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

International Nuclear Information System (INIS)

Jiang Zhenbiao; Li Guanying; Liu Zhicheng

2013-01-01

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

Siemens Nuclear Power Corporation methods development for BWR/PWR reactor licensing

International Nuclear Information System (INIS)

Pruitt, D.W.

1992-01-01

This presentation addresses the Siemens Nuclear Power Corporation (SNP) perspective on the primary forces driving methods development in the nuclear industry. These forces are fuel design, computational environment and industry requirement evolution. The first segment of the discussion presents the SNP experience base. SNP develops, manufactures and licenses both BWR and PWR reload fuel. A review of this experience base highlights the accelerating rate at which new fuel designs are being introduced into the nuclear industry. The application of advanced BWR lattice geometries provides an example of fuel design trends. The second aspect of the presentation is the rapid evolution of the computing environment. The final subject in the presentation is the impact of industry requirements on code or methods development

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)

Ventilation and air-conditioning system for PWR nuclear power plant

International Nuclear Information System (INIS)

Ohmoto, Kenji

1987-01-01

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

A digital simulation of a pressurizer in a PWR nuclear power plant

International Nuclear Information System (INIS)

Sato, E.F.

1980-11-01

A model for pressurizer digital simulation of a PWR nuclear power plant during transients, considering all pressurizer control features, is presented. The pressurizer is divided into two regions separated by a water-vapor interface and non-equilibrium conditions are considered. The particular thermodynamic process followed during insurge and outsurges is determined at each instant of analysis without any previous assumption. The pressure behavior is defined by an explicit equation in any of four possible pressurizer thermodynamic conditions. Thermodynamic properties of steam and water are computed by ASME subroutines and the mathematical formulation presented in this study was programed in FORTRAN IV for a Burroughs-6700 digital computer system. This program was employed to simulate the Shippingport Atomic Power Station and Almirante Alvaro Alberto Nuclear Power Plant - Unit 1 pressurizers. The test results compared with experimental or vendor data show the validity of this analysis method. (Author) [pt

Water chemistry control of PWR nuclear power plant

International Nuclear Information System (INIS)

Hino, Yuichi; Makino, Ichiro; Yamauchi, Sumio; Fukuda, