Evaluation of full MOX core capability for a 900 MWe PWR

International Nuclear Information System (INIS)

Joo, Hyung-Kook; Kim, Young-Jin; Jung, Hyung-Guk; Kim, Young-Il; Sohn, Dong-Seong

1996-01-01

Full MOX capability of a PWR core with 900 MWe capacity has been evaluated in view of plutonium consumption and design feasibility as an effective means for spent fuel management. Three full MOX cores have been conceptually designed; for annual cycle, for 18-month cycle, and for 18-month cycle with high moderation lattice. Fissile and total plutonium quantities at discharge are significantly reduced to 60% and 70% respectively of initial value for standard full MOX cores. It is estimated that one full MOX core demands about 1 tonne of plutonium per year to be reloaded, which is equivalent to reprocessing of spent nuclear fuels discharged from five nuclear reactors operating with uranium fuels. With low-leakage loading scheme, a full MOX core with either annual or 18-month cycle can be designed satisfactorily by installing additional rod cluster control system and modifying soluble boron system. Overall high moderation lattice case promises better core nuclear characteristics. (author)

Experimental study of the tritium inventory in the BR3 and extrapolation to a P.W.R. of 900 MWe

International Nuclear Information System (INIS)

Charlier, A.; Gubel, P.; Vandenberg, C.; Haas, D.

1982-01-01

The aim of this report is to evaluate the tritium production and diffusion in uranium and plutonium fuel in the primary circuit of a PWR and to improve the knowledge about the production difference between the two kinds of isotopes. The first part of the work is relative to the experimental PWR BR3, cycle 4A, during which a constant control of the tritium activity has been performed in the primary circuit. These experimental evaluation was compared with the the theoretical estimation of the tritium production during the cycle 4A. From these observations and calculations, a tritium release fraction was deduced and estimated to be 0.81% of the total tritium produced in the fuel. The second part of the work is devoted to post-irradiation examinations on a few uranium and plutonium rods irradiated in the BR3 reactor. The tritium content was measured in the cladding, in the fuel and in the gas plenum for various samples of fuel rods. These results show the relationship between the release rate from the fuel matrix, the linear power and the burnup. The last part of the work is the estimate of the tritium production in a PWR of 900 MWe in operating conditions. The tritium production was calculated for an uranium fuelled core and for a core containing 30% of all plutonium fuel assemblies in a generic power plant of 900 MWe. From this study, it results that the loading with 30% plutonium assemblies at equilibrium increases the tritium balance in the moderator water of less than 5%

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

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

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)

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

Full Text Available Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC. Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih

Modular simulation of the dynamics of a 925 MWe PWR electronuclear type reactor and design of a multivariable control algorithm

International Nuclear Information System (INIS)

Mansouri, S.

1985-06-01

This work has been consecrated to the modular simulation of a PWR 925 MWe power plant's dynamic and to the design of a multivariable algorithm control: a mathematical model of a plant type was developed. The programs were written on a structured manner in order to maximize flexibility. A multivariable control algorithm based on pole placement with output feedback was elaborated together with its correspondent program. The simulation results for different normal transients were shown and the capabilities of the new method of multivariable control are illustrated through many examples

French 900 MWe PWR PSA preliminary results

International Nuclear Information System (INIS)

Lanore, J.M.; Brisbois, J.

1988-10-01

A PSA is performed by the Safety Assessment Department of CEA for a 900 MWe standardized plant. The paper presents the objectives, the scope of the study and the relative preliminary results. Some general insights are drawn, especially the benefit related to the implementation of emergency procedures

Seismic analysis for safety related structures of 900MWe PWR NPP

International Nuclear Information System (INIS)

Liu Wei

2002-01-01

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

Qinshan 300Mwe NPP full scope simulator upgrade

International Nuclear Information System (INIS)

Qi Kelin; Li Qing; Liu Wei, Lai Shengyuan

2006-01-01

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

Obligations and characteristics applicable to the French unit of the 1400 MWe series. Adaptation to the 900 and 1300 MWe series

International Nuclear Information System (INIS)

Conte, M.

1985-10-01

This report presents the directives concerning the obligations and the main characteristics of the nuclear PWR units of 1400 MWe, notified Electricite de France on the 06th of October 1983 by the Industry and Research Department. They reflect the concept of defence in depth [fr

Security of nuclear power in operation. Results from the first PWR 900 MWe stages of Electricity of France (EDF)

Energy Technology Data Exchange (ETDEWEB)

Capel, R; Chaubaron, J F [Electricite de France, 93 - Saint-Denis. Service de la Production Thermique

1980-06-01

The security and reliability objectives of the PWR 900 MWe stages are acquiring particular importance in the present energetic and nuclear context. This article presents the general framework wherein the superintendence and maintenance of plant equipment are situated. E.D.F. applies to all of its activities, the assurance of quality principles. The General Rules of Operating constitute the basic document. The Operating Technical Specifications specify the conditions for the correct operating and safety of the installations. The Organization of Quality handbook sets the rules to be obeyed in the management of all operations. Examples from Fessenhein and Bugey illustrate the subject and elucidate the practical dimension of security. Lastly, the lessons of experience are recalled.

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)

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

International Nuclear Information System (INIS)

Brisbois, J.; Lanore, J.M.

1991-08-01

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

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

International Nuclear Information System (INIS)

Brisbois, J.; Lanore, J.M.

1991-01-01

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

Experimental research progress on passive safety systems of Chinese advanced PWR

International Nuclear Information System (INIS)

Xiao Zejun; Zhuo Wenbin; Zheng Hua; Chen Bingde; Zong Guifang; Jia Dounan

2003-01-01

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

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.

China: the long wait

International Nuclear Information System (INIS)

Anon.

1990-01-01

The repercussions of Tiananmen Square coupled with worse than expected balance of payments problems have been made it even more difficult than usual to discern the likely future direction of China's nuclear power programme. Currently there are three power reactors under construction, an indigenously designed 300MWe PWR at Qinsham and the two 900MWe Framatome units at Daya Bay, Guang-dong. Prior to the events of June the most likely way forward, beyond these units under construction, appeared to be development of 600MWe units for Chinese conditions in a joint venture with Western companies. There are strong pressures to ''go indigenous'', and attempt to develop a home grown 600MWe design or, perhaps more likely, go for repeats of the 300 MWe plant - assuming it can be operated successfully, a point on which some factions of the Chinese nuclear industry have their doubts. Nuclear district heating still has strong advocates, as an environmentally sound way of meeting the country's huge projected increases in primary energy consumption over the coming years. Construction of a 5MWt district heating reactor at Tsinghua University, Beijing, started in 1986, entered operation at the end of 1989 following a test phase and is now supplying 3MWt to heating of university buildings. (author)

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

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

GTHTR 300 economic calculation with Mini G4ECONS as a basis for generation cost of GTHTR 10 MWe calculation

International Nuclear Information System (INIS)

Mochamad Nasrullah; Nurlaila

2014-01-01

The government plan to build Experimental Power Reactor (EPR) requires measurable economic assessment. The purpose of the study was to recalculate Gas Turbine High Temperature Reactor of 300 MWe (GTHTR 300) and compare the results with reference data. Then calculate generation cost of GTHTR 3, 5 and 10 MWe using the scale factor calculation. The methodology used is covered the generation cost calculation using the Mini G4Econs spread sheet models published by IAEA. Result of the verification calculation showed that a relatively similar, which means that the calculation model could be used to calculate for same other cases. Afterward, using scale factor, smaller scale reactor could be calculated. The calculation result show that electricity generation cost of SMR-HTR type with load factor 90% and discount rate 10% for power capacity 3, 5 and 10 MWe are 29.5, 22.68 and 16.17 cents$/kWh respectively. However, because the EPR is planning to be built as a non-commercial power reactors, then 5 % and 3 % of discount rate could be chosen, each of those discount rate will result electricity generation cost of 10.37 cents$/kWh and 8.56 cents$/kWh respectively. These result could be considered by the government for developing SMR type of HTR. (author)

The french 900 MWe PWR PSA results and specificities

International Nuclear Information System (INIS)

Lanore, J.M.

1990-01-01

A probabilistic Safety Assessment has been performed by the Safety Analysis Department of CEA for a 900 MWe standardized plant. The paper presents the objectives, the scope of the study and the level 1 results. Some general insights are drawn, especially the benefit related to the implementation of emergency procedures and the importance of risk during shutdown situations

Some failures of diesel generators during commissioning tests of 1300 MWe PWR

International Nuclear Information System (INIS)

Colas, A.F.; Morzelle, C.

1985-10-01

During commissioning tests of the French 1300 MWe units, which are equipped with different diesel generator from the 900 MWe units, some devices and components failures were experienced. These components include: Alarm sensors on fuel, lubricating, cooling circuits; Injection pumps and speed governors; Fuel delivery; Vibrations of fuel and lubrication lines. This paper shows how and when the above elements can affect the reliability of Diesel-generator units and how commissioning tests should show the defects

Some failures of diesel-generators during commissioning tests of 1300 MWe PWR

International Nuclear Information System (INIS)

Colas, A.F.; Morzelle, C.

1986-01-01

During commissioning tests of the French 1300 MWe units, which are equipped with different diesel generator from the 900 MWe units, some devices and components failures were experienced. These components include: - Alarm sensors on fuel, lubricating, cooling circuits. - Injection pumps and speed governors. - Fuel delivery. - Vibrations of fuel and lubrication lines. This paper will try to show how and when the above elements can affect the reliability of Diesel-generator units and how commissioning tests should show the defects. (authors)

Some failures of diesel-generators during commissioning tests of 1300 MWe PWR

Energy Technology Data Exchange (ETDEWEB)

Colas, A. F. [Commissariat a l' Energie Atomique, Institut de Protection et Surete Nucleaire, Departement d' Analyse de Surete, CEA/IPSN, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, B.P. No. 6, 92260 Fontenay-aux-Roses (France); Morzelle, C. [Service Etudes et Projets Thermiques et Nucleaires, EdF Lyon (France)

1986-02-15

During commissioning tests of the French 1300 MWe units, which are equipped with different diesel generator from the 900 MWe units, some devices and components failures were experienced. These components include: - Alarm sensors on fuel, lubricating, cooling circuits. - Injection pumps and speed governors. - Fuel delivery. - Vibrations of fuel and lubrication lines. This paper will try to show how and when the above elements can affect the reliability of Diesel-generator units and how commissioning tests should show the defects. (authors)

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.)

China: the long wait. [Developments in the nuclear power programme

Energy Technology Data Exchange (ETDEWEB)

1990-03-01

The repercussions of Tiananmen Square coupled with worse than expected balance of payments problems have been made it even more difficult than usual to discern the likely future direction of China's nuclear power programme. Currently there are three power reactors under construction, an indigenously designed 300MWe PWR at Qinsham and the two 900MWe Framatome units at Daya Bay, Guang-dong. Prior to the events of June the most likely way forward, beyond these units under construction, appeared to be development of 600MWe units for Chinese conditions in a joint venture with Western companies. There are strong pressures to ''go indigenous'', and attempt to develop a home grown 600MWe design or, perhaps more likely, go for repeats of the 300 MWe plant - assuming it can be operated successfully, a point on which some factions of the Chinese nuclear industry have their doubts. Nuclear district heating still has strong advocates, as an environmentally sound way of meeting the country's huge projected increases in primary energy consumption over the coming years. Construction of a 5MWt district heating reactor at Tsinghua University, Beijing, started in 1986, entered operation at the end of 1989 following a test phase and is now supplying 3MWt to heating of university buildings. (author).

A through calculation of 1,100 MWe PWR large break LOCA by THYDE-P1 EM model

International Nuclear Information System (INIS)

Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

1984-07-01

THYDE-P1 is a code to analyze both the blowdown and refill-reflood phases of loss-of-coolant accidents (LOCAs) of pressurized water reactors (PWRs). Up to now, THYDE-P1 has been applied to various experiment analyses, which show its high capability to analyze LOCAs as a best estimate (BE) calculation code. In this report, evaluation model (EM) calculation method, especialy in the blowdown and refill phases, is established equivalently to WREM/J2 which is regarded as appropriate for an EM calculation code, and the results of them are compared and discussed. The present calculation was the first executed by THYDE-P1-EM, and was performed as Sample Calculation Run 80 which was a part of a series of THYDE-P sample calculations. The calculation was carried out from the LOCA initiation till 400 seconds for a guillotine break at the cold leg of a commercial 1,100 MWe PWR plant. The calculated results agreed well to that of the WREM/J2 code. (author)

Nuclear power perspective in China

International Nuclear Information System (INIS)

Liu Xinrong; Xu Changhua

2003-01-01

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

Technical Support to an Operating PWR vis-a-vis Safety Analysis

International Nuclear Information System (INIS)

Gul, Subhan; Khan, M.; Chughtai, M. Kamran

2011-01-01

Currently a PWR of 300 MWe capacity CHASNUPP-I is in operation since the year 2000. Technical support being provided includes in-core fuel management and corresponding safety analysis for the reshuffled core for the next cycle. Currently calculation and analysis were performed for Cycle 6 to achieve the safe and economical loading pattern. The technique used is designated as out in mode (modified). In this technique, most of the fresh fuel assemblies are not directly located at the periphery of the core, but near the boundary. This technique has the advantage that without using burnable absorber we can design a low leakage core with extended cycle and maximum batch averaged burnup. (author)

A 600 MWe advanced PWR for the 1990's

International Nuclear Information System (INIS)

Lemon, J.E.; Malloy, J.D.; Allen, R.E.

1987-01-01

The Babcock and Wilcox Company (B and W) and United Engineers and Constructors (UE and C) have prepared a conceptual design of an advanced 600 MWe Presurized Water Nuclear Power Plant. This design utilizes the large body of design and operating experience on PWRs in the U.S. and abroad and incorporates improvements emphasizing simplicity, safety, licensability, ease of construction, operability, reliability and maintainability. Cost and schedule estimates based on U.S. utility experience indicate that this plant design should be competitive with alternate options

Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

International Nuclear Information System (INIS)

1989-02-01

Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

Study of the accidental risk of the German fast breeder prototype reactor SNR-300

International Nuclear Information System (INIS)

Koeberlein, K.

1983-01-01

A fact-finding committee of the German Federal Parliament in July 1980 recommended to perform a 'risk-oriented study' of the SNR-300, the German 300 MWe fast breeder prototype reactor being under construction at Kalkar. The main aim of this study was to allow a comparative safety evaluation between the SNR-300 and a modern PWR, thus to prepare a basis for a political decision on the SNR-300. Methods and main results of the study are presented in this paper. In the first step of the study six groups of accidents have been identified which may initiate core destruction. By reliability analyses, expected frequency of each group has been estimated. Conditional probabilities for conceivable reactor tank failure modes have been analysed. Tank failure after core destruction leads to release of energy and radioactive material into the containment system. Such accident sequences have been pursued further. Based on a number of core destruction initiators and tank failure modes and various combinations of success and failure states of the containment systems, detailed calculations of different containment scenarious were carried out. From the results of the plant systems analysis, five release categories have been defined. Possible effects of external events and releases of radioactivity from the spent fuel storage pool have also been analysed. In order to quantify the degree of uncertainty of the calculated frequencies, subjective probability distributions of fixed, but inaccurately known quantities have been propagated through the calculations. Using the release categories as input, accident consequences were calculated for the site Kalkar. Though the uncertainty bandwidths for the accident frequencies estimated in the SNR-300 analysis are much wider than for the PWR, the analysis indicates that frequencies of severe accidents, and consequences, are smaller for the SNR-300 than for the PWR as analysed in the 'German Risk Study'. (orig.)

Reliability investigation for the ECC subsystem of a 1300 MWe-PWR

International Nuclear Information System (INIS)

Lalovic, M.

1983-01-01

In this study, a fault-tree analysis is used for reliability investigation of Emergency Core Cooling Sub-system of a 1300 MWe pressurised water reactor. Basic assumptions of the study are large break in the reactor coolant system and independence of the pseudo-components. Relatively high non-availability of the sub-system was calculated. Critical component and minimum cut set are determined. (author)

Main problems experienced on diesel generators of French 900 MWe operating units

Energy Technology Data Exchange (ETDEWEB)

Dredemis, Geoffroy; Jude, Francois [Commissariat a l' Energie Atomique, centre d' Etudes Nucleaires de Fontenay-aux-Roses, Institut de Protection et Surete Nucleaire, Departement d' Analyse de Surete, B.P. No. 6, 92260 Fontenay-aux-Roses (France)

1986-02-15

Each unit of all the French nuclear power plant is equipped with two diesel emergency generator sets., For the totality of standards PWRs of 900 MWe, they are identical. We present in this communication the most significative failures met with diesel engines on operating units, such as rupture of fuel injection pipes, breaking of the connecting rods, and cylinder lubrication failures. All these incidents, which affected the emergency power sources of concerned units, had generic characteristics. In view of their potential consequences, it was proceeded in each case to an immediate control of the components concerned of all PWR 900 MWe diesel engines. At the same time, studies were started as to what modifications would permit to solve rapidly each one of the problems met with. (authors)

Twenty-five years of transient counting experience in French PWR units

Energy Technology Data Exchange (ETDEWEB)

Barthelet, B. [Electricite de France (EDF DPN), 93 - Saint-Denis (France); Savoldelli, D.; Fritz, R. [Electricite de France (EDF DPN), 93 - Noisy le Grand (France)

2001-07-01

For nearly twenty five years, EDF has been checking that the actual operating transients are neither more severe nor more numerous than the design basis transients. This activity of transient cycle counting and bookkeeping has enabled EDF to own a database of more than 800 reactor.years for the PWR units. The current method of transient cycle counting is presented. In the paper, we will point out the main results of transient cycle counting and lessons learned. In general, the frequencies of transients are lower than the design frequencies. In few cases, they are higher, such as the transient frequencies of the RCS lines connected to auxiliary systems often due to operating procedures or particular periodic testing. Few periodic tests were not taken into account in the design basis transient file ; they have been detected thanks to the transient cycle counting. In the last 1980's, we achieved the first updating of the design basis transient file for the PWR 900 MWe series. In the early 1990's, we updated the design basis transient file of the PWR 1300 MWe series. In fact, since design and start-up, the operating conditions have been modified (fuel cycle with stretch-out, modification of the hot leg and cold leg temperatures for the PWR 1300 MWe,...). This was the cause of many unclassified transients. In the new design basis transient file, we have created new transients and increased the frequencies of some of them. This has enabled to consider the updated design basis transient file more representative of actual operating transients. For some years, we have increasingly associated the operators with the transient cycle counting concern. We noticed progress (decreased frequencies of most transients). (authors)

Twenty-five years of transient counting experience in French PWR units

International Nuclear Information System (INIS)

Barthelet, B.; Savoldelli, D.; Fritz, R.

2001-01-01

For nearly twenty five years, EDF has been checking that the actual operating transients are neither more severe nor more numerous than the design basis transients. This activity of transient cycle counting and bookkeeping has enabled EDF to own a database of more than 800 reactor.years for the PWR units. The current method of transient cycle counting is presented. In the paper, we will point out the main results of transient cycle counting and lessons learned. In general, the frequencies of transients are lower than the design frequencies. In few cases, they are higher, such as the transient frequencies of the RCS lines connected to auxiliary systems often due to operating procedures or particular periodic testing. Few periodic tests were not taken into account in the design basis transient file ; they have been detected thanks to the transient cycle counting. In the last 1980's, we achieved the first updating of the design basis transient file for the PWR 900 MWe series. In the early 1990's, we updated the design basis transient file of the PWR 1300 MWe series. In fact, since design and start-up, the operating conditions have been modified (fuel cycle with stretch-out, modification of the hot leg and cold leg temperatures for the PWR 1300 MWe,...). This was the cause of many unclassified transients. In the new design basis transient file, we have created new transients and increased the frequencies of some of them. This has enabled to consider the updated design basis transient file more representative of actual operating transients. For some years, we have increasingly associated the operators with the transient cycle counting concern. We noticed progress (decreased frequencies of most transients). (authors)

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)

Evaluation of Two 300 MWe Fourth Generation Pb-Bi Reactor System Concepts

International Nuclear Information System (INIS)

Miller, Laurence F.; Khuram Khan, M.; Williams, Wesley; Mynatt, F.R.

2002-01-01

This paper describes the evaluation of two 300 MWe modular Pb-Bi cooled reactor system concepts that can be field assembled from components shipped on standard rail cars or on trucks. Thus, the largest components must be smaller than 12' x 12' x 80' (3.66 m x 3.66 m x 24.4 m) and should weigh no more than 80 tons. One of these systems utilizes a cylindrical two-loop containment vessel for the core and the other is a slab design. The fuel for both designs consists of standard-sized metallic IFR fuel in 17 x 17 square array assemblies with a pitch-to-diameter ratio of 1.15. The coolant outlet temperature is limited by current material technology, which is estimated to be 550 C. The primary coolant inlet temperature is selected to be 350 C. This is well above the melting temperature of Pb-Bi, and it is expected to be sufficiently high to limit transient-induced thermal stresses to acceptable values. Coolant flow rates through the core and external piping are below 1 m/s. The results from neutronics calculations include power distributions, reactivity coefficients, and fuel depletion, and results from heat transfer calculations include temperatures and flow rates at various locations in the primary and secondary systems. The neutronic design calculations are accomplished by using a discrete ordinate transport code and a cross section processing system developed at Oak Ridge National Laboratory. Two-dimensional flux distributions are obtained with the DOORS system, and ORIGEN-S, coupled with KENO, is used for time-dependent depletion calculations. The thermal-hydraulic design of the core consists of heat transfer and fluid flow calculation for an average channel. The inlet and outlet temperatures, along with the fuel centerline temperature, are determined in conjunction with core flow rates, pumping power, and total power output. This is accomplished by using a lumped parameter steady-state model with a spreadsheet and by using a one-dimensional time-dependent model

Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

Science.gov (United States)

Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

2018-02-01

Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

Domestic design and validation of natural circulation steam generator of China 1000 MWe PWR NPP

International Nuclear Information System (INIS)

Liu, H.Y.; Wang, X.Y.; Wu, G.; Qin, J.M.; Xiong, Ch.H.; Wang, W.; Chen, J.L.; Cheng, H.P.; Zuo, Ch.P.

2005-01-01

In order to meet the requirements of domestic design of China intending built NPP projects, Research Institute of Nuclear Power Operation (RINPO) has achieved design of 1000 MWe NPP steam generator, called RINSG-1000(means 1000MWe SG designed by RINPO), which is based on SG research ,experiments and service experience accumulated by RINPO in more 40 years. Testing validation of two steam generator key technologies, advanced moisture separate device and sludge collector, has been accomplished during the period of 2000 to 2002. This paper describes the design features of RINSG-1000, and provides some validation test results. (authors)

Instrumentation of steam cycle HTR's up to 900 MWe

International Nuclear Information System (INIS)

Leithner, D.E.; Winkenbach, B.

1982-06-01

Due to basic design features and inherent safety qualities in-core instrumentation is not needed in an HTR. Reactor safety requirements can be met by integral measurements. A modest spatial resolving power of the out-of-core instrumentation is sufficient for all operational purposes in small and medium sized steam cycle HTR's. Thus, the instrumentation concept of the THTR 300 MWe prototype reactor can be adopted without major changes for the HTR 450 MWe reactor project, as is demonstrated here for the neutron flux and temperature measurements. (author)

Influence of declivitous secondary air on combustion characteristics of a down-fired 300-MWe utility boiler

Energy Technology Data Exchange (ETDEWEB)

Zhengqi Li; Feng Ren; Zhichao Chen; Zhao Chen; Jingjie Wang [Harbin Institute of Technology, Harbin (China). School of Energy Science and Engineering

2010-02-15

Industrial experiments were performed with a 300-MWe full-scale down-fired boiler. New data is reported for (i) gas temperature distributions within the primary air and coal mixture flows, (ii) gas compositions, such as O{sub 2}, CO, CO{sub 2} and NOx, and (iii) gas temperatures within the near-wall region. The data complements previously-obtained data from the same utility boiler before being modified by declination of the F-tier secondary air. By directing secondary air under the arches, the region where the primary air and pulverized coal mixture is ignited is brought forward within the boiler. Gas temperatures rose in the fuel-burning zone and fell in the fuel-burnout zone. As a result the quantity of unburned carbon in fly ash and the gas temperature at the furnace outlet were both lowered. 20 refs., 7 figs., 2 tabs.

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

Science.gov (United States)

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

2010-05-15

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

Test of small-scale central-core-cavity closure for a 300-MW(e) GCFR

International Nuclear Information System (INIS)

Robinson, G.C.; Dougan, J.R.; Naus, D.J.

1981-01-01

Under the Prestressed Concrete Reactor Vessel (PCRV) Program at the Oak Ridge National Laboratory, model tests are conducted to verify the design of the PCRV for a 300 MW(e) Gas-Cooled Fast Reactor (GCFR). Prominent features of the 1:20-scale central core cavity model included a close pitched array of fifty-five penetration tubes, forty-four segmented gusset/bearing plate assemblies, and intermeshed reinforcing steel. The closure model which was designed for a maximum cavity pressure (MCP) of 10.08 MPa was initially tested by applying 10 pressurization cycles from essentially no load to the MCP with strain and deflection data obtained during each cycle. This was followed by pressurization cycles to 32.8 MPa, 41.3 MPa, 48.3 MPa, 58.4 MPa and 79.3 MPa. At a pressure of 79.3 MPa an end cap on a penetration tube developed leaks and the test was terminated. An inelastic analysis was conducted to provide an estimate of the ultimate strength of the closure plug and to determine the potential mode of failure

Thermal hydraulic behavior of a PWR under beyond-design-basis accident conditions: Conclusions from an experimental program in a 4-loop test facility (PKL)

International Nuclear Information System (INIS)

Umminger, K.J.; Kastner, W.; Mandl, R.M.; Weber, P.

1993-01-01

Within the scope of German reactor safety research, extensive experiments covering the behavior of nuclear power plants under accident conditions have been carried out in the PKL test facility which simulates a 4-loop, 1,300 MWe KWU-designed PWR. While the investigations dealing with design-basis accidents and with the efficiency of the emergency core cooling systems have been largely completed, the main interest nowadays concentrates on the investigation of beyond-design-basis accidents to demonstrate the safety margins of nuclear power plants and to investigate the contribution of the built-in safety features for a further reduction of the residual risk. The thermal hydraulic behavior of a PWR under these extreme accident conditions was experimentally investigated within the PKL III B test program. This paper presents the fundamental findings with some of the most important results being discussed in detail. Future plans are also outlined

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.

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

Evaluation on the habitability of a reactor control room for a 1300 MWe PWR following a LOCA

International Nuclear Information System (INIS)

Chang, Si Young; Ha, Chung Woo

1988-01-01

An evaluation on the habitability of a reactor control room for a French 1300 MWe P'4 type PWR following a LOCA has been performed through exposure dose assessment for a reactor operator. A computer code COREX calculating the time-integrated exposure dose has been developed to provide a reasonable basis in this evaluation. Using COREX the exposure dose reduction factors in the reactor control room, the time--integrated radioactivities released into the atmosphere and the time-integrated exposure dose up to 30 days following the LOCA can be also calculated. From the exposure dose assessment, the time-integrated exposure dose to whole body and thyroid of a reactor operator were 0.36 mSv(0.036 rem) and 480 mSv(48.0 rem), respectively after 30 days following the LOCA. The thyroid dose of 480 mSv was nearly 10 times greater than the dose equivalent limit of 50 mSv(5.0 rem) set by the ICRP. Regarding the habitability of a reactor control room, this exceeding thyroid exposure dose could be reduced to 1.2 mSv(0.12 rem), which is 400 times less than the original, by considering the practical 4 work-shifts a day, and by improving the iodine removal efficiency of the filtration system n the reactor control room through the reinforcement of charcoal bed filters for iodine removal. The radiological habitability of a reactor control room, therefore, could be assured by comparing with the dose equivalent limit of the ICRP

Proposal for Dual Pressurized Light Water Reactor Unit Producing 2000 MWe

International Nuclear Information System (INIS)

Kang, Kyoung Min; Noh, Sang Woo; Suh, Kune Yull

2009-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the 21 st century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well

Failures of the thermal barriers of 900 MWe reactor coolant pumps

International Nuclear Information System (INIS)

Peyrouty, P.

1997-01-01

This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enables faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk represented by deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible. (author)

Failures of the thermal barriers of 900 MWe reactor coolant pumps

Energy Technology Data Exchange (ETDEWEB)

Peyrouty, P.

1996-12-01

This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible.

Failures of the thermal barriers of 900 MWe reactor coolant pumps

International Nuclear Information System (INIS)

Peyrouty, P.

1996-01-01

This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible

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

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

International Nuclear Information System (INIS)

2008-01-01

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

On-site control of 900 and 1300 MWe nuclear reactors control rod assemblies

International Nuclear Information System (INIS)

Lacroix, R.; Lebuffe, C.; Bour, D.; Pasquier, T.

1990-01-01

To measure the external wear of clads of the RCCA rodlets in both 900 and 1300 MWe P.W.R., two on site examination tools was developed by FRAGEMA. They have been used in 42 inspections between 1986 and 1989. The examination is performed in two successive phases: - longitudinal detection of wear by eddy currents, - characterization of wear by ultrasonic profilometry. Moreover, at the instance of E.D.F., an equipment is developing by INTERCONTROLE. These measurement tools allow a suitable monitoring system adapted to the phenomenon kinetics [fr

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-12-15

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

A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

International Nuclear Information System (INIS)

Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

1996-01-01

This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

Robots in P.W.R. nuclear powerplants

International Nuclear Information System (INIS)

Dubourg, M.

1987-01-01

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

Evaluation of the reliability of the protection system of 1300 MWE PWR'S

International Nuclear Information System (INIS)

Blin, A.

1990-01-01

An assesment of the reliability of the Digital Integrated Protection System (SPIN) of the 1300 MWe type french reactors has been carried out by treating an example: the emergency shutdown, which can be called upon by several initiating events. The whole chain, from sensors to breakers and control rods, is taken into account. The reliability parameters used for the quantification are evaluated essentially from the experience feedback of french reactors. The not wellknown parameters being the common cause failure rates of electronic components and the efficiency rate of the self-tests, the results of the study are then presented in a parametric form, according to these two factors

Acceptance test for 900 MWe PWR unit replacement steam generators; Essai de reception des generateurs de vapeur de remplacement des tranches REP 900

Energy Technology Data Exchange (ETDEWEB)

Gourguechon, B.

1993-12-31

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.

Evolution of Flux Mapping System (FMS) from 540 MWe to 700 MWe Indian PHWR: design perspective

International Nuclear Information System (INIS)

Sonavani, Manojkumar; Kelkar, M.G.; Singhvi, P.K.; Roy, S.; Ingle, V.J.

2013-01-01

The Flux Mapping System (FMS) of 700 MWe PHWR computes a detailed flux/power distribution of the reactor core using modal synthesis method and is also generate setback on different parameters by monitoring thermal neutron flux at more than 100 points inside the reactor core. These types of setbacks are introduced first time in Indian PHWRs. The paper brings out the Evolution of Flux Mapping System (FMS) from 540 MWe to 700 MWe and the overall design philosophy. The paper emphasizes on comparisons between 540 MWe and 700 MWe design, considerations for architectural design and setbacks for 700 MWe. (author)

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

A dual pressurized water reactor producing 2000 MWe

International Nuclear Information System (INIS)

Kang, K. M.; Suh, K. Y.

2010-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is proposed as a new design concept for large nuclear power plant. DUO is being designed to meet economic and safety challenges facing the 21. century green and sustainable energy industry. DUO2000 has two nuclear steam supply systems (NSSSs) of the Unit Nuclear Optimizer (UNO) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. UNO is anchored to the Optimized Power Reactor 1000 MWe (OPR1000). The concept of DUO can be extended to any number of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the Small and Medium sized Reactors (SMRs) be built as units, the concept of DUO2000 will apply to SMRs as well. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for Generation III+ nuclear systems. Also, the strengths of DUO2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS. Two prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The Coolant Unit Branching Apparatus (CUBA) is proposed

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Light water reactors development in Japan. (1) Introduction of LWR technology (PWR)

International Nuclear Information System (INIS)

Yamada, Ichita; Suzuki, Shigemitsu

2008-01-01

Evolutionary progress of the LWR plants in the last half-century was reviewed in series. Introduction of LWR technology (PWR) in Japan was reviewed in this article. Kansai Electric Power imported the Mihama-1 - a 340 MWe PWR built by Westinghouse Corp. It began operating in 1970 to supply power to the World Exposition (EXPO70). There followed a period in which designs was purchased from US vendors and they were constructed with the co-operation of Mitsubishi Heavy Industry, who would then receive a license to build similar plants in Japan and develop the capacity to design and construct PWRs by itself. Progress of designs, fabrications, project management and construction of PWRs were reviewed from technology transfer to its autonomy age. (T. Tanaka)

Monitoring-control of the 900 MWe and 1300 MWe nuclear reactors

International Nuclear Information System (INIS)

Meyer, J.

1982-01-01

After a short definition of the monitoring-control of the 900 MWe and 1300 MWe nuclear reactors, and a recall of requirements of nuclear energy, this paper presents the following points concerning the whole system of monitoring-control: the organization, the systems (instrumentation, automation), the technologies, the imperfections and the improvements brought to the system [fr

3-D full core calculations for the long-term behaviour of PWR's

International Nuclear Information System (INIS)

Winter, H.J.; Koebke, K.; Wagner, M.R.

1987-01-01

Presently, the most realistic simulation of a PWR core is by means of three-dimensional (3-D) full core calculations. Only by such 3-D representations can the large scope of axial effects be treated in an accurate and direct way, without the need to perform various auxiliary calculations. Although the computationally efficient burnup-corrected nodal expansion method (NEM-BC) is used, the computing effort for 3-D reactor calculations becomes rather high, e.g. a storage of about 320000 words is required to describe a 1300 MWe PWR. NEM-BC was introduced (1979) into KWU's package of PWR design codes because of its high accuracy and the great reduction of computing time and storage requirements in comparison to other methods. The application of NEM-BC to 3-dimensional PWR design is strongly correlated with the progress achieved in the solution of the homogenization and dehomogenization problem. By means of suitable methods (equivalence theory) the transport-theoretical information of the pinwise power and burnup distribution for the heterogeneous fuel assemblies is transferred in a consistent manner to the full core reactor solution. The new methods and the corresponding code system are explained in some detail. (orig.)

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)

Management of radioactive waste nuclear power plants

International Nuclear Information System (INIS)

Dlouhy, Z.; Marek, J.

1976-01-01

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

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

International Nuclear Information System (INIS)

De Baglion, L.

2011-01-01

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

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.

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)

Sodium fast reactor: an asset for a PWR UOX/MOX fleet - 5327

International Nuclear Information System (INIS)

Tiphine, M.; Coquelet-Pascal, C.; Girieud, R.; Eschbach, R.; Chabert, C.; Grosman, R.

2015-01-01

Due to its low fissile content, Pu from spent MOX fuels is sometimes regarded as not recyclable in LWR. Based on the existing French nuclear infrastructure (La Hague reprocessing plant and MELOX MOX manufacturing plant), AREVA and CEA have evaluated the conditions of Pu multi recycling in a 100% LWR fleet. As France is currently supporting a Fast Reactor prototype project, scenario studies have also been conducted to evaluate the contribution of a 600 MWe SFR in the LWR fleet. These scenario studies consider a nuclear fleet composed of 8 PWR 900 MWe, with or without the contribution of a SFR, and aim at evaluating the following points: -) the feasibility of Pu multi-recycling in PWR; -) the impact on the spent fuels storage; -) the reduction of the stored separated Pu; -) the impact on waste management and final disposal. The studies have been conducted with the COSI6 code, developed by CEA Nuclear Energy Direction since 1985, that simulates the evolution over time of a nuclear power plants fleet and of its associated fuel cycle facilities and provides material flux and isotopic compositions at each point of the scenario. To multi-recycle Pu into LWR MOX and to ensure flexibility, different reprocessing strategies were evaluated by adjusting the reprocessing order, the choice of used fuel assemblies according to their burn-up and the UOX/MOX proportions. The improvement of the Pu fissile quality and the kinetic of Pu multi-recycling in SFR depending on the initial Pu quality were also evaluated and led to a reintroduction of Pu in PWR MOX after a single irradiation in SFR, still in dilution with Pu from UOX to maintain a sufficient fissile quality. (authors)

'Kazmer' a complex noise diagnostic system for 1000 MWe PWR WWER type nuclear power units

International Nuclear Information System (INIS)

Por, G.

1992-06-01

Noise diagnostic systems have previously been developed and installed for the WWER-440 type reactors at the Paks Nuclear Power Plant, Hungary. Based on the experiences, the system has been extended and modified for use in 1000 MWe, WWER-1000 type units. KAZMER consists of three subsystem, the KARD reactor noise diagnostic system, ARGUS vibration monitoring system for rotation machinery, and ALMOS acoustic monitoring system. The installation of the KAZMER system at the Kalinin Nuclear Power Station, Russia, and the first operational experiences are outlined. (R.P.) 15 refs.; 9 figs

The main security problems encountered in the definition of the PWR 1300 MWe stage

International Nuclear Information System (INIS)

Guimbail, H.; Auvergnon, F.

1980-01-01

The main problems of security encountered in the definition of the PWR 1300 stage originated from the necessity to ensure the continuity of a stage with the preceding one, the overture to technical progress, the acceptance of responsibility for the gaines experience and the control of cost prices. A few examples show how desirable it is for a project to be managed with the design and construction rules stabilized as early as possible [fr

Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.; Generino, G.

1984-07-01

Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

The KINA neutronic module of the LEGO code for steady-state and transient PWR plant simulations

International Nuclear Information System (INIS)

Nicolopoulos, D.; Pollacchini, L.; Vimercati, G.; Spelta, S.

1989-01-01

The Automation Research Center (CRA) of ENEl has implemented some models for analyzing both incidental and operational transients in PWR power plants. For such models an axial neutron kinetics module characterized by high computational efficency with adequate results accuracy was called for. CISE has been entrusted with the task of implementing such a module named KINA and based on IQS (Improved Quasi Static) method, to be included in the library of LEGO modular code used by CRA to set up PWR power models. Moreover, The KINA module has been adapted to the neutron constants computing model developed by the EdF-SEPTEN, which has been using and improving the LEGO code for a long time in cooperation with ENEL-CRA. In this paper, after some remarks on the LEGO code, a general description of KINA neutronic module is given. The resylts of a preliminary validation activity of KINA for an EdF 1300 MWe PWR plant are also presented

Recycling option search for a 600 MWE sodium-cooled transmutation fast reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Kyo; Kim, Myung Hyun [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

2015-02-15

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro- SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to 20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

Directory of Open Access Journals (Sweden)

Yong Kyo Lee

2015-02-01

Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

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

International Nuclear Information System (INIS)

Meyer, P.J.

1981-01-01

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

The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

International Nuclear Information System (INIS)

Huber, Horacio

1989-01-01

Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author) [es

Control rod cluster drop time anomaly Guandong nuclear power station (Daya bay) and Electricite de France nuclear power stations (1450 MWe N4 Series)

International Nuclear Information System (INIS)

Olivera, J.J.; Naury, S.; Tricot, N.; Tran Dai, P.; Gama, J.M.

1996-01-01

The anomaly of control rod cluster drop time revealed at Guandong Nuclear Power Station in Daya Bay and in the Chooz B1 pilot unit for the N4 series, led to the replacement of the M1 type control rod cluster guide tubes with 1300 MWe PWR type guide tubes, adapted to the geometry of the Guandong reactors and the 1450 MWe reactors of the N4 series. The comparison of the drop times obtained with the 1300 MWe type control rod cluster guide 1300 MWe type control rod cluster guide tubes gave satisfactory results. These met the safety criterion for N4 series control rod cluster drop times (2.15 under hot shutdown conditions). The drop time tests which will be carried out in middle of and at the end of cycle 1 of Chooz B1 should make it possible to finally validate the solution already successfully implemented at Guandong. However, this anomaly has revealed the limits of representativeness of the experimental test loops with regard to the real reactor configuration. In view of this, it has been deemed necessary to ask Electricite de France to pursue its analysis both on the understanding of the phenomena which led to this anomaly and on the limits of the representativeness of the experimental test loops. (authors)

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

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

CHOOZ-A expert assessment program

International Nuclear Information System (INIS)

Bouat, M.; Godin, R.

1993-01-01

CHOOZ-A Nuclear Power Plant, the first French-Belgian PWR unit (300 MWe) was definitively shut down at the end of October 1991, after 24 years in operation. Since summer 1991, the steering committee of the French (EDF) Lifetime Project has initiated a large inquiry to the different technical specialists of EDF and external organizations, trying to define a wide expert assessment program on this plant. The aim is to improve the knowledge of aging mechanisms such as those observed on the 52 PWR French nuclear power plants (900 and 1,300 MWe), and contribute to the validation of non-destructive in-service testing methods. This paper presents the retained CHOOZ-A expert assessment program and technical lines followed during its set up. First major project stages are described, then technical choices are explained, and at last the final program is presented with the specific content of each expert assessment. The definitive program is scheduled for a three year period starting at the moment of final shutdown license acquisition, with a provisional total budget of more than US $10 million

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

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.

The future 700 MWe pressurized heavy water reactor

International Nuclear Information System (INIS)

Bhardwaj, S.A.

2006-01-01

The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

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)

Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

International Nuclear Information System (INIS)

De Rosa, Felice

2006-01-01

In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when ΔTsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its

CAREM project 15 to 150 MWe

International Nuclear Information System (INIS)

1992-05-01

The main goal of the Carem Project is the introduction of a Inherent Safe Nuclear Power Reactor in the range of low power (15 to 150 MWe). For this low-power application, light-water and low enriched uranium was selected, since using those concepts permits to take full advantage of the special characteristics of low power reactors. INVAP has been involved in the last years in the design and construction of a Carem Reactor, which could cover a range up to 150 MWe, using a multiple-unit approach. It would furnished the 150 MWe, using six Carem Reactors, of proper power, which would share most of the services. INVAP is a reliable supplier of not only the nuclear reactor but also of the fuel

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)

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

International Nuclear Information System (INIS)

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

1998-01-01

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

In-core fuel management activities in China

International Nuclear Information System (INIS)

Ruan Keqiang; Chen Renji; Hu Chuanwen

1990-01-01

The development of nuclear power in China has reached such a stage that PWR in-core fuel management becomes an urgent problem. At present the main effort is concentrated on solving the Qinshan nuclear power plant and Daya Bay nuclear power plant fuel management problems. For the Qinshan PWR (300 MWe) two packages of in-core fuel management code were developed, one with simplified nodal diffusion method and the other uses advanced Green's function nodal method. Both were used in the PWR core design. With the help of the two code packages first two cycles of the Qinshan PWR core burn-up were calculated. Besides, several research works are under way in the following areas: improvement of the nodal diffusion method and other coarse mesh method in terms of computing speed and accuracy; backward diffusion technique for fuel management application; optimization technique in the fuel loading pattern searching. As for the Daya Bay PWR plant (twin 900 MWe unit), the problem about using what kind of code package for in-core fuel management is still under discussion. In principle the above mentioned code packages are also applicable to it. Besides PWR, in-core fuel management research works are also under way for research reactors, for example, heavy water research reactor and high flux research reactor in some institutes in China. China also takes active participation in international in-core fuel management activities. (author). 19 refs

RCC-E: Design and construction rules for electrical equipment of PWR nuclear islands

International Nuclear Information System (INIS)

2016-01-01

RCC-E describes the rules for designing, building and installing electrical and I and C systems and equipment for pressurized water reactors. The code was drafted in partnership with industry, engineering firms, manufacturers, building control firms and operators, and represents a collection of best practices in accordance with IAEA requirements and IEC standards. The code's scope covers: architecture and the associated systems, materials engineering and the qualification procedure for normal and accidental environmental conditions, facility engineering and management of common cause failures (electrical and I and C) and electromagnetic interference, testing and inspecting electrical characteristics, quality assurance requirements supplementing ISO 9001 and activity monitoring. Use: RCC-E has been used to build the following power plants: France's last 12 nuclear units (1,300 MWe (8) and 1,450 MWe (4)), 2 M310 reactors in Korea (2), 44 M310 (4), CPR-1000 (28), CPR-600 (6), HPR-1000 (4) and EPR (2) reactors in service or undergoing construction in China, 1 EPR reactor in France. RCC-E is used for maintenance operations in French power plants (58 units) and Chinese M310 and CPR-1000 power plants. RCC-E has been chosen for the construction of the EPR plants in Hinkley Point, UK. Contents of the 2016 edition of the RCC-E code: Volume 1 - General requirements and quality assurance; Volume 2 - Specification of requirements; Volume 3 - I and C systems; Volume 4 - Electrical systems; Volume 5 - Materials engineering; Volume 6 - Installation of electrical and I and C systems; Volume 7 - Inspection and test methods

Electricite de France Strategy for its nuclear power plants' decommissioning programme

International Nuclear Information System (INIS)

Knockaert, J.M.; Gatineau, J.P.

1992-01-01

Although final shutdown of the first large PWR Power Stations should not occur before 2015, Electricity of France is nevertheless directly concerned by the decommissioning of its nuclear plants. The shutdown programme of the gas-graphite units is in progress and the medium-power PWR plant (300 MWe) installed at Chooz in the Ardennes will be finally shutdown at the end of 1991. This solution requires EDF to have a policy available which enables it to simultaneously run the double operation 'Plant shutdown-decommissioning' and 'New constructions-increasing available power' from both the technical and financial viewpoints. (author)

Towards a PSA harmonization French-Belgian comparison of the level 1 PSA for two similar PWR types

International Nuclear Information System (INIS)

Dupuy, P.; Corenwinder, F.; Lanore, J.M.; Gryffroy, D.; Gelder, P. de; Hulsmans, M.

2002-06-01

In the framework of the cooperation between French and Belgian regulatory authorities, a PSA (Probabilistic Safety Assessment) comparison exercise has been carried out for several years. This comparison deals with two PSA level 1 studies for internal events, performed for both power and shutdown states: the French PSA of the 900 MWe-series PWR, and the Belgian PSA of the Tihange 1 PWR, which both concern PWRs with a similar Framatome design. The purpose of this paper is to describe the PSA comparison methodology and to present, in a qualitative way, an overview of the insights obtained up to now. It also shows that such an 'a posteriori' benchmark exercise turns out to be a step towards PSA harmonization, and gives more confidence in the results of plant specific PSA when used for applications like precursor analysis or evaluations of importance to safety. (authors)

Paluel: the first of the 1300 MWe class

International Nuclear Information System (INIS)

Anon.

1980-01-01

The 1300 MWe class follows that of 900 MWe class. It is the result of studies which have taken into account the evolution of projects made by manufacturers, of research into economies of scale and site optimisation and the attempt to secure a reputation in the export field. In comparison with the 900 MWe class, the 1300 MWe class offers both similarities and differences. Similarities: the general design of the pressure vessel and the fuel elements is the same, as is the design of the loops on the primary cooling circuit. With the aim of reducing costs, the Equipment department carried out a study in 1978 regarding a number of slight modifications in the design called P'4, consisting of at least 14 units, orders for which will be given in the period up to 1983-84 [fr

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

International Nuclear Information System (INIS)

Pan Xiren; Zhang Chen.

1988-01-01

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

PWR neutron ex-vessel detection calculations using three-dimensional codes

International Nuclear Information System (INIS)

Dekens, O.; Lefebvre, J.C.; Rohart, M.; Chiron, M.

1997-01-01

During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l'Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors)

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

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)

Influence of the overfire air ratio on the NO(x) emission and combustion characteristics of a down-fired 300-MW(e) utility boiler.

Science.gov (United States)

Ren, Feng; Li, Zhengqi; Chen, Zhichao; Fan, Subo; Liu, Guangkui

2010-08-15

Down-fired boilers used to burn low-volatile coals have high NO(x) emissions. To find a way of solving this problem, an overfire air (OFA) system was introduced on a 300 MW(e) down-fired boiler. Full-scale experiments were performed on this retrofitted boiler to explore the influence of the OFA ratio (the mass flux ratio of OFA to the total combustion air) on the combustion and NO(x) emission characteristics in the furnace. Measurements were taken of gas temperature distributions along the primary air and coal mixture flows, average gas temperatures along the furnace height, concentrations of gases such as O(2), CO, and NO(x) in the near-wall region and carbon content in the fly ash. Data were compared for five different OFA ratios. The results show that as the OFA ratio increases from 12% to 35%, the NO(x) emission decreases from 1308 to 966 mg/Nm(3) (at 6% O(2) dry) and the carbon content in the fly ash increases from 6.53% to 15.86%. Considering both the environmental and economic effect, 25% was chosen as the optimized OFA ratio.

Safety design of SNR-300

International Nuclear Information System (INIS)

Traube, K.

1976-01-01

The joint German-Belgian-Dutch loop-type 300 MW(e) LMFBR prototype is being constructed at Kalkar on the lower Rhine in Germany. Among the many arguments put forward in defense of SNR-300, that of acquiring licensing exprience has proven to be of major importance to the international breeder scene. The severity of the licensing procedures and of the safety standards imposed are unique in several respects, including timing: generally growing scepticism towards nuclear power increased severity of the licensing practice; organizational features: the procedure and criteria developed for commercial light water reactors have been applied without exemptions. This relates to the commercial-type contract under which SNR-300 is being built for private utilities by a private company; and German nuclear safety standards, known worldwide to be most stringent. The following three important areas are discussed in which SNR-300 decidedly deviates from its forerunners: protection against the hypothetical core disruptive accident (HCDA), protection against external events, and provisions for in-service inspection

Three-dimensional studies of the 700 MWe steam generator design

International Nuclear Information System (INIS)

John, B.; Pietralik, J.

2006-01-01

The next stage in the Indian nuclear power programme envisions building 700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) units. This involves up-rating of all the plant equipment including the reactor, steam generators (SGs), turbo-generator, major pumps, etc. The SG used in the current generation of 540 MWe IPHWRs, is a mushroom type, inverted U-tube, natural-circulation SG. The 700 MWe SG is of the same type and has the same tube bundle design and the same heat transfer area. The tube diameter, tube pitch, and outer diameter of the SG sections are the same as for the 540 MWe SG. The geometry of the feedwater header, the flow restrictor in the downcomer and the flow distribution plate are different in the two designs. The changes were required due to a 26% increase in steam flow rate while maintaining the same circulation ratio. This paper describes the design of the 700 MWe SG and a thermalhydraulic analysis using a one-dimensional, in-house code and a three-dimensional code called THIRST developed by AECL. The codes were validated against the 540 MWe SG data. The analysis was made for the 700 MWe SG for two versions: with and without integral preheater. The results of the THIRST runs were used for a flow-induced vibration analysis. The results of the flow-induced vibration analysis show that the vibrations are not excessive. (author)

Living probabilistic safety assessment of French 1300 MWe PWR nuclear power plant unit: methodology, results and teaching

International Nuclear Information System (INIS)

Dubreuil Chambardel, A.; Villemeur, A.; Berger, J.P.; Moroni, J.M.

1991-02-01

Launched in 1986 by Electricite de France, the Probabilistic Safety Assessment of a French 1300 MWe Pressurized Water Reactor (called PSA 1300) was completed in 1989. The first objective was to assess the annual core damage frequency by identifying all the accident scenarii likely to contribute significantly to this frequency. The second objective of the study was to provide an automated computerized tool (software) for updating the assessment - in order to take new data and knowledge into account - and for performing numerous sensitivity studies easily. Its scope and characteristics render this study unique. Indeed, it required an effort amounting to 50 engineer-years. The results and the first lessons are presented in this paper. The PSA 1300 teachings will be extensively used for the design and operation of existing or future French nuclear power reactors

A study for small-medium LWR development of JAPC

International Nuclear Information System (INIS)

Okazaki, Toshihiko; Hida, Takahiko; Hoshi, Takashi; Kawahara, Hiroto; Tominaga, Kenji; Asano, Hiromitsu

2011-01-01

LWR (Light Water Reactor) power stations have accumulated many experiences of design, construction and operation. In addition, large-sized reactors have an advantage of economy of scale and 1,000 MWe LWR has therefore become the mainstream reactor in Japan. Meanwhile, introduction of the medium and small-sized LWRs (SMRs) has also been under review in Japan in order to respond to stagnant growth in electricity demand and electricity market liberalization or for investment risk mitigation; however, it has not been realized due to the economic disadvantage of scale. Therefore, JAPC has been developing the concept of SMR (300 MWe - 600 MWe) which is competitive to the large-sized LWR cooperating with Japanese plant makers (Hitachi, Toshiba Corporation and Mitsubishi Heavy Industries), assessing the possibility of realization of SMRs as one of the electric power sources in the future. As the result of the JAPC's study, we developed SMR concepts whose cost and safety are almost equal to large-sized LWR and confirmed technical feasibility of the concept in order to start developing basic design. In this paper, the outline of the SMR concepts and the current development status are presented. Concepts have been developed for two types of SMRs (i.e. BWR and PWR). As for the BWR type, reactor system is simplified by adopting natural circulation core method and CRD falling under gravity in order to downsize the reactor containments. As for the PWR type, the risk of LOCA occurrence is eliminated by unifying the primary system (e.g. incorporating steam generator into reactor). Furthermore, the primary system is simplified by adopting natural circulation core method in operation and containment vessel also become compact for the PWR. As for JAPC's further development of SMRs, key elements of SMR concepts are studied. In addition, the environment surrounding the SMRs has changed in recent years and the one with capacity exceeding 300-600 MWe class or small-sized reactor with

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

International Nuclear Information System (INIS)

Bedier, P.O.; Libmann, M.

1995-01-01

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

A nuclear desalination complex with a VK-300 boiling type reactor facility

International Nuclear Information System (INIS)

Kuznetzov, Y.N.; Mishanina, Y.A.; Romenkov, A.A.

2004-01-01

RDIPE has developed a detailed design of an enhanced safety nuclear steam supply system (NSSS) with a VK-300 boiling water reactor for combined heat and power generation. The thermal power of the reactor is 750 MW. The maximum electrical power in the condensation mode is 250 MWe. The maximum heat generation capacity of 400 Gcal/h is reached at 150 MWe. This report describes, in brief, the basic technical concepts for the VK-300 NSSS and the power unit, with an emphasis on enhanced safety and good economic performance. With relatively small power, good technical and economic performance of the VK-300 reactor that is a base for the desalination complex is attained through: reduced capital costs of the nuclear plant construction thanks to technical approaches ensuring maximum simplicity of the reactor design and the NSSS layout; a single-circuit power unit configuration (reactor-turbine) excluding expensive equipment with a lot of metal, less pipelines and valves; reduced construction costs of the basic buildings thanks to reduced construction volumes due to rational arrangement concepts; higher reliability of equipment and reduced maintenance and repair costs; longer reactor design service life of up to 60 years; selection of the best reactor and desalination equipment interface pattern. The report considers the potential application of the VK-300 reactor as a source of energy for distillation desalination units. The heat from the reactor is transferred to the desalination unit via an intermediate circuit. Comparison is made between variants of the reactor integration with desalination units of the following types: multi-stage flash (MSF technology); multi-effect distillation horizontal-tube film units of the DOU GTPA type (MED technology). The NDC capacity with the VK-300 reactor, in terms of distillate, will be more than 200,000 m 3 /day, with the simultaneous output of electric power from the turbine generator buses of around 150 MWe. The variants of the

Determination of the sources of the airborne physico-chemical 131I species in a PWR power plant

International Nuclear Information System (INIS)

Deuber, H.; Wilhelm, J.G.

1978-01-01

In a 1300 MWE PWR power plant the sources of the airborne 131 I species were determined over a period of 5 months. During power operation the main source of the radiologically decisive elemental 131 I was the exhaust from the hoods in which samples from the primary coolant are taken and processed. During refueling outage elemental 131 I was mainly contributed by the containment purge air. By efficient filtration of these exhausts, a reduction of the ingestion dose, caused by the total 131 I stack release, by a factor of nearly 4 during power operation and of possibly 10 during refueling outage can be accomplished. (author)

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

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)

Source term aspects associated with future PWR containment systems

International Nuclear Information System (INIS)

Kuczera, B.; Kebler, G.; Ehrhardt, J.; Scholtyssek, W.

1994-01-01

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

PWR passive plant heat removal assessment: Joint EPRI-CRIEPI advanced LWR studies

International Nuclear Information System (INIS)

1991-03-01

An independent assessment of the capabilities of the PWR passive plant heat removal systems was performed, covering the Passive Residual Heat Removal (PRHR) System, the Passive Safety Injection System (PSIS) and the Passive Containment Cooling System (PCCS) used in a 600 MWe passive plant (e.g., AP600). Additional effort included a review of the test programs which support the design and analysis of the systems, an assessment of the licensability of the plant with regard to heat removal adequacy, and an evaluation of the use of the passive systems with a larger plant. The major conclusions are as follows. The PRHR can remove core decay heat, prevents the pressurizer from filling with water for a loss-of-feedwater transient, and provides safety-grade means for maintaining the reactor coolant system in a safe shutdown condition for the case where the non-safety residual heat removal system becomes unavailable. The PSIS is effective in maintaining the core covered with water for loss-of-coolant accident pipe breaks to eight inches. The PCCS has sufficient heat removal capability to maintain the containment pressure within acceptable limits. The tests performed and planned are adequate to confirm the feasibility of the passive heat removal system designs and to provide a database for verification of the analytical techniques used for the plant evaluations. Each heat removal system can perform in accordance with Regulatory requirements, with the exception that the PRHR system is unable to achieve the required cold shutdown temperature of 200 F within the required 36-hour period. The passive heat removal systems to be used for the 600 MWe plant could be scaled up to a 900 MWe passive plant in a straightforward manner and only minimal, additional confirmatory testing would be required. Sections have been indexed separately for inclusion on the data base

The THESEUS project -- 50 MWe solar thermal power for Crete

Energy Technology Data Exchange (ETDEWEB)

Schillig, F.; Geyer, M.; Kistner, R.; Aringhoff, R.; Nava, P.; Brakmann, G.

1998-07-01

A consortium of European industry, utilities and research institutions from Greece, Germany, Spain and Italy attempts to implement a 52 MWe solar thermal power plant with parabolic trough technology on the Greek island of Crete sponsored by the EU' s THERMIE program. The increased demand for electricity on the island, a consequence of the growing allurement of the island as a tourist resort, makes it necessary to expand the installed capacity on Crete during the next years. According to the capacity expansion plans of Greek' s utility PPC a 160 MWe heavy fuel-fired power plant complex--two 30 MWe diesel units and two 50 MWe steam turbine units--is foreseen to be built by the year 2002. In this paper a description of the technical, economical and environmental aspects of the THESEUS project is provided. Moreover a market entry strategy for solar thermal power generation is discussed.

Severe accident mitigation strategy for the generation II PWRs in France. Some outcomes of the on-going periodic safety review of the French 1300 MWe PWR series

Energy Technology Data Exchange (ETDEWEB)

Cenerino, G.; Rahni, N.; Chevrier, P.; Dubreuil, M.; Guigueno, Y.; Raimond, E.; Bonnet, J.M. [IRSN/PSN-RES/SAG (France)

2013-07-01

The 3{sup rd} Periodic Safety Review of the French 1300 MWe PWRs series includes some modifications to increase their robustness in case of a severe accident. Their review is based on both deterministic and probabilistic approaches, keeping in mind that severe accidents frequencies and radiological consequences should be as low as reasonably practicable, severe accidents management strategies should be as safe as possible and the robustness of equipment used for severe accident management should be ensured. Consequently, the IRSN level 2 probabilistic safety assessment (L2 PSA) studies for the 1300 MWe reactors have been used to re-assess the results of the utility's L2 PSA and rank them to identify the containment failure modes contributing the most to the global risk. This ranking helped the review of plant modifications. Regarding strategies for accident management, the EDF management of water in the reactor cavity during a severe accident for the 1300 MWe PWRs is presented as well as the IRSN position on this strategy: this is an example where the optimal severe accident management strategy choice is not so easy to define. Regarding the robustness of equipment used for severe accident management, the interest of a diversification or redundancy of the French emergency filtered containment venting opening is one example among many others. (orig.)

Aerosols behavior inside a PWR during an accident

International Nuclear Information System (INIS)

Hervouet, C.

1983-01-01

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

CAREM: an innovative-integrated PWR

International Nuclear Information System (INIS)

Mazzi, R.

2005-01-01

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

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

Nuclear power. 2008 world report - evaluation

International Nuclear Information System (INIS)

Anon.

2010-01-01

In 2008, 438 nuclear power plants were available for power supply in 31 countries, 1 plant less than in 2007. No unit was commissioned for the first time, 1 nuclear power plant was decommissioned for good in 2008. At a cumulated gross power of 392,597 MWe and a cumulated net power of 372,170 MWe, respectively, worldwide nuclear generating capacity has reached a high level. Nine different reactor lines are operated in commercial plants: PWR, PWR-VVER, BWR, CANDU, D2O PWR, GCR, AGR, LWGR, and LMFBR. Light water reactors (PWR and BWR) continue to top the list with 358 plants. By the end of 2008, in 14 countries 43 nuclear power plants with an aggregate gross power of 39,211 MWe and an aggregate net power of 36,953 MWe were under construction. Of these, 37 are light water reactors, 3 are CANDU-type reactors, 2 are fast breeder and 1 D2O-PWR. 124 commercial reactors with an aggregate power in excess of 5 MWe have so far been decommissioned in 19 countries. Most of them are prototype plants of low power. About 70% of the nuclear power plants in operation, namely 304 plants, were commissioned in the eighties and nineties. The energy availability and operating availability factors of the nuclear power plants reached good levels: 80.80% for operating availability and 80,00% for energy availability. The four nuclear power plants in Finland continuecontinue to be in the lead worldwide with a cumulated average operating capacity factor of 91,60%. (orig.)

RHTF 2, a 1200 MWe high temperature reactor

International Nuclear Information System (INIS)

Brisbois, Jacques

1978-01-01

After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

Thermal circuit and supercritical steam generator of the BGR-300 nuclear power plant

International Nuclear Information System (INIS)

Afanas'ev, B.P.; Godik, I.B.; Komarov, N.F.; Kurochnkin, Yu.P.

1979-01-01

Secondary coolant circuit and a steam generator for supercritical steam parameters of the BGR-300 reactor plant are described. The BGR-300 plant with a 300 MW(e) high-temperature gas-cooled fast reactor is developed as a pilot commercial plant. It is shown that the use of a supercritical pressure steam increases the thermal efficiency of the plant and descreases thermal releases to the environment, permits to use home-made commercial turbine plants of large unit power. The proposed supercritical pressure steam generator has considerable advantages from the viewpoint of heat transfer and hydrodynamical processes

Study by the disco method of critical components of a P.W.R. normal feedwater system

International Nuclear Information System (INIS)

Duchemin, B.; Villeneuve, M.J. de; Vallette, F.; Bruna, J.G.

1983-03-01

The DISCO (Determination of Importance Sensitivity of COmponents) method objectif is to rank the components of a system in order to obtain the most important ones versus availability. This method uses the fault tree description of the system and the cut set technique. It ranks the components by ordering the importances attributed to each one. The DISCO method was applied to the study of the 900 MWe P.W.R. normal feedwater system with insufficient flow in steam generator. In order to take account of operating experience several data banks were used and the results compared. This study allowed to determine the most critical component (the turbo-pumps) and to propose and quantify modifications of the system in order to improve its availability

Towards commercial fast breeder reactors the first 1200 MWe unit

International Nuclear Information System (INIS)

Banal, M.; Carle, R.

The public probably thinks of these fast breeder reactors in terms of their rising unit capacity: RAPSODIE 20 MW (thermal), raised to 40 MW, PHENIX 25 MWe, and now 1200 MWe. However, the purposes of the project and the framework of construction have been fundamentally different in each case. Design parameters and the development program of the LMFBR are presented. (auth)

Power distribution monitoring and control in 500 MWe PHWR

International Nuclear Information System (INIS)

Kumar, A.

1996-01-01

The 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) is expected to be commissioned in a few years. It has a relatively large sized core with complex material distribution in comparison to the currently operating 220 MWe PHWRs. The resulting neutronically loosely coupled system demands continuous control of the core power distribution. This paper gives a brief description and analysis of the reactor monitoring and control system proposed for this reactor. (author). 11 refs, 8 figs, 3 tabs

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)

ANALISIS SENSITIVITAS TURBULENSI ALIRAN PADA KANAL BAHAN BAKAR PWR BERBASIS CFD

Directory of Open Access Journals (Sweden)

Endiah Puji Hastuti

2015-04-01

Full Text Available Turbulensi aliran pendingin pada proses perpindahan panas berfungsi untuk meningkatkan nilai koefisien perpindahan panas, tidak terkecuali aliran dalam kanal bahan bakar. Program CFD (CFD=computational fluid dynamics, FLUENT adalah program komputasi berbasis elemen hingga (finite element yang mampu memprediksi dan menganalisis fenomena dinamika aliran fluida secara teliti. Program perhitungan CFD dipilih dalam penelitian ini karena selain akurat juga dapat memberikan visualisasi dengan baik. Penelitian ini bertujuan untuk memahami karakteristika perpindahan panas, massa dan momentum dari dinding rod bahan bakar ke pendingin secara visual, pada medan temperatur, medan tekanan, dan medan energi kinetika pendingin, sebagai fungsi dinamika aliran di dalam kanal, pada kondisi tunak dan transien. Analisis dinamika aliran pada kanal bahan bakar PWR berbasis CFD dilakukan dengan menggunakan sampel data reaktor PWR dengan daya 1000 MWe dengan susunan bahan bakar 17x17. Untuk menguji sensitivitas persamaan aliran yang sesuai dengan model aliran turbulen pada kanal bahan bakar dilakukan pemodelan dengan menggunakan persamaan k-omega (Ƙ-ω, k-epsilon (Ƙ-ε, dan Reynold stress model (RSM. Pada analisis sensitivitas aliran turbulen di dalam kanal digunakan model mesh hexahedral dengan memilih tiga geometri sel yang masing masing berukuran 0,5 mm; 0,2 mm dan 0,15 mm. Hasil analisis menunjukkan bahwa pada analisis kondisi tunak (steady state, terdapat hasil yang mirip pada model turbulen Ƙ-ε standard dan Ƙ-ω standard. Pengujian terhadap kriteria Dittus Boelter untuk bilangan Nusselt menunjukkan bahwa model Reynold stress model (RSM direkomendasikan. Analisis sensitivitas terhadap geometri mesh antara sel yang berukuran 0,5 mm, 0,2 mm dan 0,15 mm, menunjukkan bahwa geometri sel sebesar 0,5 mm telah mencukupi. Aliran turbulen berkembang penuh telah tercapai pada model LES dan DES, meskipun hanya dalam waktu singkat (3 s, model LES memerlukan waktu komputasi

Reactor cooling systems thermal-hydraulic assessment of the ASTEC V1.3 code in support of the French IRSN PSA-2 on the 1300 MWe PWRs

International Nuclear Information System (INIS)

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

2010-01-01

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 PWRs. This PSA-2 study is 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, a wide-ranging series of comparisons 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 4 out of the 14 studied scenarios: a 12-in. cold leg Loss of Coolant Accident (LOCA), a 2-tube Steam Generator Tube Rupture (SGTR), a 12-in. Steam Line Break (SLB) and a total Loss of Feed Water scenario (LFW). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and compared to the CATAHRE 2 V2.5 results. The ASTEC results of the core degradation phase are also presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results.

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

PENENTUAN KOEFISIEN DISPERSI ATMOSFERIK UNTUK ANALISIS KECELAKAAN REAKTOR PWR DI INDONESIA

Directory of Open Access Journals (Sweden)

Pande Made Udiyani

2015-03-01

Full Text Available Atmosfer merupakan pathway penting pada perpindahan radionuklida yang lepas dari Pembangkit Listrik Tenaga Nuklir (PLTN ke lingkungan dan manusia. Penerimaan dosis pada lingkungan dan manusia dipengaruhi oleh sourceterm dan kondisi tapak PLTN. Untuk mengetahui penerimaan dosis lingkungan untuk PLTN di Indonesia, maka diperlukan nilai koefisien dispersi untuk tapak potensial yang dipilih. Model perhitungan dalam penelitian ini menggunakan model yang diterapkan pada paket program pada modul ATMOS dan CONCERN dari PC-Cosyma yaitu model perhitungan segmented plume model. Perhitungan dilakukan untuk PLTN tipe PWR kapasitas 1000 MWe berbahan bakar UO2, postulasi kejadian untuk kecelakaan DBA, kondisi tapak kasar, untuk 6 tapak contoh tapak Semenanjung Muria, Pesisir Banten, dan tapak yang didominasi oleh stabilitas cuaca C,D,E, dan F. Koefisien dispersi dihitung untuk 8 kelompok nuklida produk fisi yang lepas dari PLTN yaitu: kelompok gas mulia, lantanida, logam mulia, halogen, logam alkali, tellurium, cerium, dan kelompok stronsium & barium. Perhitungan input menggunakan paket program ORIGEN-2 dan Arc View untuk penyiapan input perhitungan. Hasil pemetaan untuk parameter dispersi maksimum rerata diperoleh pada jarak radius 800 m dari sumber lepasan untuk nuklida dari kelompok logam mulia, logam alkali dan kelompok nuklida cerium. Parameter dispersi untuk Tapak Muria maksimum 1,53E-04 s/m3, Tapak Serang adalah 1,40E-03 s/m3, tapak dengan stabilitas C: 1,72E-04 s/m3, stabilitas D: 1,40E-04 s/m3, Stabilitas E: 1,07E-04 s/m3, dan tapak dengan stabilitas F : 2,14E-05 s/m3. Kata kunci: koefisien dispersi, atmosferik, PWR, kecelakaan, Indonesia   The atmosphere is an important pathway in the migration of radionuclides transport from the Nuclear Power Plant (NPP to the environment and humans. The dose accepted in the environment and humans is influenced by the sourceterm and NPP siting condition. Distribution of radionuclides in the atmosphere is determined

Design study of a PWR of 1300 MWe of Angra-2 type operating in the thorium cycle

International Nuclear Information System (INIS)

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

1984-01-01

The utilization of the thorium-highly enriched uranium and of the thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. Reactor core design calculations were performed 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. No technical problem that precluded the utilization of any of the options analyzed was found. The savings in uranium ore introduced by the thorium cycle with fuel recycling ranges from 13% to 52% as compared with the usual uranium once-through cycle; the SWU savings goes from 13% to 22%. (Author) [pt

A 1500-MW(e) HTGR nuclear generating station

International Nuclear Information System (INIS)

Stinson, R.C.; Hornbuckle, J.D.; Wilson, W.H.

1976-01-01

A conceptual design of a 1500-MW(e) HTGR nuclear generating station is described. The design concept was developed under a three-party arrangement among General Atomic Company as nuclear steam supply system (NSSS) supplier, Bechtel Power Corporation as engineer-constructors of the balance of plant (BOP), and Southern California Edison Company as a potential utility user. A typical site in the lower Mojave Desert in southeastern California was assumed for the purpose of establishing the basic site criteria. Various alternative steam cycles, prestressed concrete reactor vessel (PCRV) and component arrangements, fuel-handling concepts, and BOP layouts were developed and investigated in a programme designed to lead to an economic plant design. The paper describes the NSSS and BOP designs, the general plant arrangement and a description of the site and its unique characteristics. The elements of the design are: the use of four steam generators that are twice the capacity of GA's steam generators for its 770-MW(e) and 1100-MW(e) units; the rearrangement of steam and feedwater piping and support within the PCRV; the elimination of the PCRV star foundation to reduce the overall height of the containment building as well as of the PCRV; a revised fuel-handling concept which permits the use of a simplified, grade-level fuel storage pool; a plant arrangement that permits a substantial reduction in the penetration structure around the containment while still minimizing the lengths of cable and piping runs; and the use of two tandem-compound turbine generators. Plant design bases are discussed, and events leading to the changes in concept from the reference 8-loop PCRV 1500-MW(e) HTGR unit are described. (author)

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

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.

IAEA’s Perspectives on Global Nuclear Power – Opportunities and Challenges

International Nuclear Information System (INIS)

Park, J.K.

2014-01-01

Status of global nuclear power: 437 reactors in operation (374.5 GWe); 2 reactors in long-term shutdown; 149 reactors in permanent shutdown; 70 reactors under construction. [As of Sep. 2014] Latest connections to the grid: - Ningde-2, 1000 MW(e), PWR, China; - Atucha-2, 692 MW(e), PHWR, Argentina; - Fuqing-1, 1000 MW(e), PWR, China). [Website: http://www.iaea.org/pris/]. IAEA projections of nuclear power: • Sep. 2014: 374.5 GWe; • 2030 - low 400.6 GWe: 7.0% increase; - high 699.2 GWe: 86.7% increase; • 2050 - low 412.9 GWe: 10.3% increase; - high 1091.7 GWe: 191.5% increase

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

International Nuclear Information System (INIS)

Bourgeois, P.; Stora, J.P.

1979-01-01

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

Nuclear power: 2006 world report - evaluation

International Nuclear Information System (INIS)

Anon.

2007-01-01

Last year, 2006, 437 nuclear power plants were available for power supply in 31 countries, 7 plants less than in 2005. One unit was commissioned for the first time, 8 nuclear power plants were decommissioned for good in 2006. At a cumulated gross power of 389,488 MWe and a cumulated net power of 370,441 MWe, respectively, worldwide nuclear generating capacity has reached a high level so far. Nine different reactor lines are operated in commercial plants: PWR, PWR-VVER, BWR, CANDU, D 2 O PWR, GCR, AGR, LWGR, and LMFBR. Light water reactors (PWR and BWR) continue to top the list with 358 plants. By the end of the year, 10 countries operated 29 nuclear power plants with an aggregate gross power of 25,367 MWe and an aggregate net power of 23,953 MWe, respectively. Of these, 21 are light water reactors, 5 are CANDU-type reactors, 2 are fast breeder and 1 a LWGR. 123 commercial reactors with an aggregate power in excess of 5 MWe have so far been decommissioned in 19 countries. Most of them are prototype plants of low power. About 70% of the nuclear power plants in operation, namely 304 plants, were commissioned in the eighties and nineties. The energy availability and operating availability factors of the nuclear power plants again reached peak levels: 82% for energy availability, and 83% for operating availability. The 4 nuclear power plants in Finland continue to be in the lead worldwide with a cumulated average operating capacity factor of 94%. (orig.)

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)

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

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)

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.)

Manufacture of the 300 MW steam generator and pressure stabilizer for Qinshan Nuclear Power Station

International Nuclear Information System (INIS)

Qian Yi; Miao Deming.

1989-01-01

A brief description of the manufacturing process of the steam generator and pressure stabilizer for 300 MWe Qinshan Nuclear Power Station in Shanghai Boiler Works is presented, with special emphasis on fabrication facilities, test procedures and technological evaluations during the manufaturing process-imcluding deep driling of tubesheets, welding of tubes to tube-sheets and tube rolling tests

1300°F 800 MWe USC CFB Boiler Design Study

Science.gov (United States)

Robertson, Archie; Goidich, Steve; Fan, Zhen

Concern about air emissions and the effect on global warming is one of the key factors for developing and implementing new advanced energy production solutions today. One state-of-the-art solution is circulating fluidized bed (CFB) combustion technology combined with a high efficiency once-through steam cycle. Due to this extremely high efficiency, the proven CFB technology offers a good solution for CO2 reduction. Its excellent fuel flexibility further reduces CO2 emissions by co-firing coal with biomass. Development work is under way to offer CFB technology up to 800MWe capacities with ultra-supercritical (USC) steam parameters. In 2009 a 460MWe once-through supercritical (OTSC) CFB boiler designed and constructed by Foster Wheeler will start up. However, scaling up the technology further to 600-800MWe with net efficiency of 45-50% is needed to meet the future requirements of utility operators. To support the move to these larger sizes, an 800MWe CFB boiler conceptual design study was conducted and is reported on herein. The use of USC conditions (˜11 00°F steam) was studied and then the changes, that would enable the unit to generate 1300°F steam, were identified. The study has shown that by using INTREX™ heat exchangers in a unique internal-external solids circulation arrangement, Foster Wheeler's CFB boiler configuration can easily accommodate 1300°F steam and will not require a major increase in heat transfer surface areas.

Influence of staged-air on airflow, combustion characteristics and NO(x) emissions of a down-fired pulverized-coal 300 MW(e) utility boiler with direct flow split burners.

Science.gov (United States)

Li, Zhengqi; Kuang, Min; Zhang, Jia; Han, Yunfeng; Zhu, Qunyi; Yang, Lianjie; Kong, Weiguang

2010-02-01

Cold airflow experiments were conducted to investigate the aerodynamic field in a small-scale furnace of a down-fired pulverized-coal 300 MW(e) utility boiler arranged with direct flow split burners enriched by cyclones. By increasing the staged-air ratio, a deflected flow field appeared in the lower furnace; larger staged-air ratios produced larger deflections. Industrial-sized experiments on a full-scale boiler were also performed at different staged-air damper openings with measurements taken of gas temperatures in the burner region and near the right-side wall, wall heat fluxes, and gas components (O(2), CO, and NO(x)) in the near-wall region. Combustion was unstable at staged-air damper openings below 30%. For openings of 30% and 40%, late ignition of the pulverized coal developed and large differences arose in gas temperatures and heat fluxes between the regions near the front and rear walls. In conjunction, carbon content in the fly ash was high and boiler efficiency was low with high NO(x) emission above 1200 mg/m(3) (at 6% O(2) dry). For fully open dampers, differences in gas temperatures and heat fluxes, carbon in fly ash and NO(x) emission decreased yielding an increase in boiler efficiency. The optimal setting is fully open staged-air dampers.

Review of some problems encountered with In-Core Fission chambers and Self-Powered Neutron Detectors in PWR's. Tests - Present use - Outlook on the near future

International Nuclear Information System (INIS)

Duchene, Jean; Verdant, Robert.

1979-01-01

The working conditions of in-core detectors are investigated as well as some reliability problems which depend on nuclear environment (such as decrease of sensibility, loss of insulation...). Then we review the long-term irradiation tests in experimental reactor that have been carried out by the CEA these last years, with fission chambers (FC) and Self-Powered Detectors (SPD). The travelling probe system with moveable FC used in the 900 MWe PWR is briefly described. Finally an outlook on future possibilities is given; for instance the use of fixed SPD and a moveable FC in the same thimble, allowing recalibration of the fixed detectors [fr

The process of the start-up of a PWR nuclear power plant in the USA

International Nuclear Information System (INIS)

Rana, B.S.

1977-01-01

The procedure is described of putting into full operation the William B. Mc Guire nuclear power plant with two PWR reactors with an output of 2x3,411 MWt (2x1,220 MWe) supplied to the Duke Power Co. lock, stock and barrel. The basic specifications are shown of units I and II and the organization of start-up activities is described. The time schedule of preoperational and start-up tests is shown and testing is reviewed preceding the first fuel charge. Also described are tests related to the first start-up of a unit comprising the period of the first fuel charge, the initial critical state, low-power tests and tests with power gradually increased. In tests of the individual systems and components of the unit, operating regulations are verified and, if required, renewed, or new regulations are introduced. (B.S.)

Risk oriented analysis of the SNR-300

International Nuclear Information System (INIS)

Koeberlein, K.

1982-01-01

The Fact Finding Committee on 'Future Nuclear Power Policy' established by the 8th German Federal Parliament in its report of June 1980 among other items published the recommendation to commission a 'risk oriented analysis' of the SNR-300 in order to enable a pragmatic comparison to be made of the safety of the German prototype fast breeder reactor and a modern light water reactor (a Biblis B PWR). The Federal Minister for Research and Technology in August 1981 officially commissioned the Gesellschaft fuer Reaktorsicherheit (GRS) to conduct the study. Following a recommendation by the Fact Finding Committee, additional studies were performed also by a group of opponents of the breeder reactor. On the instigation of the group of opponents the delivery date of the study was altered several times and finally set at April 30, 1982. GRS submitted its report by this deadline. However, a joint report by the two groups could not be compiled, as had been requested by the client, because the contributions of the opponents were not made available until mid-May 1982 and then only as an 'interim report'. In summary, the GRS study indicates that the frequency and severity of major accidents is lower for the SNR-300 at the Kalkar site than for a PWR as covered in the German Nuclear Power Plant Risk Study. (orig.) [de

Feasibility study on floating nuclear power plant (1). Conceptual design study of FNPP. Contract research

International Nuclear Information System (INIS)

Yabuuchi, Noriaki; Shimazaki, Junya; Ochiai, Masaaki; Takahashi, Masao; Nakazawa, Toshio

2001-02-01

Offshore siting methods for a nuclear power plant are classified into three types as a floating type, a settled type and a land reclamation type. The floating nuclear power plant (FNPP) has a number of advantages, such as seismic isolation, standardization of design and manufacturing, and reduction of construction period. It is, however, required for FNPP to establish the safety standards, which are different from ones for land based nuclear power plant. Investigations for this subject have not been conducted sufficiently. In this report, design study on a concept for FNPP and a review on stability evaluation for the floating platform, which were performed in order to study the safety concept of the FNPP are described. The basic concept of the FNPP are described. The basic concept for FNPP is as follows: The FNPP is sited approximately 1 - 2km off the sea coast on the open sea with water depth of about 20m and it is moored of protected sea by the breakwater, it provide a floating platform for a 1,100MWe class PWR plant. The results of design study show that the floating platform for 1,100MWe class PWR plant of 300m (L) x 80m (W) x 35m (H), and displacement of approximately 300,000 ton can be constructed in a dockyard. This floating platform guarded by the breakwater is found to be stable enough to install the nuclear power plant from the analysis simulating the movement of the platform due to sea wave or wind. (author)

Feasibility study on floating nuclear power plant (1). Conceptual design study of FNPP. Construct research

Energy Technology Data Exchange (ETDEWEB)

Yabuuchi, Noriaki; Shimazaki, Junya; Ochiai, Masaaki [Department of Nuclear Energy System, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Takahashi, Masao [Niigata Engineering Co. Ltd., Tokyo (Japan); Nakazawa, Toshio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Sato, Kazuo [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan)

2001-02-01

Offshore siting methods for a nuclear power plant are classified into three types as a floating type, a settled type and a land reclamation type. The floating nuclear power plant (FNPP) has a number of advantages, such as seismic isolation, standardization of design and manufacturing, and reduction of construction period. It is, however, required for FNPP to establish the safety standards, which are different from ones for land based nuclear power plant. Investigations for this subject have not been conducted sufficiently. In this report, design study on a concept for FNPP and a review on stability evaluation for the floating platform, which were performed in order to study the safety concept of the FNPP are described. The basic concept of the FNPP are described. The basic concept for FNPP is as follows: The FNPP is sited approximately 1 - 2km off the sea coast on the open sea with water depth of about 20m and it is moored of protected sea by the breakwater, it provide a floating platform for a 1,100MWe class PWR plant. The results of design study show that the floating platform for 1,100MWe class PWR plant of 300m (L) x 80m (W) x 35m (H), and displacement of approximately 300,000 ton can be constructed in a dockyard. This floating platform guarded by the breakwater is found to be stable enough to install the nuclear power plant from the analysis simulating the movement of the platform due to sea wave or wind. (author)

Thermoeconomic Modeling and Parametric Study of Hybrid Solid Oxide Fuel Cell â Gas Turbine â Steam Turbine Power Plants Ranging from 1.5 MWe to 10 MWe

OpenAIRE

Arsalis, Alexandros

2007-01-01

Detailed thermodynamic, kinetic, geometric, and cost models are developed, implemented, and validated for the synthesis/design and operational analysis of hybrid solid oxide fuel cell (SOFC) â gas turbine (GT) â steam turbine (ST) systems ranging in size from 1.5 MWe to 10 MWe. The fuel cell model used in this thesis is based on a tubular Siemens-Westinghouse-type SOFC, which is integrated with a gas turbine and a heat recovery steam generator (HRSG) integrated in turn with a steam turbi...

The French approach through the SMR consortium (CEA, EDF, AREVA, DCNS)

International Nuclear Information System (INIS)

Chenais, J.; Diet, A.; Grondin, Y.; Perrier, S.

2014-01-01

The SMR (Small Modular Reactor) is a new range of nuclear reactors characterized by a low output power (less than 300 MWe) and a standardized modular factory-based construction. Moderate capital costs make this type of reactor very attractive for medium-size power utilities or small countries. A consortium made around the CEA and its industrial partners (EDF, AREVA and DCNS) has performed economic and technical feasibility studies. It appears that the competitiveness of SMR require power production costs around 100 euros/MWh and that SMR will be based on PWR technology with a power output of 150 MWe per unit. 2 types of SMR are proposed: a terrestrial unit and an immersed unit. The concept of the immersed SMR, called Flexblue, is based on an unlimited access to a cold sink (the ocean) and a complete modularity. In both types the standard of safety appears to be high. (A.C.)

Progress of China's nuclear power programme

International Nuclear Information System (INIS)

Cai Jianping

1997-01-01

From a long-term point of view, nuclear power is the only solution for the shortage of energy resource. Nuclear power development strategy has been specified in China according to national condition: The electricity development of nuclear power optimizes the national energy structure and ensure the power supply, particularly in east China. China's first self-designed and self-constructed nuclear power plant--Qinshan Nuclear Power Plant (300MWe PWR) is now well under commercial operation. China is willing to cooperate with IAEA, other countries and regions in the field of nuclear energy for peaceful use on basis of mutual benefit. (author)

Analysis for the coolability of the reactor cavity in a Korean 1000 MWe PWR using MELCOR 1.8.3 computer code

International Nuclear Information System (INIS)

Lee, Byung Chul; Kim, Ju Yeul; Chung, Chang Hyun; Park, Soo Yong

1996-01-01

The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction (MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass. The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment

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.

Risk-oriented analysis of the German prototype fast breeder reactor SNR-300: off-site accident consequence model and results of the study

International Nuclear Information System (INIS)

Bayer, A.; Ehrhardt, J.

1984-01-01

Accident off-site consequence calculations and risk assessments performed for the ''risk oriented analysis'' of the German prototype fast breeder reactor SNR-300 were performed with a modified version of the off-site accident consequence model UFOMOD. The modifications mainly relate to the deposition and resuspension processes, the ingestion model, and the dose factors. Consequence calculations at the site of Kalkar on the Rhine River were performed for 115 weather sequences in 36 wind directions. They were based on seven release categories evaluated for the SNR-300 with two different fueling strategies: plutonium from Magnox reactors only and plutonium from light water reactors and Magnox reactors. In parallel, the corresponding frequencies of occurrence are determined. The following results are generated: 1. complementary cumulative frequency distribution functions for collective fatalities and collective doses 2. expected values of the collective fatalities and collective doses as well as distance-dependent expected values of individual fatality 3. contributions of the different exposure pathways to fatalities with respect to the various organs. For comparison with the risk of a PWR-1300, calculations for the PWR-1300 of the ''German Risk Study'' were repeated with the same modified consequence model. Comparison shows that smaller risks result for the SNR-300. However, the confidence interval bandwidths obtained for the frequencies of the release categories for the SNR-300 are larger than those of the PWR-1300

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)

Refitting of the 'Celimene' hot cell for following up the fuel assembly of 900 MWe PWR power reactors

International Nuclear Information System (INIS)

Lhermenier, Andre; Van Craeynest, J.-C.

1980-05-01

The 'Celimene' cell adjoining the EL3 reactor provides for the acceptance, handling and the examination of irradiated fuel assemblies from power reactors (length approximately 4m, weight approximately 700 kg). Within the framework of the PWR fuel behavior follow-up or reprocessing, it is possible to extract an assembly representative of the normal fuel cycle, carry out non destructive tests on this assembly, extract pencils from it and re-insert this assembly, after refitting the head, into the normal fuel cycle for handling in a reprocessing plant or storage pond. Given suitable refitting techniques, the re-irradiation of the assembly can be considered after examination. Significant changes have been made to the buildings and the hoist facilities for handling very heavy flasks. It was necessary to rearrange the handling, machining and in-cell storage facilities. The development of an inspection rig will make it possible, some time in 1980, to carry out non destructive tests of assemblies, optical and metrological examination of assemblies prior to dismantling or of the structure after dismantling [fr

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.)

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

International Nuclear Information System (INIS)

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

1985-08-01

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

Listing of nuclear power plant larger than 100 MWe

International Nuclear Information System (INIS)

McHugh, B.

1976-03-01

This report contains a list of all nuclear power plants larger than 100 MWe, printed out from the Argus Data Bank at Chalmers University of Technology in Sweden. The plants are listed by NSSS supply. (M.S.)

Design development of steel plate concrete modularization for the advanced PWR in Korea

International Nuclear Information System (INIS)

Mun, Taeyoup; Kim, Keunkyeong; Sun, Wonsang; Kim, Taeyoung; Hwang, Geunha

2008-01-01

APR1400 TM - an advanced PWR - has been developed in Korea since 1992. Four APR1400 units - Shin Kori no.3,4 and Shin Uljin no.1,2 - are going to be built in a next decade. As for economical efficiency, construction cost per power generation Unit(W) is improved more than 10% compared to the former 1,000 MWe PWRs. Moreover, life-cycle maintenance cost is reduced to the world's most level. For construction period from first concrete pouring to commercial operation, 54 months for APR1400 and 36 months for n-th unit have been projected. Reduction of the construction term of the Nuclear Power Plant has been emphasized increasingly for the NPP construction Project because it would reduce the interest cost and uncertainty of the project. The reduction can also advance the return of investment. Some of the PPM(Prefabrication, Preassembly, and Modularization) techniques have been studied for the shortening the construction period of nuclear power plant. Especially for the internal structure of reactor containment building (RCB) in PWR whose term of construction is critical to the whole project, Steel Plate Concrete(SC) structure has been studied as one of alternative structural systems to the conventional Reinforced Concrete(RC) structure in APR1400. SC structure is considered appropriate for the modularization of the structure with its self-supporting. In addition, formwork can be dramatically eliminated when SC structural modules are used. The MKE (Ministry of Knowledge Economy) and KHNP (Korea Hydro and Nuclear Power Co., Ltd.) initiated the research and development of SC Structure in 2005. This paper presents design examples along with Codes and Standards of SC structure in nuclear power plant. (author)

Development of essential system technologies for advanced reactor

International Nuclear Information System (INIS)

Bae, Y. Y.; Hwang, Y. D.; Cho, B. H. and others

1999-03-01

Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

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

Full Scope Modeling and Analysis on the Secondary Circuit of Chinese Large-Capacity Advanced PWR Based on RELAP5 Code

Directory of Open Access Journals (Sweden)

Dao-gang Lu

2015-01-01

Full Text Available Chinese large-capacity advanced PWR under construction in China is a new and indispensable reactor type in the developing process of NPP fields. At the same time of NPP construction, accident sequences prediction and operators training are in progress. Since there are some possible events such as feedwater pumps trip in secondary circuit may lead to severe accident in NPP, training simulators and engineering simulators of CI are necessary. And, with an increasing proportion of nuclear power in China, NPP will participate in regulating peak load in power network, which requires accuracy calculation and control of secondary circuit. In order to achieve real-time and full scope simulation in the power change transient and accident scenarios, RELAP5/MOD 3.4 code has been adopted to model the secondary circuit for its advantage of high calculation accuracy. This paper describes the model of steady state and turbine load transient from 100% to 40% of secondary circuit using RELAP5 and provides a reasonable equivalent method to solve the calculation divergence problem caused by dramatic two-phase condition change while guaranteeing the heat transfer efficiency. The validation of the parameters shows that all the errors between the calculation values and design values are reasonable and acceptable.

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)

Design, development and deployment of special sealing plug for 540 MWe PHWRs

International Nuclear Information System (INIS)

Sharma, G.; Roy, S.; Patel, R.J.

2012-01-01

The coolant channel in Pressurized Heavy Water Reactors is a pressure boundary component and is very important for reactor performance and reactor safety. Monitoring the condition of the pressure tube of each coolant channel on a periodic basis is very important. In-Service Inspection (ISI) of the coolant channels in water filled condition is done regularly for 220 MWe PHWR. For the same purpose BARC Channel Inspection System is developed for 540 MWe PHWR also. Special Sealing Plug has been developed to facilitate the channel inspection (in water filled condition) with all necessary safety features at par with normal sealing plug. Special Sealing Plug provides a 50 mm through hole for passage of drive tube of Inspection Head maintaining integrity of PHT. Lot of challenges were faced for developing the Special Sealing Plug and its associated tools. It was a first of its kind design. First ISI of TAPS-4 was conducted successfully using this plug along with associated tools in November 2011. This development has provided immense help to NPCIL in life management of 540 MWe PHWR coolant channels. (author)

Shippingport: A relevant decommissioning project

International Nuclear Information System (INIS)

Crimi, F.P.

1988-01-01

Because of Shippingport's low electrical power rating (72 MWe), there has been some misunderstanding on the relevancy of the Shippingport Station Decommissioning Project (SSDP) to a modern 1175 MWe commercial pressurized water reactor (PWR) power station. This paper provides a comparison of the major components of the reactor plant of the 72 MWe Shippingport Atomic Power Station and an 1175 MWe nuclear plant and the relevancy of the Shippingport decommissioning as a demonstration project for the nuclear industry. For the purpose of this comparison, Portland General Electric Company's 1175 MWe Trojan Nuclear Plant at Rainier, Oregon, has been used as the reference nuclear power plant. 2 refs., 2 figs., 1 tab

Secure and effective valve stem sealing in PWR power generating plants

International Nuclear Information System (INIS)

Reynolds, J.

1991-01-01

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

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

Operating experience of main steam isolation valves at Fessenheim and Bugey

International Nuclear Information System (INIS)

Dredemis, G.; Fourest, B.; Giroux, C.

1985-07-01

The paper presents the experience of Hopkinson MSIVs over about 40 reactor-years (1977 to 1984) of operation at Fessenheim and Bugey units (900 MWe PWR). The various problems encountered including ageing effects on auxiliary equipments and increases in closure time are discussed. The corrective actions undertaken by the utility and the safety assessment of these events performed by the french safety authorities are also described. This study is the synthesis of an in-depth analysis of Main Steam Isolation Valves (MSIV) and their auxiliary circuits equipping the Bugey and Fessenheim 900 MWe PWR nuclear power plants. These valves are different from those installed in the other French 900 MWe PWR reactors. The evaluation of the operation of these valves was made on the basis of incidents which occured during operation of the units or during the periodic tests, as well as anomalies discovered during maintenance operations. This analysis proved that the anomalies related to the design of the valves, as well as to their manufacture and installation, had been correctly dealt with. Furthermore, it should have also revealed potential anomalies due to ageing of the equipment

Nuclear Power Station Kalkar, 300 MWe Prototype Nuclear Power Plant with Fast Sodium Cooled Reactor (SNR-300), Plant description

International Nuclear Information System (INIS)

1984-06-01

The nuclear power station Kalkar (SNR-300) is a prototype with a sodium cooled fast reactor and a thermal power of 762 MW. The present plant description has been made available in parallel to the licensing procedure for the reactor plant and its core Mark-Ia as supplementary information for the public. The report gives a detailed description of the whole plant including the prevention measures against the impact of external and plant internal events. The radioactive materials within the reactor cooling system and the irradiation protection and surveillance measures are outlined. Finally, the operation of the plant is described with the start-up procedures, power operation, shutdown phases with decay heat removal and handling procedures

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Development of manufacturing process for production of 500 MWe calandria sheets

International Nuclear Information System (INIS)

Hariharan, R.; Ramesh, P.; Lakshminarayana, B.; Bhaskara Rao, C.V.; Pande, P.; Agarwala, G.C.

1992-01-01

Calandria tubes made of zircaloy-2 are being used as structural components in pressurised heavy water power reactors. The sheets required for producing calandria tube for 235 MWe reactors are being manufactured at Zircaloy Fabrication Plant (ZFP), NFC utilizing a 2 Hi/4 Hi rolling mill procured for the purpose, by carrying out cold rolling process to achieve the required size after hot rolling suitable extruded slabs. Due to limitation of width of the sheet that can be rolled with the mill as well as the size of the slab that can be extruded with the existing press, difficulties arose in producing acceptable full length sheets of size 6600 mm long x 435 mm wide x 1.6 mm thick for manufacturing 500 MWe calandria tube. This paper deals with the details of the process problem resolved. They are: (a)designing of suitable hot and cold rolling pass schedules, (b)selection and standardization of process parameters such as beta quenching, hot rolling and cold rolling, and (c)details of the overall manufacturing process. Due to implementation of above, sheets required for manufacturing 500 MWe calandria tube sheets were successfully rolled. About 40 nos. of acceptable full length sheets have already been manufactured. (author). 1 fig., 3 tabs

GTHTR300 cost reduction through design upgrade and cogeneration

Energy Technology Data Exchange (ETDEWEB)

Yan, Xing L., E-mail: yan.xing@jaea.go.jp; Sato, Hiroyuki; Kamiji, Yu; Imai, Yoshiyuki; Terada, Atsuhiko; Tachibana, Yukio; Kunitomi, Kazuhiko

2016-09-15

Japan Atomic Energy Agency began design and development of the Gas Turbine High Temperature Reactor of 300MWe nominal output (GTHTR300) in 2001. The reactor baseline design completed three years later was based on 850 °C core outlet temperature and a direct cycle gas turbine balance of plant. It attained 45.6% net power generation efficiency and 3.5 US¢/kW h cost of electricity. The cost was estimated 20% lower than LWR. The latest design upgrade has incorporated several major technological advances made in the past ten years to both reactor and balance of plant. As described in this paper, these advances have enabled raising the design basis reactor core outlet temperature to 950 °C and increasing power generating efficiency by nearly 5% point. Further implementation of seawater desalination cogeneration is made through employing a newly-proposed multi-stage flash process. Through efficient waste heat recovery of the reactor gas turbine power conversion cycle, a large cost credit is obtained against the conventionally produced water prices. Together, the design upgrade and the cogeneration are shown to reduce the GTHTR300 cost of electricity to under 2.7 US¢/kW h.

GTHTR300 cost reduction through design upgrade and cogeneration

International Nuclear Information System (INIS)

Yan, Xing L.; Sato, Hiroyuki; Kamiji, Yu; Imai, Yoshiyuki; Terada, Atsuhiko; Tachibana, Yukio; Kunitomi, Kazuhiko

2014-01-01

Japan Atomic Energy Agency began design and development of the Gas Turbine High Temperature Reactor of 300MWe nominal output (GTHTR300) in 2001. The reactor baseline design completed three years later was based on 850°C core outlet temperature and a direct cycle gas turbine balance of plant. It attained 45.6% net power generation efficiency and 3.5US¢/KWh cost of electricity. The cost was estimated 20% lower than LWR. The latest design upgrade has incorporated several major technological advances made in the past ten years to both reactor and balance of plant. As described in this paper, these advances have enabled raising the design basis reactor core outlet temperature to 950°C and increasing power generating efficiency by nearly 5% point. Further implementation of seawater desalination cogeneration is made through employing a newly-proposed multi-stage flash process. Through efficient waste heat recovery of the reactor gas turbine power conversion cycle, a large cost credit is obtained against the conventionally produced water prices. Together, the design upgrade and the cogeneration are shown to reduce the GTHTR300 cost of electricity to under 2.7 US¢/KWh. (author)

Model, parameter and code of environmental dispersion of gaseous effluent under normal operation from nuclear power plant with 600 MWe

International Nuclear Information System (INIS)

Hu Erbang; Gao Zhanrong

1998-06-01

The model of environmental dispersion of gaseous effluence under normal operation from a nuclear power plant with 600 MWe is established to give a mathematical expression of annual mean atmospheric dispersion factor under mixing release condition based on quality assessment of radiological environment for 30 years of Chinese nuclear industry. In calculation, the impact from calm and other following factors have been taken into account: mixing layer, dry and wet deposition, radioactive decay and buildings. The doses caused from the following exposure pathways are also given by this model: external exposure from immersion cloud and ground deposition, internal exposure due to inhalation and ingestion. The code is named as ROULEA. It contains four modules, i.e. INPUT, ANRTRI, CHIQV and DOSE for calculating 4-dimension joint frequency, annual mean atmospheric dispersion factor and doses

Experience on KKNPP VVER 1000 MWe water chemistry

International Nuclear Information System (INIS)

Ganesh, S.; Selvaraj, S.; Balasubramanian, M.R.; Selvavinayagam, P.; Pillai, Suresh Kumar

2015-01-01

Kudankulam Nuclear Power Project consists of pressurized water reactor (VVER) 2 x 1000 MWe constructed in collaboration with Russian Federation at Kudankulam in Tirunelveli District, Tamilnadu. Unit - 1 attained criticality on July 13 th 2013 and the unit was synchronized to grid on 22 nd October 2013. This paper highlights experience gained on water chemistry regime for primary and secondary circuit. (author)

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).

A world-class design for the Chinese Nuclear Program

International Nuclear Information System (INIS)

Ruiz, A.; Valero, E.

1999-01-01

Under the name Grupo Nuclear SEPI, the companies BWE, ENDESA, ENSA, ENRESA, ENUSA and INITEC, in a partnership in SEPI and with extensive experience in the nuclear sector, have in the last few years been coordinating technical and commercial actions targeted primarily at the Chinese nuclear market. These actions have been carried out under Westinghouse's leadership, for the purpose of providing a 1000 MWe nuclear Power plant design tailored to the local requirements defined by experts and authorities of the People's Republic of China (project CPWR 1000). (Author)

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)

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

Capital cost: high and low sulfur coal plants-1200 MWe. [High sulfur coal

Energy Technology Data Exchange (ETDEWEB)

1977-01-01

This Commercial Electric Power Cost Study for 1200 MWe (Nominal) high and low sulfur coal plants consists of three volumes. The high sulfur coal plant is described in Volumes I and II, while Volume III describes the low sulfur coal plant. The design basis and cost estimate for the 1232 MWe high sulfur coal plant is presented in Volume I, and the drawings, equipment list and site description are contained in Volume II. The reference design includes a lime flue gas desulfurization system. A regenerative sulfur dioxide removal system using magnesium oxide is also presented as an alternate in Section 7 Volume II. The design basis, drawings and summary cost estimate for a 1243 MWe low sulfur coal plant are presented in Volume III. This information was developed by redesigning the high sulfur coal plant for burning low sulfur sub-bituminous coal. These coal plants utilize a mechanical draft (wet) cooling tower system for condenser heat removal. Costs of alternate cooling systems are provided in Report No. 7 in this series of studies of costs of commercial electrical power plants.

Thermal stability of morpholine, AMP and sarcosine in PWR secondary systems. Laboratory and loop experiments

International Nuclear Information System (INIS)

Feron, D.; Lambert, I.

1991-01-01

Laboratory and loop tests have been carried out in order to investigate the thermal stability of three amines (morpholine, AMP and sarcosine) in PWR secondary conditions. Laboratory experiments have been performed in a titanium autoclave at 300 deg C. The results pointed out high thermal decomposition rates of AMP and sarcosine. A decomposition mechanism is proposed for the 3 amines. Loop tests have been performed in order to compare steam cycle conditioning with ammonia, morpholine and AMP. The amine concentrations and the decomposition products such as acetate and formate have been followed around the secondary circuit of the ORION loop which reproduces the main physico-chemical characteristics of a PWR secondary circuit. These concentrations are reported together with the evolution of cationic conductivities. The influence of oxygen concentration on amine thermal stability has been observed. Results are expressed also in terms of decomposition rates and of relative volatility

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)

Reactor protection systems of 500 MWe PHWRs

Energy Technology Data Exchange (ETDEWEB)

Mallik, G; Kelkar, M G; Apte, Ravindra [C and I Group, Nuclear Power Corporation, Mumbai (India)

1997-03-01

The 500 MWe PHWR has two totally independent, diverse, fast acting shutdown system called Shutdown System 1 (SDS 1) and Shutdown System 2 (SDS 2). The trip generation circuitry of SDS 1 and SDS 2 are known as Reactor Protection System 1 (RPS 1) and Reactor Protection System 2 (RPS 2) respectively. Some of the features specific to 500 MWe reactors are Core Over Power Protection System (COPPS) based upon in core Self Powered Neutron Detector (SPND) signals, use of local two out of three coincidence logic and adoption of overlap testing for RPS 2, use of Fine Impulse Testing (FIT) in RPS 2, testing of the final control elements namely electro-magnetic clutch of individual Shutoff Rods (SRs) of SDS 1 and all the fast acting valves of SDS 2, etc. The two shutdown systems have totally separate sets of sensors and associated signal processing circuitry as well as physical arrangements. A separate computerised test and monitoring unit is used for each of the two shutdown systems. Use of Programmable Digital Comparator (PDC) unit exclusively for reactor protection systems, has been adopted. The capability of PDC unit is enhanced and communication links are provided for its integration in over all system. The paper describes the design features of reactor protection systems. (author). 3 refs., 7 figs., 3 tabs.

Dynamic response of domes in CANDU 600 MWe containments

International Nuclear Information System (INIS)

Aziz, T.S.; Meng, V.; Alizadeh, A.

1981-01-01

CANDU reactors of the 600 MWe type are typically housed in a cylindrical prestressed concrete containment structure; rising from a flat slab and ending in a domed roof. The principal components of this structure are: (a) a circular base slab, (b) a vertical cylinder and (c) a spherical dome cap. A unique feature of a CANDU 600 MWe containment structure is the existence of an inner spherical concrete dome, located below the outer spherical dome, which serves as the bottom of a reservoir for the storage of 560,000 imperial gallons of douzing water. The thickness of the prestressed cylinder wall is approximately doubled between the two domes to create a ring beam. Inside the containment there exists an internal concrete structure which is independent of the containment structure except for support on the base slab. The containment boundary is a fully prestressed concrete structure. This paper deals with the seismic behaviour of the CANDU 600 MWe containment structure and the effect of its unique features; such as the lower dome and the douzing water on this behaviour. The objective of the study is to evaluate the interaction (coupling) effects between the different components of the structure. The approach taken is to study each component of the structure individually, then an assembly of the different components, and finally the total containment structure. This presentation is limited to the vertical response of the structure under a vertical earthquake only. Axisymmetric finite elements were used in all models. The vertical responses at selected points of the structure were obtained by the response spectrum method as well as the time-history method. It was observed that the response spectrum method over-estimates the vertical response of the domes and under-estimates the vertical responses of the ring girder and the containment cylinder compared to the time-history method. (orig./RW)

Operator training and requalification at GPU Nuclear

International Nuclear Information System (INIS)

Long, R.L.; Barrett, R.J.; Newton, S.L.

1982-01-01

The operator training and requalification programs at GPU Nuclear's Oyster Creek (650 MWe BWR) and Three Mile Island-1 (776 MWe PWR) nuclear plants have undergone significant revisions since the Three Mile Island-2 accident. This paper describes the Training and Education organization, the expanded training facilities, including basic principle trainers and replica simulators, and the present operator training and requalification programs

Listing of nuclear power plant larger than 100 MWe

International Nuclear Information System (INIS)

McHugh, B.

1975-06-01

This report contains a list of all nuclear power plants larger than 100 MWe, printed out from the Argus Data Bank at Chalmers University of Technology in Sweden. The plants are listed alphabetically. The report contains also a plant ranking list, where the plants are listed by the load factor (12 months) (M.S.)

Listing of nuclear power plant larger than 100 MWe

International Nuclear Information System (INIS)

McHugh, B.

1975-12-01

This report contains a list of all nuclear power plants larger than 100 MWe, printed out from the Argus Data Bank at Chalmers University of Technology in Sweden. The plants are listed by country. The report contains also a plant ranking list, where the plants are listed by the load factor (12 months). (M.S.)

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)

A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

International Nuclear Information System (INIS)

Bae, Gonghoon; Hong, Ser Gi

2015-01-01

In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO 2 –ThO 2 fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO 2 –ThO 2 fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO 2 –UO 2 fuel pins are employed to achieve long-cycle length of ∼4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods. (author)

Forbush decreases observed 40 mwe underground in 1978

International Nuclear Information System (INIS)

Benko, G.; Kecskemety, K.; Neuprandt, G.; Somogyi, A.J.

1982-01-01

Forbush decreases observed 40 mwe underground at Budapest in the first half of 1978 have been analysed together with the data of several neutron monitor stations in Europe. Assuming a power-exponential type spectrum for the variations spectrum in space as a function of rigidity, the best fitting values of power and upper cut-off rigidity have been calculated from maximum decrement by means of the weighted least squares method

ORCOST-2, PWR, BWR, HTGR, Fossil Fuel Power Plant Cost and Economics

International Nuclear Information System (INIS)

Fuller, L.C.; Myers, M.L.

1975-01-01

1 - Description of problem or function: ORCOST2 estimates the cost of electrical energy production from single-unit steam-electric power plants. Capital costs and operating and maintenance costs are calculated using base cost models which are included in the program for each of the following types of plants: PWR, BWR, HTGR, coal, oil, and gas. The user may select one of several input/output options for calculation of capital cost, operating and maintenance cost, levelized energy costs, fixed charge rate, annual cash flows, cumulative cash flows, and cumulative discounted cash flows. Options include the input of capital cost and/or fixed charge rate to override the normal calculations. Transmission and distribution costs are not included. Fuel costs must be input by the user. 2 - Method of solution: The code follows the guidelines of AEC Report NUS-531. A base capital-cost model and a base operating- and maintenance-cost model are selected and adjusted for desired size, location, date, etc. Costs are discounted to the year of first commercial operation and levelized to provide annual cost of electric power generation. 3 - Restrictions on the complexity of the problem: The capital cost models are of doubtful validity outside the 500 to 1500 MW(e) range

Changes in the design, fabrication and setting of guide tube support pins in alloy X750

International Nuclear Information System (INIS)

Benhamou, C.; Chambrin, J.L.; Todeschini, P.; Champredonde, J.; Lemaire, E.

2004-01-01

As a consequence of a problem of stress corrosion cracking (SCC) encountered on guide tube support pins (GTSP) of first generation (1982) and of second generation (1987), EDF and Framatome decided in mars 1988 to launch an important program involving a complete overhaul of the design, the material used, the fabrication and the setting in reactor of GTSP. This program has led to the implementation in 900 MWe and 1300 MWe PWR of a new tube guide support pin called NG89. This implementation began in 1989, now 15 years later, 40% of the operating GTSP in 900 MWe and 1300 MWe PWR are of NG89 type, the oldest ones cumulate 105000 hours in service without negative feedback experience. The main features of the NG89 is: - to be made from an alloy X-750 containing boron (from 25 to 45 ppm) - to have a SCC threshold set at 720 MPa - to be machined from metal bars completely treated, - to have a rolling of the fillets, and - to undergo a shot blasting on the zones of the surface the most acted upon. (A.C.)

Improvement of Candu-1000 MW(e) power cycle by moderator heat recovery

International Nuclear Information System (INIS)

Fath, H.E.S.

1988-01-01

Four different moderator heat recovery circuits are proposed for CANDU-1000 MW(e) reactors. The proposed circuits utilize all, or part, of the 155 MW(th) moderator heat load (at 70 0 C moderator outlet temperature from calandria) to the first stage of the feed water heating system. An economics study was carried out and indicated that the direct circulation of feed water through the moderator heat exchanger (with full heat recovery) is the most economical scheme. For this scheme the saved steam from the turbine extraction was found to produce additional electric power of 8 MW(e). This additional power represents a 0.7% increase in the plants nominal electric output. The outstanding features and advantages of the selected scheme are also presented. (author)

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.).

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)

Safety margin improvement by adopting the feature of interleaving in 700 MWe PHWR

International Nuclear Information System (INIS)

Kumar, Nrependra; Yadav, S.K.; Khan, T.A.; Dixit, A.; Singhal, Mukesh; Nair, Suma R.

2015-01-01

Indian Pressurised Heavy Water Reactors (IPHWRs) of 700 MWe are under construction at Kakrapar Atomic Power Project -3,4 and Rajasthan Atomic Power Project-7,8. These units have enhanced safety features with respect to standard IPHWRs. One of the enhanced features is interleaving of feeders/channels. In interleaved feeder configuration, each header located at either end of reactor gets connected to one quarter of core channels, which are uniformly distributed. The core is divided into two loops with feeder connected in interleaved fashioned. In this paper a comparative study has been performed between the two cases: 1) The core splits in two vertical halves and each vertical half is a loop of PHT (TAPS-3 and 4 Type configuration). 2) The core is divided into two loops with feeders/ channels connected in interleaved fashioned (700 MWe Configuration). LOCA studies have been performed for 700 MWe PHWR considering interleaving of feeders configuration using in-house developed computer code ATMIKA and 3-D neutron kinetics code IQS-3D. The issue of interleaving is closely linked to an inherent reactivity characteristic of PHWR reactors (viz., positive void reactivity coefficient) which leads to a power increase following a Large LOCA. In 700 MWe PHWR with intent to improve the safety margin, adopted the feature of interleaving of feeders which causes in reduction in the magnitude of void coefficient and results in reduction of peak power during LBLOCA. The systematic LBLOCA study demonstrates that interleaved configuration of feeder/channels of two loops has higher safety margins (i.e. with respect to peak power, prompt-criticality margin, adiabatic heat deposition on the fuel pins, sheath temperature excursion and clad oxidation) with regard to the effectiveness of shutdown system. (author)

Supplementary shutdown system of 220 MWe standard PHWR in India

International Nuclear Information System (INIS)

Muktibodh, U.C.

1997-01-01

The design objective of the shutdown system is to make the reactor subcritical and hold it in that state for an extended period of time. This objective must be realised under all anticipated operational occurrences and postulated abnormal conditions even during most reactive state of the core. PHWR design criteria for shutdown stipulates requirement of two independent diverse and fast acting shutdown systems, either of which acting alone should meet the above objectives. This requirement would normally call for a large number of reactivity mechanism penetrations into the calandria. From the point of view of space availability at the reactivity mechanism area on top of calandria, for the relatively small core of 220 MWe PHWRs, and ease of maintenance realisation of the total worth by either of the shutdown systems acting alone was difficult. To overcome this engineering constraint and at the same time to satisfy the design criteria, a unique approach to meet the reactivity demands for shutdown was adopted. The reactivity requirements of the shutdown consists of fast and slow reactivity changes. For the shutdown system of 220 MWe PHWRs, the approach of realizing fast reactivity changes with dual redundant, diverse, fast acting shutdown systems aided by a slow acting shutdown system to counter delayed reactivity changes was conceived. The supplementary slow acting shutdown system is called upon to act after actuation of either of the two redundant fast acting systems and is referred to as Liquid Poison Injection System (LPIS). The system adds bulk amount of neutron poison (boric acid), equivalent to 45 mk, directly into the moderator through two nozzles in calandria using pneumatic pressure. This paper describes the design of LPIS as envisaged for the standardised 220 MWe PHWRs. (author)

Utilization of El Dabaa basic simulator for manpower development

International Nuclear Information System (INIS)

Wahab, W.A.; Abdel Hamid, S.B.

1998-01-01

Training with basic simulators is considered as an essential tool for training and retraining of the nuclear power plant staff. To achieve that objective; Nuclear Power Plants Authority (NPPA) has installed a basic training simulator for PWR and PHWR simulation at El Dabaa site. The basic simulator simulates a 3-loops PWR-900 MW(e) and 2-loops PHWR-600 MW(e). The simulator has passed in-plant acceptance tests in CEA/CENG, Grenoble, France and passed successfully on-site completion tests at El Dabaa site in November 1991. This paper presents the main features of the simulator; training capabilities, hardware configuration and software architectures. Also training methodology by NPPA and the training experience gained by using the simulator are presented. (author)

Capital cost: pressurized water reactor plant. Commercial electric power cost studies

Energy Technology Data Exchange (ETDEWEB)

1977-06-01

The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

Capital cost: pressurized water reactor plant. Commercial electric power cost studies

International Nuclear Information System (INIS)

1977-06-01

The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description

Role of Fugen HWR in Japan and design of a 600 MWe demonstration reactor

International Nuclear Information System (INIS)

Sawai, Sadamu.

1982-03-01

Fugen, a 165 MWe prototype of a heavy water-moderated, boiling light water-cooled reactor, has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work for a 600 MWe demonstration plant has been carried out since 1973. The important systems and components, such as pressure tube assemblies and control rod drive mechanism, are essentially the same as those of Fugen. However, some modification is made owing to the experience obtained in Fugen and LWrs. In the HWR Fugen, plutonium and uranium are effectively used, and plutonium makes the coolant void reactivity more negative, which results in the increase of the stability and safety of the reactor. On August 4, 1981, the ad hoc committee submitted the final report to the Japanese Atomic Energy Commission, in which the construction of a 600 MWe demonstration plant was recommended. As for the research and development on reactor safety, coolant leak detectors, the performance of ECCS, and safety design codes are enumerated. Since 1965, mixed oxide fuel has been developed, and 168 fuel assemblies were loaded in Fugen, but failure did not occur. (Kako, I.)

NPI one year on [Framatome-Siemens common product

International Nuclear Information System (INIS)

Ruess, F.; Lebreton, G.

1990-01-01

Nuclear Power International (NPI), a joint subsidiary of Framatome and Siemens was established in April 1989 to coordinate the development and marketing of a common PWR technology for the world market. Both parent companies command well proven PWR technology and can rely on experience gained with over 100 000MWe of nuclear capacity in operation, under construction or on order. The technology from each is being offered through NPI until the joint PWR technology which is under development, becomes available. NPI's experience to date is reviewed. (author)

A multinode digital control system for 500 MWe PHWR

Energy Technology Data Exchange (ETDEWEB)

Patil, G N; Suresh Babu, R M; Jangra, L R; Das, Shantanu; Mallik, S B [Bhabha Atomic Research Centre, Bombay (India). Reactor Control Div.

1994-12-31

A fault tolerant distributed digital computer system for 500 MWe reactor power regulation is configured around standard microcomputer boards designed indigenously. The system is configured as functionally partitioned distributed control system having 8 nodes linked by high-speed dual redundant high-way. The paper gives the details of the configuration of system and how the features of fault-tolerance and fail-safeness are achieved through design. (author). 1 fig.

Conceptual core designs for a 1200 MWe sodium cooled fast reactor

International Nuclear Information System (INIS)

Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

2008-01-01

The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

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

ALIBABA, an assistance system for the detection of confinement leaks in a PWR reactor; ALIBABA, un systeme d`aide a la detection des voies de fuites du confinement sur un reacteur a eau sous pression

Energy Technology Data Exchange (ETDEWEB)

Bedier, P.O.; Libmann, M. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie

1995-12-31

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

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)

A decade of nuclear development in France

International Nuclear Information System (INIS)

Renon, G.

1984-01-01

After the first energy crisis, in France depended heavily on oil import, the broad lines of strategy were the improvement of efficiency in energy use, the diversification of energy sources and the development of domestic substitute energy sources. The energy conservation policy achieved in 1983 up to 35 million tons of oil equivalent saving. The coal and natural gas output in France decreased during the last ten years, but the national energy production covered 38.7 % of the national needs in 1983. This result is mainly due to the implementation of electro-nuclear program whose contribution to the total energy consumption was more than 17 % in 1983. In 10 reference years, 29 PWR units of 900 MWe and 17 PWR units of 1300 MWe have been launched, and 27 units of 900 MWe each have been commissioned. The power cost per kWh generated by nuclear energy remained about 2/3 of that by coal. The cumulative production of 900 MWe PWRs is already equivalent to 50 reactor-years at full power. The average availability of the units was 68 % in 1983, which showed that the problems have been solved. Also the state of nuclear fuel cycle in France, the trend of reactor improvement and the long term plan are explained. (Kako, I.)

Role of Fugen-HWR in Japan and design of a 600 MWe demonstration reactor

International Nuclear Information System (INIS)

Sawai, S.

1982-01-01

Fugen, a 165 MWe prototype of a heavy water moderated boiling light water cooled reactor; has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work of the 600 MWe demonstration plant has been carried out since 1973. Important system and components, such as pressure tube assemblies, control rod drive mechanism, etc., are essentially the same as those of Fugen. Some modifications, however, are made especially from the stand point of experiences In the Fugen-HWR, plutonium and uranium would be effectively used; and plutonium could make the coolant void reactivity more negative which would give good results in increasing the reactor stability and safety. On the other hand, nuclear power plants are mainly consisted of LWRs in Japan. Considering the above situations, the Fugen-HWR, coupled with LWRs, is now considered in Japan to contribute to our energy security by using plutonium and depleted uranium extracted from spent fuels of LWRs: thereby reducing the demands On August 4, 1981, the ad hoc committee on the 600 MWe demonstration Fugen-HWR submitted the final report to the Japan AEC, after having had discussions and evaluations. In the report, the ad hoc committee recommended to build the 600 MWE demonstration plant with appropriate supports of the Government. The Japan AEC will be expected to make her decision on the program in the near future. As for the reactor safety R and C, development has been stressed on coolant leak detectors and ECCS performances or Since 1965, many development works have been done for mixed oxide fuel assemblies, both for establishment of the fabrication technology and for clarification of irradiation performances. 196 mixed oxide fuel assemblies have been manufactured for Fugen. 168 of them were loaded and 92 were withdrawn. No fuel has been failured yet. (author)

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

Flux mapping algorithm (FMA) for 700 MWe PHWR

International Nuclear Information System (INIS)

Sonavani, Manoj; Ingle, V.J.; Singhvi, P.K.; Raj, Manish; Fernando, M.P.S.; Kumar, A.N.

2012-01-01

For large reactor like 700 MWe PHWR effective spatial control is essential and is provided by RRS. For spatial control purpose reactor core is divided into 14 power zones. Corresponding to each zone is a light water zonal compartment. The 14 ZCCs are located in two radial planes, each containing 7 ZCCs. For each zone, power measurement is carried out using inconel (3 pitch long) self powered neutron detector (SPND) at appropriate location close to the respective ZCC. Since the zone power as obtained by the healthy zone control detector (ZCD) reading belonging to a particular zone may not correspond to its actual power because the detector per zone, measure only average fluxes but the zone extends over a large core region. Therefore accurate estimation of zone power calibration factors is required to estimate the zone powers and also to provide effective spatial power control to avoid the xenon induced spatial power oscillations in large PHWRs like 700 and 540 MWe Reactors. This accurate calculation of zone power is carried out by FMS which uses λ modes in its algorithm. Flux at any point inside the reactor can be represented in terms of the linear combination of these modes. Coefficients used in the expansion are called combining coefficient. If the readings of the detectors are known, then combining coefficients can be estimated by simple matrix operations. Once these combining coefficients are known, flux at any point inside the reactor can be found. (author)

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.

Water treatment for 500 MWe PHWR plants

International Nuclear Information System (INIS)

Vasist, Sudheer; Sharma, M.C.; Agarwal, N.K.

1995-01-01

Large quantities of treated water is required for power generation. For a typical 500 MWe PHWR inland station with cooling towers, raw water at the rate of 6000 m 3 /hr is required. Impurities in cooling water give rise to the problems of corrosion, scaling, microbiological contamination, fouling, silical deposition etc. These problems lead to increased maintenance cost, reduced heat transfer efficiency, and possible production cut backs or shutdowns. The problems in coastal based power plants are more serious because of the highly corrosive nature of sea water used for cooling. An overview of the cooling water systems and water treatment method is enumerated. (author). 2 refs., 1 fig

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

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

International Nuclear Information System (INIS)

Ireland, J.R.

1982-01-01

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

Two metals welded joints analysis. Specific problems and solution proposal

International Nuclear Information System (INIS)

Bodson, F.; Launay, J.P.; Thomas, A.

1983-03-01

This paper summarizes the non destructive quality control of bimetallic welded joints on pipes and metallic structures of PWR type reactors (1300 MWe): radiographic and metrasonic failure detection, standardization and in service control processes [fr

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.)

A reactivity hold-down strategy for soluble boron free operation by introducing Pu-238 added fuel

International Nuclear Information System (INIS)

Kim, Soon Young; Kim, Jong Kyung

2000-01-01

A new concept of Pu-238 added fuel is introduced to control the reactivity and power distribution in soluble boron free (SBF) pressurized water reactor (PWR) core. Though extensive use of burnable poison and control rods is inevitable for reactivity suppression in SBF core, it causes the core power distribution control to be so difficult that a practical SBF operation is far distant. In this work, it is confirmed that the excess reactivity can be greatly suppressed by introducing the Pu-238 added fuel. As a result of the conceptual core design of the 600 MWe SBF PWR using Pu-238 added fuel, the core reactivity is well controlled in comparison with the results obtained from the earlier 600 MWe SBF core design works. Especially, the axial power shape control is performed successfully with the aid of simple axial zoning scheme, developed in this study, by using Pu-238 enrichment zoning. The Pu-238 added fuel is also tested for 1300 MWe SBF PWR core design, in which the power distribution control can be more difficult than that of smaller plants if soluble boron control is not available. The results show that the core excess reactivity and the power distribution can be well controlled without using soluble boron even in a large-sized PWR. Hence, one of the difficult control problems arising in SBF core design can be greatly mitigated by introducing the new fuel concept. It is further expected that the Pu-238 added fuel, the simple axial zoning scheme, and the control bank operation strategy introduced in this study are directly applicable to practical SBF core design

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)

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)

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)

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.)

Safety options for the 1300 MWe program

International Nuclear Information System (INIS)

Cayol, A.; Dupuis, M.C.; Fourest, B.; Oury, J.M.

1980-04-01

Standardization of the nuclear plants built in France implies an examination of the main technical safety options to be taken for a given type of reactor. By this procedure the subjects for which detailed studies will be needed to confirm the decisions made for the project can be defined in advance. In this context the technical safety option analysis for the 1300 MWe plants was conducted from the end of 1975 to the middle of 1978 according to usual regulation examination practice. The main conclusions are presented on the following subjects: safety methods; technical options concerning the containment vessel, primary fluid activity, fuel elements, steam generators; general organization of the lay-out [fr

Commerical electric power cost studies. Capital cost addendum multi-unit coal and nuclear stations

International Nuclear Information System (INIS)

1977-09-01

This report is the culmination of a study performed to develop designs and associated capital cost estimates for multi-unit nuclear and coal commercial electric power stations, and to determine the distribution of these costs among the individual units. This report addresses six different types of 2400 MWe (nominal) multi-unit stations as follows: Two Unit PWR Station-1139 MWe Each, Two Unit BWR Station-1190 MWe Each, Two Unit High Sulfur Coal-Fired Station-1232 MWe Each, Two Unit Low Sulfur Coal-Fired Station-1243 MWe Each, Three Unit High Sulfur Coal-Fired Station-794 MWe Each, Three Unit Low Sulfur Coal-Fired Station-801 MWe Each. Recent capital cost studies performed for ERDA/NRC of single unit nuclear and coal stations are used as the basis for developing the designs and costs of the multi-unit stations. This report includes the major study groundrules, a summary of single and multi-unit stations total base cost estimates, details of cost estimates at the three digit account level and plot plan drawings for each multi-unit station identified

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.)

Seismic isolation of nuclear power plants - EDF's philosophy

International Nuclear Information System (INIS)

Coladant, C.

1989-01-01

The elastomer bearing pads used since 1963 as supports for prestressed concrete pressure vessels (PCPVs) was quickly chosen by Electricite de France (ED) to improve the capability of nuclear power plants (NPPs) to withstand strong earthquakes and to reduce the seismic loads on structures and equipment. The standardized units for 900 and 1,300 MW(e) pressurized water reactor (PWR) plants have moderate seismic design loads of 0.2 and 0.15 g, respectively. These design loads were exceeded by the site dependent spectra of Cruas (France) and Koeberg (South Africa). To keep the plant design unchanged and to take the advantages of standardization, these units were put on laminated bearings with or without sliding plates. For the future French 1,500 MW(e) fast breeder reactors (FBRs), which are more sensitive to seismic loads, the base isolation is considered by EDF at the beginning of the design, even for low ground motions of 0.1 g. The buildings are placed on laminated bearings while the reactor block is supported by springs and dampers. The isolated plant has identical costs as a conventional design such as SPX1 at Creys-Malville

Computer code for thermal-hydraulic simulation of heat pressurizer tanks operation (Simterm-H)

International Nuclear Information System (INIS)

Sellos, R.F.

1987-01-01

It is presented the Simtherm-H computer code, developed for calculating the thermodynamic properties of the high pressure heating system and the feedwater tank in transient state for PWR nuclear power plants (1300 MWe). (E.G.) [pt

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

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.

Investigations on construction material and construction concepts in order to obtain dose-reducing effects in the dismantling of the biological shield of a 1300 MWe-PWR

International Nuclear Information System (INIS)

Bittner, A.; Jungwirth, D.; Knell, M.; Schnitzler, L.

1984-04-01

Numerical values of neutron fluxes, activations, dose rates etc. as a function of characteristic values of materials required for optimization purposes to reduce the radiation effect of the biological shield of a PWR are not available. Design concepts are presented for biological shields of PWRs made of concrete with respect to both the most suitable application of materials and the design principles aiming at reduced radiation exposure as compared to present designs during entering, waste disposal and ultimate storage. To evaluate the present-state design the above values have been calculated. Suggested alternative designs are biological shields with selective material application, built from precast elements with or without boron carbide layer arranged in front of it. (orig./HP) [de

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.)

Studies on flow induced vibration of reactivity devices of 700 MWe Indian PHWR

Energy Technology Data Exchange (ETDEWEB)

Prabhakaran, K.M., E-mail: kmprabha@yahoo.com [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Goyal, P.; Dutta, Anu; Bhasin, V.; Vaze, K.K.; Ghosh, A.K. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Pillai, Ajith V.; Mathew, Jimmy [Nuclear Power Corporation of India Ltd., Mumbai 400 094 (India)

2012-03-15

Highlights: Black-Right-Pointing-Pointer FIV studies on internals of heavy water filled calandria of 700 MWe Indian PHWR is presented. Black-Right-Pointing-Pointer This includes CFD and structural dynamic analysis to predict the dynamic behavior of component lying inside calandria. Black-Right-Pointing-Pointer Results of these calculations as well as conclusions from this investigation are presented. Black-Right-Pointing-Pointer It is established that FIV is not a concern in the present design of calandria internals. - Abstract: Component failures due to excessive flow-induced vibration are still affecting the performance and reliability of nuclear power stations. Tube failures due to fretting-wear in nuclear steam generators, and vibration related damage of reactor internals are of particular concern. In the Indian nuclear industry, flow induced vibrations are assessed early in the design process and the results are incorporated in the design procedures. In this paper the details of flow induced vibration studies on internals like liquid zone control unit and poison injection units of heavy water filled calandria of 700 MWe Indian pressurized heavy water reactor is given. This includes computational fluid dynamics studies from which the velocities are extracted for the components lying inside the calandria. With these velocities as input, further studies are performed to predict the dynamic behavior of these components. Results of these calculations as well as conclusions derived from this investigation are presented. Based on the studies it has been established that flow induced vibration is not a concern in the present design of 700 MWe calandria internals.

Capital cost: pressurized water reactor plant. Commerical electric power cost studies

International Nuclear Information System (INIS)

1977-06-01

The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate

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)

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

Possibilities of optimizing non-nuclear simulation of pressurized water reactor transients

International Nuclear Information System (INIS)

Silva Filho, E.

1985-01-01

The GKSS-Forschungszentrum Geesthacht GmbH has instituted the concept of a scaled test facility (volume scale factor of 1/100) of a typical PWR of the 1 300 MWe class for the purpose of studying small breaks Loss-of-Coolant Accidents (LOCA) and transients. Having in mind the goal of an optimization of this concept has been choosen a station blackout with and without reactor shutdown and a small break LOCA in a primary loop piping to investigate the thermohydraulic behaviour of the test facility in comparison to the reactor plant. The computer code RELAP 5/MOD 1 has been utilized to compare the test facility behaviour with the reactor plant one. Recommendations are given for minimization of distortions between test facility and reactor plant. (orig./HP) [de

Qinshan NPP large break LOCA safety analysis

International Nuclear Information System (INIS)

Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

1997-01-01

Qinshan NPP is the first nuclear power plant in the mainland of China, a 300 MW(e) two-loop PWR. Large break LOCA is the design-basis accident of Qinshan NPP. Based on available computer codes, the own analysis method which complies with Appendix k of 10 CFR 50 has been established. The RELAP4/MOD7 code is employed for the calculations of blowdown, refill and reflood phase of the RCS respectively. The CONTEMPT-LT/028 code is used for the containment pressure and temperature analysis. The temperature transient in the hot rod is calculated using the FRAP-6T code. Conservative initial and functional assumptions were adopted during Qinshan NPP large break LOCA analysis. The results of the analysis show the applicable acceptance criteria for the loss-of-coolant accident are met

The status of safeguarding 600 MW(e) CANDU reactors

International Nuclear Information System (INIS)

Von Baeckmann, A.; Rundquist, D.E.; Pushkarjov, V.; Smith, R.M.; Zarecki, C.W.

1982-09-01

There has been extensive work in the development of CANDU safeguards since the last International Conference on Nuclear Power, and this has resulted in the development of improved equipment for the safeguards system now being installed in the 600 MW(e) CANDU generating stations. The overall system is designed to improve on the existing IAEA safeguards and to provide adequate coverage for each plausible nuclear material diversion route. There is sufficient sensitivity and redundancy to enable the timely detection of the possible diversion of significant quantities of nuclear material

Internal Technical Report, Safety Analysis Report 5 MW(e) Raft River Pilot Plant

Energy Technology Data Exchange (ETDEWEB)

Brown, E.S.; Homer, G.B.; Spencer, S.G.; Shaber, C.R.

1980-05-30

The Raft River Geothermal Site is located in Southern Idaho's Raft River Valley, southwest of Malta, Idaho, in Cassia County. EG and G idaho, Inc., is the DOE's prime contractor for development of the Raft River geothermal field. Contract work has been progressing for several years towards creating a fully integrated utilization of geothermal water. Developmental progress has resulted in the drilling of seven major DOE wells. Four are producing geothermal water from reservoir temperatures measured to approximately 149 C (approximately 300 F). Closed-in well head pressures range from 69 to 102 kPa (100 to 175 psi). Two wells are scheduled for geothermal cold 60 C (140 F) water reinjection. The prime development effort is for a power plant designed to generate electricity using the heat from the geothermal hot water. The plant is designated as the ''5 MW(e) Raft River Research and Development Plant'' project. General site management assigned to EG and G has resulted in planning and development of many parts of the 5 MW program. Support and development activities have included: (1) engineering design, procurement, and construction support; (2) fluid supply and injection facilities, their study, and control; (3) development and installation of transfer piping systems for geothermal water collection and disposal by injection; and (4) heat exchanger fouling tests.

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

The development of the Chinese nuclear power program

International Nuclear Information System (INIS)

Tournyol du Clos, A.

2009-01-01

The electricity production is steadily growing in China and the intense use of coal has led to various problems such as -) important contamination of urban areas, -) safety concerns in coal mines, and -) coal transport as the production regions are not the consumption regions. In 2004, Chinese authorities decided to accelerate the recourse to nuclear power. The main measures that were adopted followed: -) the development of the CPR1000 that is a Chinese improved version of the Areva 900 MW PWR; -) the development of a Chinese third generation reactor based on foreign design: 4 AP1000 and 2 EPR have been ordered; -) the adoption of a closed cycle for the fuel; -) the participation to international collaborations like GIF (Generation 4 International Forum) and to ITER. Today the number of operating power reactors in China is 11 and 26 have been ordered. The objective is to reach a share of 20% for nuclear power in the electricity production which means 250 GW installed by 2050. It appears that the nuclear renaissance has begun in Asia and real opportunities for the European nuclear industry are there. (A.C.)

Performances and reliabilities comparisons by multidimensional statistic studies

International Nuclear Information System (INIS)

Coudray, R.

1993-01-01

After an overview of the methods utilizables in this type of analyse, we insist on the method used in the experience return for the French PWR 900 MWe chemical and volume control system pump. 7 figs., 3 tabs., 8 refs

GTM (Guangdong ten year modifications): the Chinese experience in plant safety alignment

International Nuclear Information System (INIS)

Louvat, J.-P.; Chauvin, T.

2005-01-01

The present article introduces the recent experience by Framatome-ANP of developing a large scale engineering project in China. The so-called GTM project deals with the studies and the site implementation of a dozen of major design improvements at the Guangdong Nuclear Power Station on Daya Bay site, stemming from the Periodic Safety Reassessment (PSR) carried-out by the Utility. This article focuses on the benefits of sharing with the Chinese utility the long experience of the study of the modification batches 'lot93' and 'VD2' implemented on the French CPY 900 MWe series units and on the confidence gained throughout years of continuous sinofrench technical cooperation that was established by the plant owner in Framatome- ANP's capabilities to carry-out the successful development of the project. Indeed Framatome-ANP involved its Chinese partners Beijing Institute of Nuclear Engineering, China Nuclear Industry 23rd Company and Shenzhen Nuclear Engineering Company to localize design and erection activities. The successful achievement of the GTM studies proved the maturity of the Chinese engineering competencies to set-up a sustainable industrial scheme devoted to the further development of nuclear engineering services in China. (authors)

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

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)

The costs of nuclear power in the Netherlands

International Nuclear Information System (INIS)

1978-01-01

A study on the costs of nuclear power generation in the Netherlands is presented. Light water cooled reactors are chosen as nuclear power plants and no difference is made in calculating the costs between a PWR type reactor and a BWR type reactor. The power plants have an output of 1000 MWe. From each part of the whole fuel cycle the costs are determined, taking into account interest, investments, time of construction, labor costs, insurances etc. Also are determined from each part of the fuel cycle the energy costs; the costs per kWh. Finally a comparison is made in costs between a 1000 MWe power plant and a 600 MWe power plant

Comparative of fuel cycle cost for light water nuclear power plants; Uporedna analiza cene gorivnog ciklusa lakovodne nuklearne elektrane

Energy Technology Data Exchange (ETDEWEB)

Kocic, A; Dimitrijevic, Z [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1978-07-01

Starting from ost general fuel cycle scheme for light water reactors this article deals with conceptual differences of BWR, PWR and WWER as well as with the influence of certain phases of fuel cycle on economic parameters of an equivalent 1000 MWe reactor using a computer program CENA /1/ and typical parameters of each reactor type. An analysis of two particular power plants 628 MWe and 440 MWe WWER by means of the same program is given in the second part of this paper taking into account the differences of in-core fuel management. This second approach is especially interesting for the economy of the power plant itself in the period of planning. (author)

Comparative of fuel cycle cost for light water nuclear power plants

International Nuclear Information System (INIS)

Kocic, A.; Dimitrijevic, Z.

1978-01-01

Starting from ost general fuel cycle scheme for light water reactors this article deals with conceptual differences of BWR, PWR and WWER as well as with the influence of certain phases of fuel cycle on economic parameters of an equivalent 1000 MWe reactor using a computer program CENA /1/ and typical parameters of each reactor type. An analysis of two particular power plants 628 MWe and 440 MWe WWER by means of the same program is given in the second part of this paper taking into account the differences of in-core fuel management. This second approach is especially interesting for the economy of the power plant itself in the period of planning. (author)

Damage evaluation of 500 MWe Indian pressurized heavy water reactor nuclear containment for air craft impact

International Nuclear Information System (INIS)

Kukreja, Mukesh; Singh, R.K; Vaze, K.K; Kushwaha, H.S.

2003-01-01

Non-linear transient dynamic analysis of 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) nuclear containment has been carried out for the impact of Boeing and Airbus category of aircraft operated in India. The impulsive load time history is generated based on the momentum transfer of the crushable aircraft (soft missiles) of Boeing and Airbus families on the containment structure. The case studies include the analyses of outer containment wall (OCW) single model and the combined model with outer and inner containment wall (ICW) for impulsive loading due to aircraft impact. Initially the load is applied on OCW single model and subsequently the load is transferred to ICW after the local perforation of the OCW is noticed in the transient simulation. In the first stage of the analysis it is demonstrated that the OCW would suffer local perforation with a peak local deformation of 117 mm for impact due to B707-320 and 196 mm due to impact of A300B4 without loss of the overall integrity. However, this first barrier (OCW) cannot absorb the full impulsive load. In the second stage of the analysis of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.19 sec for B707-320 and 0.24 sec for A300B4 due to the local perforation of OCW. This results in the local deformation of approx. 115 mm for B707-320 and 124 mm for A300B4 in ICW and together both the structures (OCW and ICW) are capable of absorbing the full impulsive load. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due commercial aircraft operated in India. (author)

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

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.)

Technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations

International Nuclear Information System (INIS)

Wittenbrock, N.G.

1982-01-01

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWR) and large (1155-MWe) boiling water reactors (BWR) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services

Enhanced safety features of CHASHMA NPP UNIT-2 to encounter selected severe accidents, various challenges involved to prove the adequacy of severe accidents prevention/mitigation measures and to write management guidelines with one possible solution to these challenges

International Nuclear Information System (INIS)

Iqbal, Z.; Minhaj, A.

2007-01-01

This paper describes enhanced safety features of Chashma Nuclear Power Plant Unit-2 (C-2), a 325 MWe PWR to encounter selected severe accidents and discusses various challenges involved to prove the adequacy of severe accidents encountering measures and to write severe accident management guidelines (SAMGs) in compliance with the recently introduced national regulations based on the new IAEA nuclear safety standards. C-2 is being built by China National Nuclear Corporation (CNNC) for Pakistan Atomic Energy Commission (PAEC). Its twin, Unit-1 (C-1) also a 325 MWe PWR, was commissioned in 2000. Nuclear power safety with reference to severe accidents should be treated as a global issue and therefore the developed countries should include the people of developing countries in nuclear power industry's various severe accidents based research and development programs. The implementation of this idea may also deliver few other useful and mutually beneficial byproducts. (author)

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

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

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

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

The status of fast reactor technology development in China

International Nuclear Information System (INIS)

Xu, M.

2002-01-01

China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basic strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan (2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before 2010 respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor (CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached 16.8m above the ground. Forty seven components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000m2 floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which is only under consideration up to now. Some important technical selections have been settled, but its design has not yet started. (author)

The Progress of Fast Reactor Technology Development in China

International Nuclear Information System (INIS)

Yang, Hongyi; Xu, Mi

1994-01-01

China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basis strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan(2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor(CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached the top namely 57m above the ground. More than one hundred components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000m 2 floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which in only under consideration up to now. Some important technical selections have been settled, but its design has not yet started

Multifractality, efficiency analysis of Chinese stock market and its cross-correlation with WTI crude oil price

Science.gov (United States)

Zhuang, Xiaoyang; Wei, Yu; Ma, Feng

2015-07-01

In this paper, the multifractality and efficiency degrees of ten important Chinese sectoral indices are evaluated using the methods of MF-DFA and generalized Hurst exponents. The study also scrutinizes the dynamics of the efficiency of Chinese sectoral stock market by the rolling window approach. The overall empirical findings revealed that all the sectoral indices of Chinese stock market exist different degrees of multifractality. The results of different efficiency measures have agreed on that the 300 Materials index is the least efficient index. However, they have a slight diffidence on the most efficient one. The 300 Information Technology, 300 Telecommunication Services and 300 Health Care indices are comparatively efficient. We also investigate the cross-correlations between the ten sectoral indices and WTI crude oil price based on Multifractal Detrended Cross-correlation Analysis. At last, some relevant discussions and implications of the empirical results are presented.

Heat transport analysis in a district heating and snow melting system in Sapporo and Ishikari, Hokkaido applying waste heat from GTHTR300

International Nuclear Information System (INIS)

Kasahara, Seiji; Kamiji, Yu; Terada, Atsuhiko; Yan Xing; Inagaki, Yoshiyuki; Murata, Tetsuya; Mori, Michitsugu

2015-01-01

A district heating and snow melting system utilizing waste heat from Gas Turbine High temperature Gas Reactor of 300 MW_e (GTHTR300), a heat-electricity cogeneration design of high temperature gas-cooled reactor, was analyzed. Application areas are set in Sapporo and Ishikari, the heavy snowfall cities in Northern Japan. The heat transport analyses are carried out by modeling the components in the system; pipelines of the secondary water loops between GTHTR300s and heat demand district and heat exchangers to transport the heat from the secondary water loops to the tertiary loops in the district. Double pipe for the secondary loops are advantageous for less heat loss and smaller excavation area. On the other hand, these pipes has disadvantage of more electricity consumption for pumping. Most of the heat demand in the month of maximum requirement can be supplied by 2 GTHTR300s and delivered by 9 secondary loops and around 5000 heat exchangers. Closer location of GTHTR300 site to the heat demand district is largely advantageous economically. Less decrease of the distance from 40 km to 20 km made the heat loss half and cost of the heat transfer system 22% smaller. (author)

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)

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)

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

Generation of consistent nuclear properties of DUPIC fuel by DRAGON with ENDF/B-VI nuclear data library

International Nuclear Information System (INIS)

Shen, W.; Rozon, D.

