UKAEA PWR plant studies

    International Nuclear Information System (INIS)

    Kinnersly, S.R.; Richards, C.G.; O'Mahoney, R.

    1983-05-01

    Plant data represented by the RETRAN, RELAP4 and TRAC models used at Winfrith for studies of pressurised faults and small and large break loss of coolant accidents for the UK PWR are presented together with comparable data for the Sizewell B design taken from the Pre-Construction Safety Report (PCSR). The main components of the plant are described, and modelling issues, which may affect the interpretation and assessment of the data, and the historical development and use of the models, are outlined. The bulk of the report consists of tables of data with supporting figures and text for all the main items of plant modelled in the Winfrith accident studies. The data presented should be adequate to allow assessments of the Winfrith models and results to be carried out and provide a firm basis for the development of models more representative of the Sizewell B PCSR design. (U.K.)

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

    International Nuclear Information System (INIS)

    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

  6. Vulnerability analysis of a PWR to an external event

    International Nuclear Information System (INIS)

    Aruety, S.; Ilberg, D.; Hertz, Y.

    1980-01-01

    The Vulnerability of a Nuclear Power Plant (NPP) to external events is affected by several factors such as: the degree of redundancy of the reactor systems, subsystems and components; the separation of systems provided in the general layout; the extent of the vulnerable area, i.e., the area which upon being affected by an external event will result in system failure; and the time required to repair or replace the systems, when allowed. The present study offers a methodology, using Probabilistic Safety Analysis, to evaluate the relative importance of the above parameters in reducing the vulnerability of reactor safety systems. Several safety systems of typical PWR's are analyzed as examples. It was found that the degree of redundancy and physical separation of the systems has the most prominent effect on the vulnerability of the NPP

  7. Transient analysis of multifailure conditions by using PWR plant simulator

    International Nuclear Information System (INIS)

    Morisaki, Hidetoshi; Yokobayashi, Masao.

    1984-11-01

    This report describes results of the analysis of abnormal transients caused by multifailures using a PWR plant simulator. The simulator is based on an existing 822MWe power plant with 3 loops, and designed to cover wide range of plant operation from cold shutdown to full power at the end of life. Various malfunctions to simulate abnormal conditions caused by equipment failures are provided. In this report, features of abnormal transients caused by concurrence of malfunctions are discussed. The abnormal conditions studied are leak of primary coolant, loss of charging and feedwater flows, and control systems failure. From the results, it was observed that transient responses caused by some of the malfunctions are almost same as the addition of behaviors caused by each single malfunction. Therefore, it can be said that kinds of malfunctions which are concurrent may be estimated from transient characteristics of each single malfunction. (author)

  8. Behaviour of organic iodides under pwr accident conditions

    International Nuclear Information System (INIS)

    Alm, M.

    1982-01-01

    Laboratory experiments were performed to study the behaviour of radioactive methyl iodide under PWR loss-of-coolant conditions. The pressure relief equipment consisted of an autoclave for simulating the primary circuit and of an expansion vessel for simulating the conditions after a rupture in the reactor coolant system. After pressure relief, the composition of the CH 3 sup(127/131)I-containing steam-air mixture within the expansion vessel was analysed at 80 0 C over a period of 42 days. On the basis of the values measured and of data taken from the literature, both qualitative and quantitative assessments have been made as to the behaviour of radioactive methyl iodide in the event of loss-of-coolant accidents. (author)

  9. Failure probability of PWR reactor coolant loop piping

    International Nuclear Information System (INIS)

    Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

    1984-02-01

    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

  10. Design of large steam turbines for PWR power stations

    International Nuclear Information System (INIS)

    Hobson, G.

    1984-01-01

    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)

  11. Analysis of reactivity insertion accidents in PWR reactors

    International Nuclear Information System (INIS)

    Camargo, C.T.M.

    1978-06-01

    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 UO 2 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) [pt

  12. Simplified model of a PWR primary coolant circuit

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  13. Modeling local chemistry in PWR steam generator crevices

    International Nuclear Information System (INIS)

    Millett, P.J.

    1997-01-01

    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

  14. Development of failed element monitoring system for PWR

    International Nuclear Information System (INIS)

    Liu Yupu; Liu Haojie

    2005-01-01

    Aiming at the existent problems of failed element monitoring system in the PWR, the detector, the spiral tube, the neutron-moderator and the shielding of neutron bas improved on in this task. These improvements decrease the backgrounds effectively, raise the work stability of the detectors and resolve the failed element error action problem which can not be resolved for the long time, and the detecting sensitivity is raise ten times. The γ-ray detector is arranged spiral with outside, so the γ-rays with shorter half-life can be detected. The structure of gross gamma detection station has improved, so the solid angle is expanded, the transmissivity of γ-rays and β-rays are increased, and the ratio of signal to background is raised. The measurement instrument has been intellectualized. This system is above criticism for the users in operation. (authors)

  15. System for stress corrosion conditions tests on PWR reactors

    International Nuclear Information System (INIS)

    Castro, Andre Cesar de Jesus

    2007-01-01

    The study of environmentally assisted cracking (EAC) involves the consideration and evaluation of the inherent compatibility between a material and the environment under conditions of either applied or residual stress. EAC is a critical problem because equipment, components and structure are subject to the influence of mechanical stress, water environment of different composition, temperature and different material history. Testing for resistance to EAC is one of the most effective ways to determine the interrelationships among this variables on the process of EAC. Up to now, several experimental techniques have been developed worldwide, which address different aspects of environmental caused damage. Constant loading of CT specimens test is a typical example of test, which is used for the estimation of parameters of stress corrosion cracking. To assess the initiation stages and kinetics of crack growth, the testing facility should allow active loading of specimens in the environment that is close to the actual operation conditions of assessed component. This paper presents a testing facility for stress corrosion cracking to be installed at CDTN, which was designed and developed at CDTN. The facility is used to carry out constant load tests under simulated PWR environment, where temperature, water pressure and chemistry are controlled, which are considered the most important factors in SCC. Also, the equipment operational conditions, its applications, and restrictions are presented. The system was developed to operate at temperature until 380 degree C and pressure until 180 bar. It consists in a autoclave stuck at a mechanical system, responsible of producing load , a water treatment station, and a data acquisition system. This testing facility allows the evaluation of cracking progress, especially at PWR reactor. (author) operational conditions. (author)

  16. Probe for detection of denting in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gerardin, J.P.; Germain, J.L.; Nio, J.C.

    1994-07-01

    In certain types of PWR steam generator, oxide deposits can lead to embedding, and subsequently to deformation of a tube (the phenomenon of ''denting''). Such embedding changes the vibratory behavior of the tubes and can result in fatigue cracking. This type of cracking can also be worsened in the event of improper assembly of the anti-vibration spacer bars supporting the U-bends. To prevent such incidents and provide for effective preventive condition-directed maintenance of its PWR steam generators, EDF has undertaken the study and development of a probe to detect this type of phenomenon. The studies began in 1990 and led to the building of an initial prototype probe. The principle behind the probe consists in inducing vibration in the U-bend and determining the main resonance modes of the tube. Measurements of frequency and amplitude and calculation of damping enable characterization of the mechanical behavior of the U-bend. The most important parameter is damping, for which the value must be sufficiently high to ensure that the tube is not subjected to major vibratory amplitudes during operation. Numerous tests have been performed with the first prototype version of the probe, on a mock-up in the test area and on one of the demounted steam generators on the Dampierre site. These different tests have enabled validation of the operating principle, fine-tuning the process, pinpointing certain mechanical problems in the probe design, and obtaining the first indications as to the real vibratory behavior of U-bends on a steam generator. On the basis of these preliminary tests, the specifications were drawn up for an industrial version of the probe. Following a call for bids and the choice of a manufacturer, work began on fabrication of a new probe model in 1993. This version was delivered at the end of 1993 and testing began in 1994. (authors). 5 figs., 2 tabs

  17. Thermo-mechanical analysis of PWR bolts susceptible to IASCC

    International Nuclear Information System (INIS)

    Matteoli, C.; Hannink, M.H.C.; Blom, F.J.; Marck, S.C. van der; Charpin-Jacobs, F.

    2015-01-01

    Irradiation Assisted Stress Corrosion Cracking (IASCC) is considered a primary ageing issue for the Reactor Pressure Vessel (RPV) internals of Pressurized Water Reactors (PWR). In particular, this complex phenomenon which develops in an environment featuring thermal and mechanical stresses, interaction with corrosive compounds and irradiation, is affecting the bolts connecting the baffles and the formers in the Nuclear Power Plants' RPVs. The baffle-former assembly is the structure that borders the fuel assemblies region, contributing to keep them in position and separating in the radial direction, the core region from the downcomer region. An evaluation of the stresses and temperatures reached in the baffle-former bolts during normal operation was performed by means of a coupled thermo-mechanical study which uses reactor physics calculations to obtain the fluence in the reactor core and as a consequence the heat deposition in the RPV internals. The heat deposition data are coupled with a finite element model of the bolts and the RPV internals in order to perform a complete analysis taking in account thermal, mechanical and radiation loadings. The study is first carried out focusing on a section of the RPV internals, showing a single row of baffle-former bolts. Then the work is extended to the full core height. The model set up in this work, includes an in-depth study of the behavior of the core internals, in particular baffle-former bolts. The model has the capability of understanding the mechanical and thermal behavior of essential internal components in a PWR. (authors)

  18. Study of a Station Blackout Event in the PWR Plant

    International Nuclear Information System (INIS)

    Ching-Hui Wu; Tsu-Jen Lin; Tsu-Mu Kao

    2002-01-01

    On March 18, 2001, a PWR nuclear power plant located in the Southern Taiwan occurred a Station Blackout (SBO) event. Monsoon seawater mist caused the instability of offsite power grids. High salt-contained mist caused offsite power supply to the nuclear power plant very unstable, and forced the plant to be shutdown. Around 24 hours later, when both units in the plant were shutdown, several inadequate high cycles of bus transfer between 345 kV and 161 kV startup transformers degraded the emergency 4.16 kV switchgears. Then, in the Train-A switchgear room of Unit 1 occurred a fire explosion, when the degraded switchgear was hot shorted at the in-coming 345 kV breaker. Inadequate configuration arrangement of the offsite power supply to the emergency 4.16 kV switchgears led to loss of offsite power (LOOP) events to both units in the plant. Both emergency diesel generators (EDG) of Unit 1 could not be in service in time, but those of Unit 2 were running well. The SBO event of Unit 1 lasted for about two hours till the fifth EDG (DG-5) was lined-up to the Train-B switchgear. This study investigated the scenario of the SBO event and evaluated a risk profile for the SBO period. Guidelines in the SBO event, suggested by probabilistic risk assessment (PRA) procedures were also reviewed. Many related topics such as the re-configuration of offsite power supply, the addition of isolation breakers of the emergency 4.16 kV switchgears, the betterment of DG-5 lineup design, and enhancement of the reliability of offsite power supply to the PWR plant, etc., will be in further studies. (authors)

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

    International Nuclear Information System (INIS)

    Tsuda, S.; Otoha, K.; Takiguchi, H.

    2002-01-01

    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)

  20. PWR primary system chemistry control during hot functional testing

    International Nuclear Information System (INIS)

    Reid, Richard D.; Little, Michael J.

    2014-01-01

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