1998-01-01

DRAGON code with 89-groups ENDF/B-VI cross section library was used in this paper to generate consistent nuclear properties of DUPIC fuel. The reference feed material used for the DUPIC fuel cycle is a 17x17 French standard 900 MWe PWR spent fuel assembly with 3.2 w/o initial enrichment and 32500 MWD/7 discharge burnup. The PWR fuel assembly was modeled by JPMT/SYBILT transport method in DRAGON to generate nuclide fields of spent PWR fuel. The resultant nuclide fields constitute the initial fuel composition files for reference DUPIC fuel which can be accessed by DRAGON for CANDU 2D cluster geometry depletion calculation and 3D supercell calculation. Because of uneven spatial power distribution in PWR assemblies and full core, unexpected transition cycle, and various fuel management strategy, the spent PWR fuel composition is expected to be different from one assembly to the next. This heterogeneity was characterized also by modeling various spent PWR fuel assembly types in the paper. (author)

Safety experience on EDF's PWRs

International Nuclear Information System (INIS)

Tanguy, P.

1986-01-01

The french nuclear programme has been widely publicized. In 1985, the total nuclear electricity generated was around 216 GWh, i. e. 70% of the electricity produced by electricity de France (EDF). If we consider only pressurized water reactors, at the end of 1985, 37 units were in operation (32 900 MWe and 5 1300 MWe) and 18 were under construction. I intend to review our experience with the safety of PWR's, but I will first present briefly some aspects related to the safety organization in France and the standardization policy. (author) [pt

Road-map design for thorium-uranium breeding recycle in PWR - 031

International Nuclear Information System (INIS)

Shengyi, Si

2010-01-01

The paper was focused on designing a road-map to finally approach sustainable Thorium-Uranium ( 232 Th- 233 U) Breeding Recycle in current PWR, without any other change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. At first, the paper presented some insights to the inherence of Thorium-Uranium fuel conversion or breeding in PWR based on the neutronics theory and revealed the prerequisites for Thorium-Uranium fuel in PWR to achieve sustainable Breeding Recycle; And then, various Thorium-based fuels were designed and examined, and the calculation results further validated the above theoretical deductions; Based on the above theoretical analysis and calculation results, a road-map for sustainable Thorium-Uranium breeding recycle in PWR was outlined finally. (authors)

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)

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

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

Comparative analysis of a LOCA for a German PWR with ASTEC and ATHLET-CD

International Nuclear Information System (INIS)

Reinke, N.; Chan, H.W.; Sonnenkalb, M.

2013-01-01

This paper presents the results of a comparative analysis performed with ASTEC V2.02 and a coupled ATHLET-CD V2.2c /COCOSYS V2.4 calculation for a German 1300 MWe KONVOI type PWR. The purpose of this analysis is mainly to assess the ASTEC code behaviour in modelling of both the thermal-hydraulic phenomena in the coolant circuit arising during a hypothetical severe accident and the early phase of the core degradation versus the more mechanistic code system ATHLET-CD/COCOSYS. The performed analyses cover a loss of coolant accident sequence (LOCA). Such comparison has been done for the first time. The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed since 1996 by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. The thermal-hydraulic mechanistic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by GRS for the analysis of the whole spectrum of leaks and transients in PWRs and BWRs. For modeling of core degradation processes the CD part (Core Degradation) of ATHLET can be activated. For analyses of the containment behavior, ATHLET-CD has been coupled to the GRS code COCOSYS (COntainment COde SYStem). (orig.)

Combustion and NOx emission characteristics of a retrofitted down-fired 660 MWe utility boiler at different loads

Energy Technology Data Exchange (ETDEWEB)

Li, Z.Q.; Liu, G.K.; Zhu, Q.Y.; Chen, Z.C.; Ren, F. [Harbin Institute of Technology, Harbin (China)

2011-07-15

Industrial experiments were performed for a retrofitted 660 MWe full-scale down-fired boiler. Measurements of ignition of the primary air/fuel mixture flow, the gas temperature distribution of the furnace and the gas components in the furnace were conducted at loads of 660, 550 and 330 MWe. With decreasing load, the gas temperature decreases and the ignition position of the primary coal/air flow becomes farther along the axis of the fuel-rich pipe in the burner region under the arches. The furnace temperature also decreases with decreasing load, as does the difference between the temperatures in the burning region and the lower position of the burnout region. With decreasing load, the exhaust gas temperature decreases from 129.8{sup o}C to 114.3{sup o}C, while NOx emissions decrease from 2448 to 1610 mg/m{sup 3}. All three loads result in low carbon content in fly ash and great boiler thermal efficiency higher than 92%. Compared with the case of 660 MWe before retrofit, the exhaust gas temperature decreased from 136 to 129.8{sup o}C, the carbon content in the fly ash decreased from 9.55% to 2.43% and the boiler efficiency increased from 84.54% to 93.66%.

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

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

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)

User's guide for signal validation software: Final report

International Nuclear Information System (INIS)

Swisher, V.I.

1987-09-01

Northeast Utilities has implemented a real-time signal validation program into the safety parameter display systems (SPDS) at Millstone Units 2 and 3. Signal validation has been incorporated to improve the reliability of the information being used in the SPDS. Signal validation uses Parity Space Vector Analysis to process SPDS sensor data. The Parity Space algorithm determines consistency among independent, redundant input measurements. This information is then used to calculate a validated estimate of that parameter. Additional logic is incorporated to compare partially redundant measurement data. In both plants the SPDS has been designed to monitor the status of critical safety functions (CSFs) and provide information that can be used with plant-specific emergency operating procedures (EOPs). However the CSF logic, EOPs, and complement of plant sensors vary for these plants due to their different design characteristics (MP2 - 870 MWe Combustion Engineering PWR, MP3 - 1150 MWe Westinghouse PWR). These differences in plant design and information requirements result in a variety of signal validation applications

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)

Predisposal of Radioactive Waste from NPP 1000 MWe

International Nuclear Information System (INIS)

Suryantoro

2007-01-01

Predisposal of radioactive waste from NPP 1000 MW which was planned to be operated in 2016 has been conducted. In this study NPP applying PWR type was assumed. This assessment comprises all aspects of radioactive waste coming from NPP. One through cycle was chosen consequently no reprocessing step will be conducted. The assessment shows that technologically all radioactive waste treatment process rising from NPP operation has similarities to the existing radioactive waste process conducted by RWI which has lower scale of waste amount. (author)

Nuclear safety in eastern countries. Background of IPSN's actions

International Nuclear Information System (INIS)

1999-01-01

In this document, IPSN presents its opinion about the safety level that might be reached by the nuclear power plants situated in the former-USSR countries. In these countries 2 types of fission reactors are operating: VVER and RBMK with respectively 46 units and 14 units. 3 generations of VVER-type reactors are coexisting: 440 MWe-230, 440 MWe-213 and 1000 MWe-320. The first generation (440 MWe-230) which involves 11 operating units are the least safe and by no means is it possible to make them reach the western standard of safety. The second generation (440 MWe-213) require technical modifications to near western safety standards. The last generation (1000 MWe-320) has safety levels very similar to PWR's if operating procedures are modified and adapted. RBMK-type reactors have been designed in the years 60-70, they suffer from generic defects due to their design, the poor quality of materials and their low reliability. IPSN fears that any incident in such reactors might turn into a major accident. In order to improve nuclear safety in eastern countries, the European Union has launched an international cooperation, the programmes PHARE and TACIS are presented. (A.C.)

Use of gadolinium as neutron poison in 540 MWe PHWR

International Nuclear Information System (INIS)

Nag, P.K.; Fernando, M.P.S.; Kumar, A.N.

2006-01-01

In Pressurised heavy water reactors (PHWRs), neutron poison in the moderator is used to compensate the excess reactivity present in the core on different occasions such as xenon decay during synchronization just after poison out period or start ups from xenon free conditions. It is also used in secondary shutdown system (SDS-2), where required amount of neutron poison is injected directly into the moderator within 2.5 seconds. Further, it is also used for over poisoning the moderator to achieve the guaranteed shutdown state when the regular shutdown systems are taken for maintenance. Generally, two types of moderator poisons are used in power reactors to balance the reactivity of the core and they are boron and gadolinium. Gadolinium is used in the form of gadolinium nitrate (Gd(NO 3 ) 3 .6H 2 O). The paper gives the details of estimation of reactivity coefficients of gadolinium for 540 MWe PHWR for different operating conditions. These neutron poisons are converted into non-absorbing elements and therefore their effective worth will decrease as reactor operation proceeds. The rate of burning of neutron absorbing isotopes depends on its magnitude of absorption cross-section and thermal flux seen by them. The present study discusses the burning characteristics of gadolinium during power operation in 540 MWe PHWR. It is established by detailed analysis that the rate of positive reactivity realized due to burning of neutron absorbing Gd isotopes almost match with the build up rate of xenon. The burning half lives of boron and gadolinium is worked out for different power levels. (author)

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)

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.

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.

Internal Technical Report, Safety Analysis Report 5 MW(e) Raft River Research and Development Plant

Energy Technology Data Exchange (ETDEWEB)

Brown, E.S.; Homer, G.B.; Shaber, C.R.; Thurow, T.L.

1981-11-17

The Raft River Geothermal Site is located in Southern Idaho's Raft River Valley, southwest of Malta, Idaho, in Cassia County. EG and G idaho, Inc., is the DOE's prime contractor for development of the Raft River geothermal field. Contract work has been progressing for several years towards creating a fully integrated utilization of geothermal water. Developmental progress has resulted in the drilling of seven major DOE wells. Four are producing geothermal water from reservoir temperatures measured to approximately 149 C (approximately 300 F). Closed-in well head pressures range from 69 to 102 kPa (100 to 175 psi). Two wells are scheduled for geothermal cold 60 C (140 F) water reinjection. The prime development effort is for a power plant designed to generate electricity using the heat from the geothermal hot water. The plant is designated as the ''5 MW(e) Raft River Research and Development Plant'' project. General site management assigned to EG and G has resulted in planning and development of many parts of the 5 MW program. Support and development activities have included: (1) engineering design, procurement, and construction support; (2) fluid supply and injection facilities, their study, and control; (3) development and installation of transfer piping systems for geothermal water collection and disposal by injection; and (4) heat exchanger fouling tests.

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)

Evolution of MMI for 500 MWe PHWR plant

International Nuclear Information System (INIS)

Surendar, Ch.; Sharma, M.P.; Jayanthi, S.

1994-01-01

The Indian nuclear power programme for building Pressurized Heavy Water Reactors began with the construction of two units at Kota, Rajasthan. Although the concept of a centralized control room has been used since the beginning, the man-machine interface design has evolved with technological developments. The man-machine interaction in the earliest plants imposed a considerable burden on the operators and led to a need for more sophisticated instrumentation. Several microprocessor and computer based systems were identified and developed and many were retrofitted into existing plants providing immediate advantages. This paper traces the evolution of many of these systems and also describes the basis and the architecture for the man-machine interaction scheme in the 500 MWe nuclear power plants currently being designed. (author). 7 refs., 2 figs., 1 tab

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

The lateral distribution of muons in showers at 40 mwe underground

International Nuclear Information System (INIS)

Bergamasco, L.; Castagnoli, C.; Dardo, M.; D'Ettorre Piazzoli, B.; Mannocchi, G.; Picchi, P.; Visentin, R.; Sitte, K.; Freiburg Univ.

1975-01-01

The multiplicity distribution of muon showers at 40 mwe underground was studied with a 4 m 2 spark chamber telescope. The observed frequencies deviate systematically from those calculated with the 'standard' lateral distributions of Vernov or of Greisen. Agreement can be attained if an enhancement of the muon component at small shower sizes is assumed, in accordance with the assumptions of a two-component theory of cosmic ray origin. It is improved by introducing an age dependence of the lateral structure function. (orig.) [de

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

Two, three and four buttressed PWR containment vessels. A comparative study

International Nuclear Information System (INIS)

Dufour, C.J.; Cheyrezy, M.H.; Thorsen, N.E.

1977-01-01

The purpose of the paper is to analyse the advantages and drawbacks of different arrangements of hood tendons and buttresses for a PWR containment vessel. It is shown that the solution with two buttresses and full hoop tendons (through 360 0 ) gives: acceptable secondary stresses and strains; a lower total cost of the prestressing; a shorter time schedule for prestressing operations; a smaller 'forbidden area' for the penetration lay out. This arrangement has been used for the first time (to our knowledge) in PWR containment design for the nuclear plant of DOEL III, Belgium. The comparative study has been carried out for a containment vessel with a 85 cm thick wall and an interior diameter of 42.50 m. The Guaranteed Ultimate Tensile Strength of the tendons is 9000 KN. The minimum required prestressing force is 10 300 KN per linear meter of height. The buttresses are 4.0 m long and 1.70 m thick on their vertical centerline. The following arrangements have been studied: 4 buttresses and 3/4 hoop tendons; 3 buttresses and 2/3 hoop tendons; 3 buttresses and full hoop tendons; 2 buttresses and full hoop tendons. The ovalisation and other secondary effects result mainly from three different causes: the average prestressing force is not constant around a hoop; the buttress itself constitutes a sudden thickening and creates local disturbances even under an axisymmetric loading: the anchored tendons are straight over their first 6 meters and they initiate local stress perturbations in the vicinity of the junction of the buttress and the wall. The determination of these secondary effects has been performed by a plane stress finite element analysis

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

High Cooling Water Temperature Effects on Design and Operational Safety of NPPs in the Gulf Region

Energy Technology Data Exchange (ETDEWEB)

Kim, Byung Koo [Khalifa Univ., Abu Dhabi (United Arab Emirates); Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2013-12-15

The Arabian Gulf region has one of the highest ocean temperatures, reaching above 35 degrees and ambient temperatures over 50 degrees in the summer. Two nuclear power plants (NPP) are being introduced in the region for the first time, one at Bushehr (1,000 MWe PWR plant from Russia), and a much larger one at Barakah (4Χ1,400 MWe PWR from Korea). Both plants take seawater from the Gulf for condenser cooling, having to modify the secondary/tertiary side cooling systems design by increasing the heat transfer surface area from the country of origin. This paper analyses the secondary side of a typical PWR plant operating under the Rankine cycle with a simplified thermal-hydraulic model. Parametric study of ocean cooling temperatures is conducted to estimate thermal efficiency variations and its associated design changes for the secondary side. Operational safety is reviewed to deliver rated power output with acceptable safety margins in line with technical specifications, mainly in the auxiliary systems together with the cooling water temperature. Impact on the Gulf seawater as the ultimate heat sink is considered negligible, affecting only the adjacent water near the NPP site, when compared to the solar radiation on the sea surface.

HIGH COOLING WATER TEMPERATURE EFFECTS ON DESIGN AND OPERATIONAL SAFETY OF NPPS IN THE GULF REGION

Directory of Open Access Journals (Sweden)

BYUNG KOO KIM

2013-12-01

Full Text Available The Arabian Gulf region has one of the highest ocean temperatures, reaching above 35 degrees and ambient temperatures over 50 degrees in the summer. Two nuclear power plants (NPP are being introduced in the region for the first time, one at Bushehr (1,000 MWe PWR plant from Russia, and a much larger one at Barakah (4X1,400 MWe PWR from Korea. Both plants take seawater from the Gulf for condenser cooling, having to modify the secondary/tertiary side cooling systems design by increasing the heat transfer surface area from the country of origin. This paper analyses the secondary side of a typical PWR plant operating under the Rankine cycle with a simplified thermal-hydraulic model. Parametric study of ocean cooling temperatures is conducted to estimate thermal efficiency variations and its associated design changes for the secondary side. Operational safety is reviewed to deliver rated power output with acceptable safety margins in line with technical specifications, mainly in the auxiliary systems together with the cooling water temperature. Impact on the Gulf seawater as the ultimate heat sink is considered negligible, affecting only the adjacent water near the NPP site, when compared to the solar radiation on the sea surface.

IRSN-ANCCLI partnership. IRSN-ANCCLI seminar on decennial inspections of 900 MWe reactors - November 2010

International Nuclear Information System (INIS)

Rollinger, Francois; Delalonde, Jean-Claude; Hubert, F.; Paulmaz, X.; Tindillere, M.; Lheureux, Y.; Junker, R.; Sene, M.

2010-11-01

This seminar addressed the commitment of local information commissions (CLI) in the analysis and follow-up of the third decennial inspections of the French 900 MWe nuclear reactors. A first session addressed topics directly related to these inspections. Contributions proposed under the form of Power Point presentations by experts and representatives of the IRSN, EDF and CLIs addressed the following issues: safety re-examination of EDF 900 MWe reactors at the occasion of the third decennial inspection, activities of the IRSN related to skill management in nuclear power stations, implementation of the third decennial inspection of the unit 1 of the Fessenheim nuclear power station, the issue of follow-up by a local information commission after a decennial inspection. A second session addressed topics not related to decennial inspections and were proposed by Gravelines and Dampierre local information commissions: analysis of significant safety events, issues of skill management in nuclear power stations

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.)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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)

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

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

AP1000: Meeting economic goals in a competitive world. Annex 7

International Nuclear Information System (INIS)

Davis, G.; Cummins, E.; Winters, J.

2002-01-01

In the U.S., conditions are becoming more favorable for considering the nuclear option again for new baseload generation. While oil and natural gas prices have risen, the cost of operating the existing fleet of nuclear plants has decreased. Furthermore, an advanced 1000 MWe nuclear plant that will be even more cost-competitive with fossil fuels and natural gas will be available by 2005. Westinghouse, in an effort to further improve on the AP600's cost competitiveness, has developed the AP1000, a two-loop, 1000 MWe, advanced pressurized water reactor (PWR) with passive safety features and extensive plant simplifications to enhance the construction, operation, and maintenance. Like the AP600, the AP1000 uses proven technology that builds on over 30 years of operating PWR experience. Westinghouse has completed design studies that demonstrate that it is feasible to increase the power output of the AP600 to at least 1000 MWe, maintaining its current design configuration and licensing basis. To maximize the cost savings, the AP1000 has been designed within the space constraints of the AP600, while retaining the credibility of proven components and substantial safety margins. The affect on the plant's overnight cost of the increased major components that is required to uprate the AP600 to 1000 MWe is small. This overall cost addition is on the order of 11 percent, while the overall power increase is almost 80 percent. This paper describes the changes made to uprate the AP600 and gives an overview of the plant design. (author)

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

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

Comparative studies on trace metal geochemistry in Indian and Chinese rivers

Digital Repository Service at National Institute of Oceanography (India)

Alagarsamy, R.; Zhang, J.

concentration of Indian rivers (~300–1000 µg g–l) is higher than that of Chinese rivers (~150–300 µg g–l) and the global average (~170–350 µg g–l). Molar ratios of alkaline and alkaline earth metals to Fe and Al (i.e. segregation factor, SF) were calculated...

CNSS plant concept, capital cost, and multi-unit station economics

Energy Technology Data Exchange (ETDEWEB)

1984-07-01

United Engineers and Constructors (UE and C) and the Babcock and Wilcox Company (B and W) have performed several studies over the last eight years related to small integral pressurized water reactors. These reactors include the 365 MWt (100 MWe) Consolidated Nuclear Steam Generator (CNSG) and the 1200 MWt Consolidated Nuclear Steam System (CNSS). The studies, mostly performed under contract to the Oak Ridge National Laboratory, have led to a 1250 MWt (400 MWe) Consolidated Nuclear Steam System (CNSS) plant concept, with unique design and cost features. This report contains an update of earlier studies of the CNSS reactor and balance-of-plant concept design, capital costs, and multi-unit plant economics incorporating recent design developments, improvements, and post-TMI-2 upgrades. The economic evaluation compares the total system economic impact of a phased, three stage 400 MWe CNSS implementation program, i.e., a three-unit station, to the installation of a single 1200 MWe Pressurized Water Reactor (PWR) into a typical USA utility system.

The role of nuclear desalination in the Kingdom of Saudi Arabia

International Nuclear Information System (INIS)

Aljohani, M.S.; Abdul-Rahman A.F.; Almarshad, A.I.

2004-01-01

In this study, the role of nuclear desalination in Saudi Arabia is investigated. A water demand forecast between the years 2000 and 2025 was established for the Eastern region of Saudi Arabia as part of the collected input data for the DEEP computer code. The DEEP computer code was run for several options of energy sources such as PWR (600 MWe), SPWR (160 MWth), PHWR (450 MWe), HR (200 MWth) and GT (125 MWe or 175 MWe). These energy sources were investigated for different desalination plants such as RO, MSF, MED and the hybrid MED-RO. The levelised power cost, average daily water production, net saleable power and levelised water cost are presented for all cases. Two scenarios were investigated, the first assumes no interest and discount rates and the second assumes interest and discount rates equal to 8%. The first scenario assumes that the water utility will continue under the control of the government and the second assumes that the water utility will be privatised. (author)

CNSS plant concept, capital cost, and multi-unit station economics

International Nuclear Information System (INIS)

1984-07-01

United Engineers and Constructors (UE and C) and the Babcock and Wilcox Company (B and W) have performed several studies over the last eight years related to small integral pressurized water reactors. These reactors include the 365 MWt (100 MWe) Consolidated Nuclear Steam Generator (CNSG) and the 1200 MWt Consolidated Nuclear Steam System (CNSS). The studies, mostly performed under contract to the Oak Ridge National Laboratory, have led to a 1250 MWt (400 MWe) Consolidated Nuclear Steam System (CNSS) plant concept, with unique design and cost features. This report contains an update of earlier studies of the CNSS reactor and balance-of-plant concept design, capital costs, and multi-unit plant economics incorporating recent design developments, improvements, and post-TMI-2 upgrades. The economic evaluation compares the total system economic impact of a phased, three stage 400 MWe CNSS implementation program, i.e., a three-unit station, to the installation of a single 1200 MWe Pressurized Water Reactor (PWR) into a typical USA utility system

Fire probability safety analysis in France for 900 MWe nuclear power plants

International Nuclear Information System (INIS)

Bertrand, R.; Bonneval, F.; Mattei, J.M.

2000-01-01

This paper describes the methodology implemented by the Institute for Nuclear Safety and Protection (IPSN) to carry out the Fire Probabilistic Safety Assessment (Fire PSA) for French 900 MWe pressurised water reactors. The initial results obtained are presented. Additional research and development activities are indicated which IPSN carried out or decided to perform in order to reduce the amount of uncertainty associated with the data or to confirm hypotheses that can impact significantly the study results. (orig.) [de

Green County Nuclear Power Plant. License application

International Nuclear Information System (INIS)

1975-07-01

The Green County reactor, a PWR to be supplied by Babcock and Wilcox, will be a baseload generating facility planned to provide for mass transit and other public agency electrical needs. The plant is scheduled for completion by 1983 and will have a generating capacity of about 1200 MW(e). (FS)

Programmed system for nuclear power plant protection

International Nuclear Information System (INIS)

Jover, Pierre.

1980-06-01

The progress in the field of microprocessors and large scale integration circuits, have incited to introduce this new technologies into nuclear power plant protection system. The hardware and software design principles are briefly listed; then, a quad-redundant protection system for 1300 MWe PWR, developed in France is described [fr

Experience of partial dismantling and large component removal of light water reactors

International Nuclear Information System (INIS)

Dubourg, M.

1987-01-01

Not any of the French PWR reactors need to be decommissioned before the next decade or early 2000. However, feasibility studies of decommissioning have been undertaken and several dismantling scenarios have been considered including the dismantling of four PWR units and the on-site entombment of the active components into a reactor building for interim disposal. In addition to theoretical evaluation of radwaste volume and activity, several operations of partial dismantling of active components and decontamination activities have been conducted in view of dismantling for both PWR and BWR units. By analyzing the concept of both 900 and 1300 MWe PWR's, it appears that the design improvements taken into account for reducing occupational dose exposure of maintenance personnel and the development of automated tools for performing maintenance and repairs of major components, contribute to facilitate future dismantling and decommissioning operations

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

A comparison of fuzzy logic-PID control strategies for PWR pressurizer control

International Nuclear Information System (INIS)

Kavaklioglu, K.; Ikonomopoulos, A.

1993-01-01

This paper describes the results obtained from a comparison performed between classical proportional-integral-derivative (PID) and fuzzy logic (FL) controlling the pressure in a pressurized water reactor (PWR). The two methodologies have been tested under various transient scenarios, and their performances are evaluated with respect to robustness and on-time response to external stimuli. One of the main concerns in the safe operation of PWR is the pressure control in the primary side of the system. In order to maintain the pressure in a PWR at the desired level, the pressurizer component equipped with sprayers, heaters, and safety relief valves is used. The control strategy in a Westinghouse PWR is implemented with a PID controller that initiates either the electric heaters or the sprayers, depending on the direction of the coolant pressure deviation from the setpoint

Fracture mechanics analysis and evaluation for the RPV of the Chinese Qinshan 300 MW NPP and PTS

International Nuclear Information System (INIS)

He Yinbiao; Isozaki, Toshikuni

2000-03-01

One of the most severe accident conditions of a reactor pressure vessel (RPV) in service is the loss of coolant accident (LOCA). Cold safety injection water is pumped into the downcomer of the RPV through inlet nozzles, while the internal pressure may remain at high level. Such an accident is called pressurized thermal shock (PTS) transient according to 10 CFR 50.61 definition. This paper illustrates the fracture mechanics analysis for the existing RPV of the Chinese Qinshan 300 MW nuclear power plant (NPP) under the postulated PTS transients that include SB-LOCA, LB-LOCA of Qinshan NPP and Rancho Seco transients. 3-D models with the flaw depth range a/w=0.05∼0.9 (a: flaw depth; w: wall thickness) were used to probe what kind of flaw and what kind of transient are most dangerous for the RPV in the end of life (EOF). Both the elastic and elastic-plastic material models were used in the stress analysis and fracture mechanics analysis. The different types of flaw and the influence of the stainless steel cladding on the fracture analysis were investigated for different PTS transients. comparing with the material initiation crack toughness K IC , the fracture evaluation for the RPV in question under PTS transients are performed in this paper. (author)

Radioactive waste processing

International Nuclear Information System (INIS)

Dejonghe, P.

1978-01-01

This article gives an outline of the present situation, from a Belgian standpoint, in the field of the radioactive wastes processing. It estimates the annual quantity of various radioactive waste produced per 1000 MW(e) PWR installed from the ore mining till reprocessing of irradiated fuels. The methods of treatment concentration, fixation, final storable forms for liquid and solid waste of low activity and for high level activity waste. The storage of radioactive waste and the plutonium-bearing waste treatement are also considered. The estimated quantity of wastes produced for 5450 MW(e) in Belgium and their destination are presented. (A.F.)

Evolution of the on-site electric power sources on French 900 MWe PWRs

Energy Technology Data Exchange (ETDEWEB)

Bera, Jean [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, Departement d' Analyse de Surete, Service d' Analyse Fonctionnelle, Institut de Protection et Surete Nucleaire, B.P. No. 6, 92260 Fontenay-aux-Roses (France)

1986-02-15

Additional means have been provided on the French 900 MWe PWRs to improve safety if both the off-site and on-site Power sources are lost, namely: - a primary pump seal water injection device, one for two units; - a gas turbine generator for each site; - supplying any failing unit with electric power from a house load operating unit; - supplying a unit from a diesel generator of another unit. (author)

TRSM-a thermal-hydraulic real-time simulation model for PWR

International Nuclear Information System (INIS)

Zhou Weichang

1997-01-01

TRSM (a Thermal-hydraulic Real-time Simulation Model) has been developed for PWR real-time simulation and best-estimate prediction of normal operating and abnormal accident conditions. It is a non-equilibrium two phase flow thermal-hydraulic model based on five basic conservation equations. A drift flux model is used to account for the unequal velocities of liquid and gaseous mixture, with or without the presence of the noncondensibles. Critical flow models are applied for break flow and valve flow calculations. A 5-regime two phase heat convection model is applied for clad-to-coolant as well as fluid-to-tubing heat transfer. A rigorous reactor coolant pump model is used to calculate the pressure drop and rise for the suction and discharge ends with complete pump characteristics curves included. The TRSM model has been adapted in the full-scale training simulator of Qinshan Nuclear Power Plant 300 MW unit to simulate the thermal-hydraulic performance of the NSSS. The simulation results of a cold leg LOCA and a steam generator tube rupture (SGTR) accident are presented

Technical feasibility study of 60 MWe fast reactor concept: RAPID

International Nuclear Information System (INIS)

Kambe, Mitsuru; Ueda, Nobuyuki; Uotani, Masaki

1993-01-01

A study has been performed on the passive safety features and technical feasibility of an inherently safe 60 MWe fast reactor concept RAPID to meet various power requirements in Japan. The system dynamic analyses on the UTOP and ULOF transients revealed that the enhanced reactivity feedback derived from an annular core configuration and the integrated fuel assembly provides a high margin of self-protection. Structural integrity of the integrated fuel assembly has also been confirmed. The following innovative key technologies have been demonstrated; Lithium Injection Modules (LIM) for ultimate shutdown, Lithium Expansion Modulus (LEM) for inherent reactivity feedback and Void Leading Channel (VLC) for the sodium void worth reduction. (author)

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)

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

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

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

A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

Energy Technology Data Exchange (ETDEWEB)

Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S. [Bhabha Atomic Research Centre, Bombay (India)] [and others

1997-04-01

Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300{degrees}C. Two important observations of the experiments are - appreciable drop in maximum load at 300{degrees}C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis.

A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

International Nuclear Information System (INIS)

Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S.

1997-01-01

Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300 degrees C. Two important observations of the experiments are - appreciable drop in maximum load at 300 degrees C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis

Analysis of the Nonlinear Density Wave Two-Phase Instability in a Steam Generator of 600MWe Liquid Metal Reactor

International Nuclear Information System (INIS)

Choi, Seok Ki; Kim, Seong O

2011-01-01

A 600 MWe demonstration reactor being developed at KAERI employs a once-through helically coiled steam generator. The helically coiled steam generator is compact and is efficient for heat transfer, however, it may suffer from the two-phase instability. It is well known that the density wave instability is the main source of instability among various types of instabilities in a helically coiled S/G in a LMR. In the present study a simple method for analysis of the density wave two phase instability in a liquid metal reactor S/G is proposed and the method is applied to the analysis of density wave instability in a S/G of 600MWe liquid metal reactor

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

The Results of Feasibility Study of Co-generation NPP With Innovative VK-300 Simplified Boiling Water Reactor

International Nuclear Information System (INIS)

Kuznetsov, Yury N.

2006-01-01

The co-generation nuclear power plant (CNPP) producing electricity and district heating heat is planned to be constructed in Archangelsk Region of Russia. Following the 'Letter of Intent' signed by Governor of Archangelsk region and by Minister of the Russian Federation for atomic energy the feasibility study of the Project has been done. The NPP will be based on the four co-generation nuclear power units with the Russian VK-300 SBWR. The innovative passive VK-300 reactor facility has been designed on the basis of well-established nuclear technologies, proven major components, the operating experience of the prototype VK-50 reactor in RIAR, Dimitrovgrad, and the experience in designing such reactors as SBWR (GE) and SWR-1000 (Siemens). The CNPP's total power is planned to be 1000 MW(e) and district-heating heat production capacity 1600 Gcal/h. A detailed description of the results of the feasibility study is presented in the report. The results of the feasibility study have shown that the Archangelsk CGNP is feasible in terms of engineering, economics and production. (authors)

The results of feasibility study of co-generation NPP with innovative VK-300 simplified boiling water reactor

International Nuclear Information System (INIS)

Kuznetsov, Yu. N.; Gabaraev, B. A.

2004-01-01

The co-generation nuclear power plant (CNPP) producing electricity and district-heating heat is planned to be constructed in Archangelsk Region of Russia. Following the Letter of Intent signed by Governor of Archangelsk region and by Minister of the Russian Federation for atomic energy the feasibility study of the Project has been done. The NPP will be based on the four co-generation nuclear power units with the Russian VK-300 SBWR. The innovative passive VK-300 reactor facility has been designed on the basis of well-established nuclear technologies, proven major components, the operating experience of the prototype VK-50 reactor in RIAR, Dimitrovgrad, and the experience in designing such reactors as SBWR (GE) and SWR-1000 (Siemens). The CNPP's total power is planned to be 1000 MW(e) and district-heating heat production capacity 1600 Gcal /h. A detailed description of the results of the feasibility study is presented in the report. The results of the feasibility study have shown that the Archangelsk CGNP is feasible in terms of engineering, economics and production.(author)

Experimental results of the effective water head in downcomer during reflood phase of a PWR LOCA

International Nuclear Information System (INIS)

Sudo, Yukio; Murao, Yoshio; Akimoto, Hajime

1980-08-01

The results and analysis of an experiment for the effective water head in downcomer with 50mm gap size are described. The main objective of the experiment was to clarify the effect of gap size on reflooding in a PWR LOCA. The effective water head in downcomer is the driving force for feeding emergency coolant into the core during reflood phase of a PWR LOCA. Discussions presented here follow those of a previous report in which experimental results and analysis were described for the case of 200mm gap size. Experimental Conditions were: Initial Wall Temperature = 200 -- 300 0 C, Back Pressure = 1 atm., Coolant Temperature = 71 -- 100 0 C, Extraction Water Velocity = 0 -- 2 cm/s, Gap Size = 50 mm. The effective water head history obtained in the experiment was compared with those predicted with Sudo's void fraction correlation. In the prediction, heat input to coolant was calculated from the response of measured wall temperature with heat condition analysis. The experimental results and analysis reveals that: (1) The effects of the gap size and initial wall temperature are evident, (2) The effect of extraction water velocity is negligible, and (3) The predicted history of effective water head is in good agreement with the experimental results except during the transient period in which the effective water head is descreasing. (author)

Assessment of PWR Steam Generator modelling in RELAP5/MOD2

International Nuclear Information System (INIS)

Putney, J.M.; Preece, R.J.

1993-06-01

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

PC based manual and safety logic card test setup for 235 MWe PHWRs

International Nuclear Information System (INIS)

Chandgadkar, G.M.; Kohli, A.K.; Agarwal, R.G.; Chandra, Rajesh

1992-01-01

Fuel handling controls for 235 MWe PHWR make use of Manual and Logic cards (MLCs) for providing safety interlocks. These cards consist of various type of logic blocks. By connecting these logic blocks all the safety interlocks required for fuel handling controls have been provided. Previously trouble shooting of these cards was done by means of logic probe. Since the method was manual, it was laborious and time consuming. PC based test setup has overcome this drawback and detects the fault at the component level within few seconds. It also gives printout of status of faulty MLC cards. Here motherboard has been designed having slots for insertion of MLC cards. The input/output connection of these cards are coming to two 50 pin FRC connectors. PC communicates through 144 line digital input/output card with MLC card under test. Software is user friendly and outputs suitable input patterns to the card under test and checks for output pattern. It compares this output pattern with compare pattern and detects the fault and displays the symptoms. This system is currently in use at test facility for fuelling machine for 235 MWe PHWR reactor at Refuelling Technology Division, Hall-7. This test setup has been proposed for use at NAPP and future reactors. (author). 4 figs., 1 annexure

VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

Directory of Open Access Journals (Sweden)

Muhammad Darwis Isnaini

2016-03-01

Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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)

Control rod studies for alternative fuel cycles in the GA 1160 MW(e) high temperature reactor

Energy Technology Data Exchange (ETDEWEB)

Neef, H. J.

1975-06-15

The control system, which is investigated in this paper for both the low enriched uranium high enriched uranium/thorium fuel cycles, has been developed to control the General Atomics (GA) thorium fuel cycle 1160 MW(e) reactor. It has been shown in this investigation that its effectiveness in the low enriched and subsequent thorium cycle switch-over reactor is equivalent to the effectiveness in the thorium cycle. The shutdown margin in the low enriched core is even higher compared to the thorium core, mainly due to the presence of Pa-233 in the thorium cycle. As long as the fuel cycle for the thorium cycle is not closed with the recycling of U-233, the low enriched cycle will offer an attractive alternative. It was found that the GA 1160 MW(e) control system has enough built-in control rod capacity to accommodate the low enriched uranium cycle and to perform a later switch-over to a thorium-based fuel cycle.

Analysis on the Influence of Stock Index Futures on Chinese Stock Market

Institute of Scientific and Technical Information of China (English)

王钊

2014-01-01

As the first product of financial futures in China, CSI 300 Stock Index Futures is a symbol of the continual improvement and development of Chinese capital market system. So it would be bound to generate immeasurable influence on Chinese capital market and financial system. Starting from introducing the relevant summaries of stock index futures, this paper analyzes the influence of the stock index futures on the fluctuation in the international stock market;then, it analyzes influence of the stock index futures on the fluctuation in Chinese stock market, in order to propose some suggestions to the policies for developing Chinese stock index futures.

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.

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

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

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)

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

Analysis of mechanisms induced by sliding and corrosion: dedicated apparatus for PWR environments

International Nuclear Information System (INIS)

Vernot, JPh

2004-01-01

In pressurized water reactors (PWR), some components are submitted to relative motions due to necessary operational processes (localisation and positioning adjustment) or by not wished effects (flow induced vibration). Thus, components and associated supports are typically excited by a large range of kinematics so than complex combinations of wear can occur. Those excitations can lead to sliding, fretting, impact, etc. Furthermore, typical environment in PWR coupling of temperature (320 deg. C), pressure (154 bars) and chemistry solution (deaerated, low conductivity water) involve specific corrosion processes. Apparently, research performed to date did not deal with all the specific parameters involved at PWR conditions. For this purpose, a specific apparatus has been developed in Framatome Technical Center for a better understanding of this complex degradation mechanism where mechanical and corrosion effects are occurring at the same time. Thanks to electromagnets excitation, mechanical investigations can be proposed with the following combined contact type: pure impact, pure sliding and impact plus sliding for several kinds of sample as rod in a ring, rod against a guide. Motion can be induced on a local area or for the total length (orbital excitation). The relative displacement and the contact force are acquired continuously and permit to establish normal and tangential forces, angular position, sliding distance. On the other hand, electrochemistry measurements have been adapted to the specific apparatus and work in the high temperature water environment. The standard mounting with three electrodes has been qualified so that it is possible to adjust or measure current and potential. All the system is computer controlled and with the present apparatus relationship between mechanical parameters and re-passivation can be studied for specific environments, materials and solicitations. In a first step, potential dynamic polarization curves have been established for

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)

Angra 1 nuclear power plant full scope simulator development project

Energy Technology Data Exchange (ETDEWEB)

Selvatici, Edmundo; Castanheira, Luiz Carlos C.; Silva Junior, Nilo Garcia da, E-mail: edsel@eletronuclear.gov.br, E-mail: lccast@eletronuclear.gov.br, E-mail: nilogar@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (SCO/ELETRONUCLEAR), Angra dos Reis, RJ (Brazil). Superintendencia de Coordenacao da Operacao; Zazo, Francisco Javier Lopez; Ruiz, Jose Antonio, E-mail: jlopez@tecnatom.es, E-mail: jaruiz@tecnatom.es [Tecnatom S.A., San Sebastian de los Reyes, Madrid (Spain)

2015-07-01

Specific Full Scope Simulators are an essential tool for training NPP control room operators, in the formation phase as well as for maintaining their qualifications. In the last years availability of a Plant specific simulator has also become a Regulator requirement for Nuclear Power Plant operation. By providing real-time practical training for the operators, the use of a simulator allows improving the operator's performance, reducing the number of unplanned shutdowns and more effective response to abnormal and emergency operating conditions. It can also be used, among other uses, to validate procedures, test proposed plant modifications, perform engineering studies and to provide operation training for the technical support staff of the plant. The NPP site, in Angra dos Reis-RJ, Brazil, comprises the two units in operation, Unit 1, 640 MWe, Westinghouse PWR and Unit 2, 1350 MWe, KWU/Areva PWR and one unit in construction, Unit 3, 1405 MWe, KWU/Areva PWR, of the same design of Angra 2. Angra 2 has had its full scope simulator from the beginning, however this was not the case of Angra 1, that had to train its operators abroad, due to lack of a specific simulator. Eletronuclear participated in all the phases of the project, from data supply to commissioning and validation. The Angra 1 full scope simulator encompasses more than 80 systems of the plant including the Primary system, reactor core and associated auxiliary systems, the secondary system and turbo generator as well as all the Plant operational and safety I and C. The Angra 1 Main Control Room panels were reproduced in the simulator control room as well as the remote shutdown panels that are outside the control room. This paper describes the project for development of the Angra 1 NPP Full Scope Simulator, supplied by Tecnatom S.A., in the period of Feb.2012 to Feb.2015. (author)

Angra 1 nuclear power plant full scope simulator development project

International Nuclear Information System (INIS)

Selvatici, Edmundo; Castanheira, Luiz Carlos C.; Silva Junior, Nilo Garcia da

2015-01-01

Specific Full Scope Simulators are an essential tool for training NPP control room operators, in the formation phase as well as for maintaining their qualifications. In the last years availability of a Plant specific simulator has also become a Regulator requirement for Nuclear Power Plant operation. By providing real-time practical training for the operators, the use of a simulator allows improving the operator's performance, reducing the number of unplanned shutdowns and more effective response to abnormal and emergency operating conditions. It can also be used, among other uses, to validate procedures, test proposed plant modifications, perform engineering studies and to provide operation training for the technical support staff of the plant. The NPP site, in Angra dos Reis-RJ, Brazil, comprises the two units in operation, Unit 1, 640 MWe, Westinghouse PWR and Unit 2, 1350 MWe, KWU/Areva PWR and one unit in construction, Unit 3, 1405 MWe, KWU/Areva PWR, of the same design of Angra 2. Angra 2 has had its full scope simulator from the beginning, however this was not the case of Angra 1, that had to train its operators abroad, due to lack of a specific simulator. Eletronuclear participated in all the phases of the project, from data supply to commissioning and validation. The Angra 1 full scope simulator encompasses more than 80 systems of the plant including the Primary system, reactor core and associated auxiliary systems, the secondary system and turbo generator as well as all the Plant operational and safety I and C. The Angra 1 Main Control Room panels were reproduced in the simulator control room as well as the remote shutdown panels that are outside the control room. This paper describes the project for development of the Angra 1 NPP Full Scope Simulator, supplied by Tecnatom S.A., in the period of Feb.2012 to Feb.2015. (author)

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.

Meeting the physics design challenges of modern LWRs being inducted into the Indian nuclear power programme

International Nuclear Information System (INIS)

Jagannathan, V.; Pal, Usha; Karthikeyan, R.; Raj, Devesh; Srivastava, Argala; Khan, Suhail Ahmad

2007-01-01

Indian nuclear power programme started with the two Boiling Water Reactors (BWR) of 210 MWe capacity at Tarapur. Two VVER-1000 MWe reactors which are Pressurized Water Reactors (PWR) of Russian design are being constructed at Kudankulam, Tamilnadu and are expected to be commissioned by end 2008. There may be also a possibility of inducting some western PWRs in future. These reactors belong to the category of light water reactors (LWR). The LWRs are compact and have complex physical characteristics distinctly different from those of the Pressurized Heavy Water Reactors (PHWR) which, currently form the mainstay of our indigenous nuclear power programme. The physics design and analysis capability for the modern LWRs (BWR, PWR and VVER) has been developed at Light Water Reactors Physics Section, BARC. This paper presents the current state of art in this key technology area to meet the physics design and operation challenges when LWRs would be inducted in a major way into the Indian nuclear power programme and commence operating in the coming decades. (author)

BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

Energy Technology Data Exchange (ETDEWEB)

Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G. [N.A. Dollezhal Institute ' NIKIET' , PO Box 788, Moscow, 101000 (Russian Federation)

2006-07-01

This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with {beta}{sub eff}, low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

International Nuclear Information System (INIS)

Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G.

2006-01-01

This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with β eff , low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

Progress in the developing countries

International Nuclear Information System (INIS)

Simnad, M.

1981-01-01

Nuclear programmes in selective developing countries are briefly discussed. The oil rich countries of Iraq, Libya and Iran all have reactors on order. Turkey has decided to purchase a PWR from the USSR and Egypt's programme anticipates a capacity of 6600 MWe by 2000. The current projections for India are 6000 MWe by 1990 and 20,000 MWe by 2000. The progress of Pakistan, South Korea and other Asian countries are discussed. The predicted growth in reactors and population in Latin America is considered - 17 reactors presently planned for a population of 340 million and 18-57 possible additions in 2000 for an estimated population of 600 million. The role of the IAEA and experience of some Western countries in technology transfer is discussed with the ambitious Spanish nuclear power programme and the experience of Argentina in purchasing Candu reactors. (author)

Preliminary core design of IRIS-50

International Nuclear Information System (INIS)

Petrovic, Bojan; Franceschini, Fausto

2009-01-01

IRIS-50 is a small, 50 MWe, advanced PWR with integral primary system. It evolved employing the same design principles as the well known medium size (335 MWe) IRIS. These principles include the 'safety-by-design' philosophy, simple and robust design, and deployment flexibility. The 50 MWe design addresses the needs of specific applications (e.g., power generation in small regional grids, water desalination and biodiesel production at remote locations, autonomous power source for special applications, etc.). Such applications may favor or even require longer refueling cycles, or may have some other specific requirements. Impact of these requirements on the core design and refueling strategy is discussed in the paper. Trade-off between the cycle length and other relevant parameters is addressed. A preliminary core design is presented, together with the core main reactor physics performance parameters. (author)

Use of PSA for improving the safety of French PWRs

International Nuclear Information System (INIS)

Lanore, J.M.; Chambon, J.L.

1994-06-01

Two French PWR Probabilistic Safety Assessment (PSA) studies were conducted for the standardized PWR series of 900 and 1300 MWe. Both PSA 900 and PSA 1300 are level 1 PSAs, that means their objective is the evaluation of core meltdown frequency. These studies have some specific features, in particular the treatment of shutdown conditions, the treatment of long term post-accidental situations, and a wide use of French experience feedback. The PSAs are used for safety improvements of the French PWRs. Following the PSA results, several modifications to plants concerning the dominant sequences were decided. (R.P.). 2 refs., 4 figs

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)

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

International Nuclear Information System (INIS)

Choi, In-Kyu; Yeon, Jei-Won; Kim, Won-Ho

2002-01-01

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

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.)

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)

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)

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)

Design of a 2.5MW(e) biomass gasification power generation module

Energy Technology Data Exchange (ETDEWEB)

McLellan, R.

2000-07-01

The purpose of this contract was to produce a detailed process and mechanical design of a gasification and gas clean up system for a 2.5MW(e) power generation module based on the generation of electrical power from a wood chip feed stock. The design is to enable the detailed economic evaluation of the process and to verify the technical performance data provided by the pilot plant programme. Detailed process and equipment design also assists in the speed at which the technology can be implemented into a demonstration project. (author)

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

International Nuclear Information System (INIS)

Lee, C.B.

2002-01-01

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

Assessment of void swelling in austenitic stainless steel PWR core internals

International Nuclear Information System (INIS)

Chung, H.M.

2006-01-01

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

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)

Preliminary design concepts of an advanced integral reactor

International Nuclear Information System (INIS)

Moon, Kap S.; Lee, Doo J.; Kim, Keung K.; Chang, Moon H.; Kim, Si H.

1997-01-01

An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

Three-dimensional model of the thermo-hydrodynamic neutron interaction in the core of water reactors (stationary states)

International Nuclear Information System (INIS)

Mastrangelo, Victor.

1977-01-01

A thermo-hydrodynamic neutron interaction model for permanent working conditions is developed in the case of closed circuits (boiling water reactors) and open circuits (pressurized water reactors). Two numerical convergence acceleration methods are then worked out for the resolution of linear problems by successive iterations. A physical study is devoted to the convergence of the thermo-hydrodynamic neutron interaction process. The model developed is applied to the calculation of the power distribution for the core of a 980 MWe BWR-6 type boiling water power station and to the study of normal and accidental working configurations of the pressurized water core of a 900 MWe PWR-CP1 unit [fr

The use of PSA in the French regulatory practice

International Nuclear Information System (INIS)

Mennesiez, H.

1994-01-01

The presentation gives a description of fundamental documents (since 1977-1978) through which have been set up in France probabilistic objectives, and PSAs, including shutdown states, performed for 900-1300 MWe PWR-type nuclear power plants. PSA developments and use, including fire PSA, level 2 and PSA for the future French-German European Pressurized Reactor (EPR) are also discussed

In-service - pressure test of the primary circuit of the Chooz nuclear power plant

International Nuclear Information System (INIS)

Barthelemy, F.L.; Lespiaucq, P.G.

1977-01-01

A brief summary is given of the regulations governing inspection pratices of operating nuclear power plants, in France. As an example, such an inspection performed in 1976 on the Westinghouse 320 MWe PWR built in Chooz (Ardennes) is described. Emphasis is put on the administrative organization, the technical solutions, the specific problems and the difficulties encountered. (author)

Research on 3D power distribution of PWR reactor core based on RBF neural network

International Nuclear Information System (INIS)

Xia Hong; Li Bin; Liu Jianxin

2014-01-01

Real-time monitor for 3D power distribution is critical to nuclear safety and high efficiency of NPP's operation as well as the control system optimization. A method was proposed to set up a real-time monitor system for 3D power distribution by using of ex-core neutron detecting system and RBF neural network for improving the instantaneity of the monitoring results and reducing the fitting error of the 3D power distribution. A series of experiments were operated on a 300 MW PWR simulation system. The results demonstrate that the new monitor system works very well under condition of certain burnup range during the fuel cycle and reconstructs the real-time 3D distribution of reactor core power. The accuracy of the model is improved effectively with the help of several methods. (authors)

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)

Repowering options for steam power plants

International Nuclear Information System (INIS)

Wen, H.; Gopalarathinam, R.

1992-01-01

Repowering an existing steam power plant with a gas turbine offers an attractive alternative to a new plant or life extension, especially for unit sizes smaller than 300 MWe. Gas turbine repowering improves thermal efficiency and substantially increases the plant output. Based on recent repowering studies and projects, this paper examines gas turbine repowering options for 100 MWe, 200 MWe and 300 MWe units originally designed for coal firing and currently firing either coal or natural gas. Also discussed is the option for a phased future conversion of the repowered unit to fire coal-derived gas, should there be a fluctuation in the price or availability of natural gas. A modular coal gasification plant designed to shorten the conversion time is presented. Repowering options, performance, costs, and availability impacts are discussed for selected cases

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

International Nuclear Information System (INIS)

Fruzzetti, Keith; Choi, Samuel

2012-09-01

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

Study on advanced nuclear fuel cycle of PWR/CANDU synergism

International Nuclear Information System (INIS)

Xie Zhongsheng; Huo Xiaodong

2002-01-01

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

Design of shutdown system no.2 liquid poison injection system for 500 MWe PHWR

International Nuclear Information System (INIS)

Bhatnagar, S.; Balasubrahmanian, A.K.; Pillai, A.V.

1997-01-01

Defence in depth and two group system concepts form the basic design philosophy for the shutdown systems. There are two independent, diverse and fast acting shutdown systems provided for the 500 MWe PHWR. The design is based on fail-safe principle, sufficient component redundancy and on-line testing. Liquid poison injection system, as shutdown system 2, is newly developed for the 500 MWe PHWRs. The system operates by rapidly injecting gadolinium nitrate solution into bulk moderator using stored helium pressure thereby inserting negative reactivity. A high pressure helium supply tank which provides the energy for system actuation, is connected, through an array of fast acting valves in series-parallel arrangement, to the individual poison tanks storing gadolinium nitrate solution. The valves, belonging to three different channels of reactor Protection System 2, are the only active components in the system. The valves are fail safe and are periodically tested on-line without actually firing the system. The system comprising of in-core assemblies and the external process system has been engineered. Experimental work is being carried out by BARC for design validation and data generation. This paper describes the conceptual development, design basis, design parameters and detailed engineering of the system. (author)

Minimum throttling feedwater control in VVER-1000 and PWR NPPs

International Nuclear Information System (INIS)

Symkin, B.E.; Thaulez, F.

2004-01-01

This paper presents an approach for the design and implementation of advanced digital control systems that use a minimum-throttling algorithm for the feedwater control. The minimum-throttling algorithm for the feedwater control, i.e. for the control of steam generators level and of the feedwater pumps speed, is applicable for NPPs with variable speed feedwater pumps. It operates in such a way that the feedwater control valve in the most loaded loop is wide open, steam generator level in this loop being controlled by the feedwater pumps speed, while the feedwater control valves in the other loops are slightly throttling under the action of their control system, to accommodate the slight loop imbalances. This has the advantage of minimizing the valve pressure losses hence minimizing the feedwater pumps power consumption and increasing the net MWe. The benefit has been evaluated for specific plants as being roughly 0.7 and 2.4 MW. The minimum throttling mode has the further advantages of lowering the actuator efforts with potential positive impact in actuator life and of minimizing the feedwater pipelines vibrations. The minimum throttling mode of operation has been developed by the Ukrainian company LvivORGRES. It has been applied with great deal of success on several VVER-1000 NPPs, six units of Zaporizhzha in Ukraine plus, with participation of Westinghouse, Kozloduy 5 and 6 in Bulgaria and South Ukraine 1 to 3 in Ukraine. The concept operates with both ON-OFF valves and true control valves. A study, jointly conducted by Westinghouse and LvivORGRES, is ongoing to demonstrate the applicability of the concept to PWRs having variable speed feedwater pumps and having, or installing, digital feedwater control, standalone or as part of a global digital control system. The implementation of the algorithm at VVER-1000 plants provided both safety improvement and direct commercial benefits. The minimum-throttling algorithm will similarly increase the performance of PWRs. The

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.)

Validating Westinghouse atom 16 x 16 and 18 x 18 PWR fuel performance

International Nuclear Information System (INIS)

Andersson, S.; Gustafson, J.; Jourdain, P.; Lindstroem, L.; Hallstadius, L.; Hofling, C.G.

2001-01-01

Westinghouse Atom designs and fabricates PWR fuel for all major European fuel types: 17 x 17 standard (12 ft) and 17 x 17 XL (14 ft) for Westinghouse type PWRs, and 16 x 16 and 18 x 18 fuel for Siemens type PWRs. The W Atom PWR fuel designs are based on the extensive Westinghouse CE PWR fuel experience from combustion engineering type PWRs. The W atom designs utilise basic design features from the W CE fuel tradition, such as all-Zircaloy mid grids and the proven ( 6 rod years) Guardian TM debris catcher, which is integrated in the bottom Inconel grid. Several new features have been developed to meet with stringent European requirements originating from requirements on very high burnup, in combination with low-leakage core operating strategies and high coolant temperatures. The overall reliability of the Westinghouse Atom PWR fuel is very high; no fuel failure has been detected since 1997. (orig.)

Status and prospects of LWR fuel development in Korea

Energy Technology Data Exchange (ETDEWEB)

Lee, Byong Whi [Department of Nuclear Engineering, Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

1997-12-01

Currently 11 Nuclear Power Plants of 9.6 GWe comprising of 2 KSNP (Korea Standard Nuclear Power Plant - 1000 mwe PWR), 8 PWR and 1 CANDU units are in operation with average availability factor of 87.5% in 1996. 4 KSNP and 3 CANDU units of 6.1 GWe are under construction. 5 KSNP of Evolutionary PWR and CANDU are in its various planning stage. Due to industrialization and economic development, the average annual electric power consumption growth rate was 11.8% in past five years. This growth trend is likely to continue into next century. Therefore, total installed electric power generation capacity of Long Term Electric Power Expansion Planning has to be expanded to 2.2. fold of current capacity by 2010 including 4 KNGR (Korea Next Generation Reactor). 2 figs, 1 tab.

Uncertainty study of the PWR pressure vessel fluence. Adjustment of the nuclear data base

International Nuclear Information System (INIS)

Kodeli, I.A.

1994-01-01

The code system devoted to the calculation of the sensitivity and uncertainty of of the neutron flux and reaction rates calculated by the transport codes, has been developed. Adjustment of the basic data to experimental results can be performed as well. Various sources of uncertainties can be taken into account, such as those due to the uncertainties in the cross-sections, response functions, fission spectrum and space distribution of neutron source, geometry and material composition uncertainties... One -As well as two- dimensional analysis can be performed. Linear perturbation theory is applied. The code system is sufficiently general to be used for various analysis in the fields of fission and fusion. The principal objective of our studies concerns the capsule dosimetry study realized in the framework of the 900 MWe PWR pressure vessel surveillance program. The analysis indicates that the present calculations, performed by the code TRIPOLI-2, using the ENDF/B-IV based, non-perturbed neutron cross-section library in 315 energy groups, allows to estimate the neutron flux and the reaction rates in the surveillance capsules and in the most calculated and measured reaction rates permits to reduce these uncertainties. The results obtained with the adjusted iron cross-sections, response functions and fission spectrum show that the agreement between the calculation and the experiment was improved to become within 10% approximately. The neutron flux deduced from the experiment is then extrapolated from the capsule to the most exposed pressure vessel location using the calculated lead factor. The uncertainty in this factor was estimated to be about 7%. (author). 39 refs., 52 figs., 30 tabs

Localization of the native Chinese speakers language cortex by magnetic source imaging

International Nuclear Information System (INIS)

Sun Jilin; Wu Jie; Li Sumin; Wu Jing; Zhao Huadong; Wu Yujin; Liu Lianxiang

2003-01-01

Objective: To localize the language cortex associated with Chinese word processing by magnetic source imaging (MSI). Methods: Eight right handed and one left handed healthy native Chinese speakers, including 5 men and 4 women, aged from 14 to 32 years, were examined by magnetoencephalography (MEG) and 1.5 T MR unit. All subjects were given 50 times pure tone stimuli (intensity was 80 dB sound pressure level), then 150 pairs of Chinese words (the meaning of the words was related or not related) auditory stimuli (intensity was 80 dB sound pressure level), and then 50 times pure tone stimuli at last (intensity was 80 dB sound pressure level). Evoked response fields (ERFs) time locked to the pure tone and Chinese words were recorded in a magnetically shielded room using a whole-head neuromagnetometer (Model Vectorview 306, made by 4-D Neuroimaging company, Finland) in real-time. The acquired data were averaged by the acquisition computer according to the response to the pure tone, related pairs of words and not related pairs of words. The data obtained by the MEG could be superimposed on MRI. Results: There were two obvious higher magnetic waves named M50 and M100 (two peaks occurred about 50 ms and 100 ms after giving the subjects binaurally stimuli). M50 and M100 in all subjects were localized in the bilateral transverse temporal gyri. The responses to the pairs of Chinese words (the meaning of the words was related or not related) were similar in the same hemisphere of the same subjects. There was a higher peak during 300-600 ms in the right hemisphere in the left handed subject, but there was no peak during 300-600 ms in his left hemisphere. It indicated that the language dominant hemisphere localized in the right hemisphere. Superimposing the MEG data on MRI, the language area was localized in the Wernicke's areas. There were two 300-600 ms response peaks in the bilateral hemispheres (the amplitude of the 300-600 ms response peaks in the bilateral hemisphere was

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

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

Foster Wheeler's Solutions for Large Scale CFB Boiler Technology: Features and Operational Performance of Łagisza 460 MWe CFB Boiler

Science.gov (United States)

Hotta, Arto

During recent years, once-through supercritical (OTSC) CFB technology has been developed, enabling the CFB technology to proceed to medium-scale (500 MWe) utility projects such as Łagisza Power Plant in Poland owned by Poludniowy Koncern Energetyczny SA. (PKE), with net efficiency nearly 44%. Łagisza power plant is currently under commissioning and has reached full load operation in March 2009. The initial operation shows very good performance and confirms, that the CFB process has no problems with the scaling up to this size. Also the once-through steam cycle utilizing Siemens' vertical tube Benson technology has performed as predicted in the CFB process. Foster Wheeler has developed the CFB design further up to 800 MWe with net efficiency of ≥45%.

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

The assessment of technological and safety aspects of small power reactor SMART

International Nuclear Information System (INIS)

Antariksawan, A.R.; Ekariansyah, Andi S.; Sony, D.T.; Suharno; Hastowo, Hudi

2002-01-01

This paper describes and discusses the technology and safety of small nuclear power plant SMART. The reactor SMART produces 300 MWth of power is cooled and moderated with light water and integral PWR type developed by KAERI. At present, the development activities had reached the end of basic design stage. The concept design of reactor SMART is based on safety enhancement, economic competitiveness and high performance. The fuel is uranium oxide with approximately 5% w/o enrichment. The safety characteristics of the core are shown with low power density around 62.6 W/cc, high negative reactivity coefficient, and high shutdown and thermal margin. Besides the inherent safety characteristics, SMART is equipped with engineered safety features and severe accident management system which are in compliance with the IAEA recommendations. The application of SMART for dual-purpose produces 90 Mwe and 40,000 to fresh water a day. Based on the technology and core characteristics of the reactor SMART, it is very interesting to be deeply assessed

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

International Nuclear Information System (INIS)

Dukelow, J.S.; Harrison, D.G.; Morgenstern, M.

1992-03-01

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

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

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

Design and analysis of pressurized water reactor systems

International Nuclear Information System (INIS)

Juhn, P.E.; Kim, Y.H.

1979-01-01

To help develop nuclear engineering technologies in local industry sectors, technical and economical data on pressurized water reactor systems and components have been collected, systematically analyzed and computerized to a certain degree. Codes and standards necessary for engineering design of PWR systems have been surveyed and clarified in terms of NSSS, turbine-generator system and BOP, then again rearranged with respect to quality classes and seismic classes. Some design manuals, criteria and guidelines regarding design, construction, test and operation of PWR plants have also been surveyed and collected. Benchmark cost calculation for the construction of a 900 MWe PWR plant, according to the standard format, was carried out, and computer model on construction costs was improved and updated by considering the local supply of labor and materials. And for the indigeneous development of PWR equipment and materials, such data as delivery schedule and manufacturers of 52 systems and 36,000 components have also been reviewed herein. (author)

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

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.

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

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)

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

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

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

Integrity of pressurized water electronuclear reactor vessels. The case of French reactors

International Nuclear Information System (INIS)

2012-01-01

This document aims at identifying elements related to design, manufacturing and control during operation of reactor vessels of the French electronuclear fleet, and more precisely as far as vessel ferrule is concerned. It briefly describes the typical design and elements of most of French PWR vessels with respect to the reactor type (900 MWe, 1300 MWe, 1450 MWe, EPR). It recalls some measures regarding design (for embrittlement assessment) and manufacturing processes (forging operations for shell fabrication, coatings). It discusses the different manufacturing defects which have been noticed (under the coatings, due to hydrogen, and intergranular loss of cohesion due to re-heating). It more particularly comments defects noticed on a Belgium power station reactor in Doel, defects due to hydrogen and some other defects noticed in the French reactor fleet. It presents the different types of control which are performed on vessel shells during operation

Investigation of human performance events at French power stations

International Nuclear Information System (INIS)

Ghertman, F.; Griffon-Fouco, M.

1985-01-01

This paper is concerned with the collection of data on human errors that occur at operating power plants. Three collection methods are used, each relating to a difference level of analysis. (1) Simplified statistical analysis of the causes of human errors: Events which have occurred at operating power plants and which are attributable to human errors are selected. The errors thus identified are analysed briefly and are described by a simplified classification, statistical analysis then being applied to find the principal factors underlying these errors. By way of example, an analysis is given of data on emergency shut-downs involving a human error component that occurred at 900 MW(e) PWR plants during 1982, 1983, 1984. (2) In-depth statistical analysis of the causes of certain human errors: The errors selected are analysed and described by means of a detailed classification. By way of example, the collection and evaluation of data on human errors occurring during periodic tests at a 900 MW(e) power plant over a period of six months are described. (3) In-depth analysis of certain events due to human errors: The events selected are analysed by means of a method which reconstitutes the multicausal aspect of the event and of each human error. By way of example, a description is given of an emergency core cooling required at a 900 MW(e) PWR plant. In conclusion, it is explained how these three methods of collection play complementary roles

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

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

SODEXPERT: help to PWR plant management to prevent secondary circuit corrosion

International Nuclear Information System (INIS)

Eon-Duval, P.; Fiquet, J.M.; Langlet, J.P.

1996-01-01

Since about 10 years, problems of secondary circuit corrosion have raised for PWR plant management. The watch staff can't be asked the physicochemical knowledge requested for a proper interpretation of the various probes outputs. So an expert-system has been performed to help the identification of dangerous situation from a corrosion point of view, and immediately start the PWR managing action. This software has been successfully tested and validated. (D.L.). 5 figs., 4 photos

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)

Dry Ice Blast Decontamination to in-service equipment in Japanese PWR plant

International Nuclear Information System (INIS)

2016-01-01

MHI had developed several mechanical decontamination methods. Mechanical decontamination is beneficial when it is applied to equipment whose surface is narrow. Especially in terms of secondary waste reduction, MHI started the study of application of Dry Ice Blast Decontamination to actual PWR plant. This paper provides an introduction to Dry Ice Blast Decontamination principle, its system and actual application result to PWR plant. (J.P.N.)

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)

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.

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)

Maintenance service for major component of PWR plant. Replacement of pressurizer safe end weld

International Nuclear Information System (INIS)

Miyoshi, Yoshiyuki; Kobayashi, Yuki; Yamamoto, Kazuhide; Ueda, Takeshi; Suda, Naoki; Shintani, Takashi

2017-01-01

In October 2016, MHI completed the replacement of safe end weld of pressurizer (Pz) of Ringhals unit 3, which was the first maintenance work for main component of pressurized water reactor (PWR) plant in Europe. For higher reliability and longer lifetime of PWR plant, MHI has conducted many kinds of maintenance works of main components of PWR plants in Japan against stress corrosion cracking due to aging degradation. Technical process for replacement of Pz safe end weld were established by MHI. MHI has experienced the work for 21 PWR units in Japan. That of Ringhals unit 3 was planned and conducted based on the experiences. In this work, Alloy 600 used for welds of nozzles of Pz was replaced with Alloy 690. Alloy 690 is more corrosive-resistant than Alloy 600. Specially designed equipment and technical process were developed and established by MHI to replace safe end weld of Pz and applied for the Ringhals unit 3 as a first application in Europe. The application had been performed in success and achieved the planned replacement work duration and total radiation dose by using sophisticated machining and welding equipment designed to meet the requirements to be small, lightweight and remote-controlled and operating by well skilled MHI personnel experienced in maintenance activities for major components of PWR plant in Japan. The success shows that the experience, activities and technology developed in Japan for main components of PWR plant shall be applicable to contribute reliable operations of nuclear power plants in Europe and other countries. (author)

Fatigue cycles evaluation of 500 MWe PHWR coolant channel sealdisc

International Nuclear Information System (INIS)

Chawla, D.S.; Vaze, K.K.; Kushwaha, H.S.; Gupta, K.S.; Bhambra, H.S.

1998-07-01

At each end of coolant channel there is one sealing plug assembly. The sealdisc is a part of sealing plug assembly. The sealdisc is used to avoid leakage of heavy water. The importance of sealdisc can be understood by the fact that there are 784 sealdiscs in one 500 MWe PHWR unit. During the life time of reactor the sealdisc will be subjected to cyclic loads due to reactor startup, shutdown, power setback and also due to refuelling operations. Excessive reversal of stresses may lead to fatigue failure. The sealdisc failure may cause loss of coolant accidents. Since sealdisc is safety class 1 component, it has to be qualified according to ASME Section III Division 1 NB. For cyclic loads, the fatigue analysis is essential to assess the allowable number of cycles and also to check the total usage factor due to different cyclic loads. To evaluate the allowable fatigue cycles, the analysis is carried out using finite element method. The present report deals with the fatigue cycles evaluation of 500 MWe PHWR sealdisc. The finite element model having eight noded axisymmetric elements is used for the analysis. The various loads considered in the analysis are mechanical loads arising due to refuelling operations and number of temperature-pressure transients. During refuelling, the sealdisc is removed and reinstalled back by use of fuelling machine ram which applies load at centre as well as at rocker point of sealdisc. The stress analysis is carried out for each stage of loading during refuelling and fatigue cycles are evaluated. For temperature transient, decoupled thermal analysis is carried out. At various instants of time, the stresses are computed using temperatures calculated in thermal analysis. The pressure variation is also considered along with temperature variation. The fatigue cycles are evaluated for each transient using maximum alternating stress intensities. The usage factors are calculated for various temperature/pressure transients and refuelling loads

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)

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

International Nuclear Information System (INIS)

Fruzzetti, K.; Choi, S.; Haas, C.; Pender, M.; Perkins, D.

2010-01-01

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

A review of start-up operations on the first units of the 1300 MWe generation

International Nuclear Information System (INIS)

Meclot, B.; Lemagny Boc Lonlaygue, C.; Lavogiez, M.

1986-01-01

This paper describes and offers comments on the different phases of start-up on power stations of the P4 series. Then, one reviews incidents which occurred in the course of these start-up phases and, having highlighted the lessons to be learnt from the commissioning of these power stations, goes on to make a comparative study of 1300 and 900 MWe availability in the initial year of operation [fr

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

Quantitative analysis technique for Xenon in PWR spent fuel by using WDS

Energy Technology Data Exchange (ETDEWEB)

Kwon, H. M.; Kim, D. S.; Seo, H. S.; Ju, J. S.; Jang, J. N.; Yang, Y. S.; Park, S. D. [KAERI, Daejeon (Korea, Republic of)

2012-01-15

This study includes three processes. First, a peak centering of the X-ray line was performed after a diffraction for Xenon La1 line was installed. Xe La1 peak was identified by a PWR spent fuel sample. Second, standard intensities of Xe was obtained by interpolation of the La1 intensities from a series of elements on each side of xenon. And then Xe intensities across the radial direction of a PWR spent fuel sample were measured by WDS-SEM. Third, the electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to do matrix correction of a PWR spent fuel sample. Finally, the method and the procedure for local quantitative analysis of Xenon was developed in this study.

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

International Nuclear Information System (INIS)

Muhammad Subekti

2009-01-01

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

Quantitative analysis technique for Xenon in PWR spent fuel by using WDS

International Nuclear Information System (INIS)

Kwon, H. M.; Kim, D. S.; Seo, H. S.; Ju, J. S.; Jang, J. N.; Yang, Y. S.; Park, S. D.

2012-01-01

This study includes three processes. First, a peak centering of the X-ray line was performed after a diffraction for Xenon La1 line was installed. Xe La1 peak was identified by a PWR spent fuel sample. Second, standard intensities of Xe was obtained by interpolation of the La1 intensities from a series of elements on each side of xenon. And then Xe intensities across the radial direction of a PWR spent fuel sample were measured by WDS-SEM. Third, the electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to do matrix correction of a PWR spent fuel sample. Finally, the method and the procedure for local quantitative analysis of Xenon was developed in this study

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

International Nuclear Information System (INIS)

Temple, S.M.; Robbins, T.R.

1986-09-01

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

Aging management of PWR reactor internals in U.S. plants

International Nuclear Information System (INIS)

Amberge, K.J.; Demma, A.

2015-01-01

This paper describes the development, aging management strategies and inspection results of the Pressurized Water Reactor (PWR) vessel internals inspection and evaluation guidelines. The goal of these guidelines is to provide PWR owners with robust aging management strategies to monitor degradation of internals components to support life extension as well as the current period of operation and power up-rate activities. The implementation of these guidelines began in 2010 within the U.S. PWR fleet and several examinations have been performed since. Examples of inspection results are presented for selected vessel internals components and are compared with simulation results. In summary, to date there have been no observations of austenitic stainless steel stress corrosion cracking (SCC), which is consistent with expectations based on the current understanding of the mechanism. Observations of irradiation assisted stress corrosion cracking (IASCC) have been limited and only found in baffle former bolting. Additionally, no macroscopic effects or global observations of void swelling impacts on general conditions of reactor internal hardware have been observed. (authors)

Containment loading during severe core damage accidents

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

1984-11-01

The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

Neutron flux measurement by mobile detectors

International Nuclear Information System (INIS)

Verchain, M.

1987-01-01

Various incore instrumentation systems and their technological evolution are first reviewed. Then, for 1300 MWe PWR nuclear power plant, temperature and neutron flux measurement are described. Mobile fission chambers, with their large measuring range and accurate location allow a good knowledge of the core. Other incore measures are possible because of flux detector thimble tubes inserted in the reactor core [fr

Modification to 200 MW(e) CANDU for improved dynamic behaviour

International Nuclear Information System (INIS)

Chamany, B.F.; Murthy, L.G.K.; Ray, R.N.

1976-01-01

Rajasthan Atomic Power Station is inherently suitable for base load operation. Its control philosophy is based on turbine following the reactor. However, due to load fluctuations and inherent limitation of the control system, there had been considerable number of outages of the station. This limitation is further enhanced by improper choice of the operating pressure range of the boilers. Besides, existing fuel design does not permit thermal cycling and hence there is no use in attempting to make the reactor follow the turbine. Design modifications have been suggested for incorporation in the further 200 MW(e) systems. The method adopted is complete decoupling of the reactor from the load. Dynamic behaviour of the station with the suggested modifications and its comparison with the existing situation has been brought out. (author)

Effects of some structural materials on the reactivity and flux distributions in a pressurised water reactor

International Nuclear Information System (INIS)

Mondal, A.M.W.; Mannan, M.A.

1983-01-01

The effect of the structural materials of the guide tubes, spacer grids and the shroud on the reactivity and the flux distribution of a Pressurised Water Reactor (PWR) has been studied. Group constants of different cells of guide tubes, spacer grids, shroud and the fuel have been calculated using the cell codes LEOPARD, PANTHER and METHUSELAH. Core calculations have been performed using the diffusion code EQUIPOISE. It has been found for a PWR of 1300 MWe of Kraftwork Union design for Iron that the total change in reactivity due to the presence of guide tubes, spacer grids and the shroud is about -2.48x10 -2 . (author)

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

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

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

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.

PWR plant operator training used full scope simulator incorporated MAAP model

International Nuclear Information System (INIS)

Matsumoto, Y.; Tabuchi, T.; Yamashita, T.; Komatsu, Y.; Tsubouchi, K.; Banka, T.; Mochizuki, T.; Nishimura, K.; Iizuka, H.

2015-01-01

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

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.

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

In-pile loop studies of the effect of PWR coolant pH on corrosion product radionuclide deposition

International Nuclear Information System (INIS)

Driscoll, M.J.; Harling, O.K.; Kohse, G.E.

1992-02-01

An in-pile loop which simulates the primary coolant system of a PWR has been constructed and operated in the MIT research reactor. A total of seven one-month-long irradiations have been carried out to evaluate the effect of coolant pH controlled by variation in LiOH/H 3 BO 3 concentrations. With the exception of one run at zero boron, all employed 800 ppm B; pH 300degreesC values of 6.5, 7.0, 7.2, 7.5 were studied, and two runs each at 7.0 and 7.2 were carried out. Finally, one of the runs at a pH 300degreesC of 7.2 was conducted with special care to exclude zinc because of its potential effects on cobalt deposition. The results show the expected benefits of high pH in reducing the rate of activity deposition on plant surfaces, but pH 300degreesC = 7.2 is approximately as effective as 7.5, while pH 300degreesC = 6.5 exhibits much larger activity transport and qualitatively different deposition behavior. Significant heat flux effects not predicted by current models have been consistently observed. While not as extensively studied, the zero-boron run suggests that the presence of boron species, at fixed pH, may reduce the net amount of activity deposited on ex-core surfaces. Neutron activation analysis of a variety of samples ruled out Zircaloy as an important source of Co-60, since its cobalt content is less than one ppm, considerably less than the applicable ASTM specification of ≤ 20 ppm. Amendment of the latter has been recommended

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

The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

Science.gov (United States)

Cockeram, B. V.; Kammenzind, B. F.

2018-02-01

Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

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

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.)

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.

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.

Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Design Requirements Document (DRD)

Science.gov (United States)

Rigo, H. S.; Bercaw, R. W.; Burkhart, J. A.; Mroz, T. S.; Bents, D. J.; Hatch, A. M.

1981-01-01

A description and the design requirements for the 200 MWe (nominal) net output MHD Engineering Test Facility (ETF) Conceptual Design, are presented. Performance requirements for the plant are identified and process conditions are indicated at interface stations between the major systems comprising the plant. Also included are the description, functions, interfaces and requirements for each of these major systems. The lastest information (1980-1981) from the MHD technology program are integrated with elements of a conventional steam electric power generating plant.

Analysis of operator's behaviour under accidental transients

International Nuclear Information System (INIS)

Llory, M.; Lemaitre, D.; Griffon-Fouco, C.; Meslin, B.

1992-01-01

Since 1979, EDF has been conducting intensive test campaigns on full-scale PWR simulators in order to study and improve the operators behaviour under incident as well as accident conditions. This paper presents some results obtained during tests carried out in 1986 on the P4 (1300 MWe power plant series) simulators of the Paluel Training Center. These results essentially concern the observed deviations, the diagnosis and the safety engineer's role. They are compared with the results of previous tests on 900 MWe unit simulators. The test organization and methodology, the result analysis methods and the biases introduced by this kind of test are also discussed. (author). 7 refs, 1 fig., 6 figs

Prospects of power ramping and cycling supervision in Finnish power reactors

Energy Technology Data Exchange (ETDEWEB)

Antila, M; Kaikkonen, H T [Imatran Voima Oy, Helsinki (Finland); Mannola, E [Teollisuuden Voima Oy Industries Kraft Ab, Helsinki (Finland)

1983-06-01

Since 1977 2x440 MWe PWR and 2x660 MWe BWR nuclear power has been taken in operation in Finland, which until the middle of 1982 has given favourable fuel operating experiences from 10 reactor years. This paper describes the core supervision systems of the plants especially from the viewpoint of ramp surveillance and the potentials and needs to improve the supervision capability to meet the future needs in case more load follow operation is required. As a special feature for Imatran Voima is the demand of general basic understanding of the behaviour of Loviisa reactors' fuel in different operating conditions. A possibility to investigate the fuel seem to be power cycling tests in Loviisa reactors. (author)

Prospects of power ramping and cycling supervision in Finnish power reactors

International Nuclear Information System (INIS)

Antila, M.; Kaikkonen, H.T.; Mannola, E.

1983-01-01

Since 1977 2x440 MWe PWR and 2x660 MWe BWR nuclear power has been taken in operation in Finland, which until the middle of 1982 has given favourable fuel operating experiences from 10 reactor years. This paper describes the core supervision systems of the plants especially from the viewpoint of ramp surveillance and the potentials and needs to improve the supervision capability to meet the future needs in case more load follow operation is required. As a special feature for Imatran Voima is the demand of general basic understanding of the behaviour of Loviisa reactors' fuel in different operating conditions. A possibility to investigate the fuel seem to be power cycling tests in Loviisa reactors. (author)

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

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)

A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

Energy Technology Data Exchange (ETDEWEB)

NONE

1990-04-15

To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10{sup -6} per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective

A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

International Nuclear Information System (INIS)

1990-04-01

To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10 -6 per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective of

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

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

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

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

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

International Nuclear Information System (INIS)

Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

2015-01-01

Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

Factors Impacting Openness to Christianity among Chinese Graduate Students Who Attended a Christian University in the United States

Science.gov (United States)

Chamberlain, Michael A.

2017-01-01

More than 300,000 Chinese students attend U.S. universities annually (USDHS, 2017), many of whom reportedly "Leave China, Study in America, Find Jesus" (H. Zhang, 2016). However, research on this phenomenon of worldview change is thin, especially experiences of atheist or nonreligious Chinese graduate students attending Christian…

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

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

International Nuclear Information System (INIS)

Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

2002-01-01

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

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.

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

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.

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

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

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)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

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

Systems analysis of a 100-MWe modular liquid metal cooled reactor

International Nuclear Information System (INIS)

Morris, E.E.; Rhow, S.K.; Switick, D.M.

1985-01-01

The response of a 100-MWe modular liquid metal cooled reactor to unprotected loss of flow and/or loss of primary heat removal accidents is analyzed using the systems analysis code SASSYS. The reactor response is tracked for the first 1000 s following a postulated upset in the primary heat removal system. The calculations do not take credit for the functioning of any decay heat removal other than through the secondary system. In addition to the power rating, other features of the reactor are an average sodium temperature rise of 148 K, a sodium void worth (counting the core and upper axial blanket) of 1.89 $, and 3.6 $ of Doppler feedback due to a uniform e-fold fuel temperature increase

Capture of SO2 by limestone in a 71 MWe pressurized fluidized bed boiler

Directory of Open Access Journals (Sweden)

Shimizu Tadaaki

2003-01-01

Full Text Available A 71 MWe pressurized fluidized bed coal combustor was operated. A wide variety of coals were burnt under fly ash recycle conditions. Limestone was fed to the combustor as bed material as well as sorbent. The emission of SO^ and limestone attrition rate were measured. A simple mathematical model of SO? capture by limestone with intermittent solid attrition was applied to the analysis of the present experimental results. Except for high sulfur fuel, the results of the present model agreed with the experimental results.

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

International Nuclear Information System (INIS)

Mursid Djokolelono.

1976-01-01

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

34 CFR 300.300 - Parental consent.

Science.gov (United States)

2010-07-01

...; and (iii) Is not required to convene an IEP Team meeting or develop an IEP under §§ 300.320 and 300... related services; and (iv) Is not required to convene an IEP Team meeting or develop an IEP under §§ 300...

Scope and procedures of fuel management for PWR nuclear power plant

International Nuclear Information System (INIS)

Yao Zenghua

1997-01-01

The fuel management scope of PWR nuclear power plant includes nuclear fuel purchase and spent fuel disposal, ex-core fuel management, in-core fuel management, core management and fuel assembly behavior follow up. A suit of complete and efficient fuel management procedures have to be created to ensure the quality and efficiency of fuel management work. The hierarchy of fuel management procedure is divided into four levels: main procedure, administration procedure, implement procedure and technic procedure. A brief introduction to the fuel management scope and procedures of PWR nuclear power plant are given

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)

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

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

Energy Technology Data Exchange (ETDEWEB)

Matsumoto, Hideki; Onoue, Masaaki; Tahara, Yoshihisa [Mitsubishi Heavy Industries Ltd., Tokyo (Japan)

2001-08-01

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

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

International Nuclear Information System (INIS)

Tylee, J.L.

1981-01-01

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

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

Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant Conceptual Design Engineering Report (CDER)

Science.gov (United States)

1981-01-01

The reference conceptual design of the magnetohydrodynamic (MHD) Engineering Test Facility (ETF), a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commercial feasibility of open cycle MHD, is summarized. Main elements of the design, systems, and plant facilities are illustrated. System design descriptions are included for closed cycle cooling water, industrial gas systems, fuel oil, boiler flue gas, coal management, seed management, slag management, plant industrial waste, fire service water, oxidant supply, MHD power ventilating

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

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

1993-03-01

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

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)

A comparison of in-vessel behaviors between SFR and PWR under severe accident

Energy Technology Data Exchange (ETDEWEB)

Park, Sanggil; Cho, Cheon Hwey [ACT Co., Daejeon (Korea, Republic of); Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

This paper aims to provide an easy guide for experts who know well the severe accident phenomenology of Pressurized Water Reactor (PWR) by comparing both reactor design concepts and in vessel behaviors under a postulated severe accident condition. This study only provides a preliminary qualitative comparison based on available literature. The PWR and SFR in-vessel design concepts and their effects under a postulate severe accident are investigated in this paper. Although this work is a preliminary study to compare the in-vessel behaviors for both PWR and SFR, it seems that there is no possibility to lead a significant core damage in the metal fuel SFR concept. In the oxide fuel SFR, there might be a chance to progress to the severe accident initiators such as the energetic reaction, flow blockage and so on.

Krsko: A case study for western nuclear power issues

International Nuclear Information System (INIS)

Bizal, M.R.

1996-01-01

This article is a review of issues surrounding the operation of the Krsko reactor in the former Yugoslav state of Slovenia. This reactor, a Westinghouse-supplied 632-MWe PWR, supplies 25% of Slovenia's electrical energy and is therefore a major element in the country's energy mix. There are environmental issues and political issues noted, and the project is currently evaluating the need for costly steam generator replacement

Toward full MOX core design

International Nuclear Information System (INIS)

Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

1999-01-01

This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

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)

A preliminary redrafting of the Italian technical guide no.26 on radioactive waste management

International Nuclear Information System (INIS)

Tocci, M.

1993-01-01

The recent National Energy Plan confirms the need for reprocessing of the nuclear spent fuel arising from the Nuclear Power Plants (NPPs) that were definitively put out of operation after Chernobyl accident. Two campaigns concerning the Italian spent fuel reprocessing are starting at this time. The first-one is related to LATINA (MAGNOX, GCR 160 MWe) while the second one will concern GARIGLIANO (BWR 160 MWe), CAORSO (BWR 840 MWe) and TRINO (PWR, 200 MWe). The first campaign will involve about 600 tons of MAGNOX spent fuel arising from LATINA NPP which will be reprocessed in the next future by BNFL (British Nuclear Fuel Limited) at Sellafield. The vitrified High Level Wastes (HLW) shall return to Italy starting from 1994. ENEA-DISP (Italian directorate for Nuclear Safety) has received from BNFL the technical specifications of the glass blocks. The approval of those specifications is scheduled by the end of 1993, in order to allow the delivery of the wastes. During the licensing iter ENEA-DISP will use its own Technical Guide N. 26, which identifies and specifies the basic criteria and requirements for an appropriate waste management policy. (author). 4 figs., 5 refs

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

International Nuclear Information System (INIS)

Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

2012-01-01

700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

Assessment on 900–1300 MWe PWRs of the ASTEC-based simulation tool of SGTR thermal-hydraulics for the IRSN Emergency Technical Centre

Energy Technology Data Exchange (ETDEWEB)

Foucher, L., E-mail: laurent.foucher@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, Saint-Paul-lez-Durance 13115 (France); Cousin, F.; Fleurot, J. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, Saint-Paul-lez-Durance 13115 (France); Brethes, S. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-CRI/SESUC, Cadarache, Saint-Paul-lez-Durance 13115 (France)

2014-06-01

In the event of an accident occurring in a nuclear power plant (NPP), being able to predict the amount of released radioactive substances in the environment is of prime importance. Depending on the severity of the accident, it can be necessary to quickly and efficiently protect the population and the surrounding environment from the associated radiological consequences. In France, the IRSN Emergency Technical Centre provides a technical support in decision making in case of a nuclear accident. The main objectives are to evaluate and predict the plant behaviour and radioactive releases during the accident. Different types of complementary tools are used: expert assessments, pre-calculated databases, simulation tools, etc. In the case of Steam Generator Tube Rupture (SGTR) accidents that may lead to significant radioactive releases to the atmosphere through the steam generator relief valves, IRSN is currently improving the simulation tools for diagnosis in crisis management. The objective is to adapt the thermal-hydraulic and FP behaviour modules of the severe accident integral code ASTEC V2.0, jointly developed by IRSN and its German counterpart GRS, to crisis management requirements. These requirements impose a fast running, highly reliable (accurate physical results), flexible and simple tool. This paper summarizes the results of the benchmarks between the ASTEC V2.0 thermal-hydraulic module and the CATHARE 2 (V2.5) French reference thermal-hydraulics code on several SGTR scenarios both for PWR 900 and 1300 MWe, with a particular emphasis on the computational time and physical models assessment. The overall agreement between both codes is good on the primary and secondary circuit thermal-hydraulic parameters. Moreover, the reliability and fast computational time of the thermal-hydraulic module of ASTEC V2.0 code appeared very satisfactory and in accordance with the requirements of an emergency tool.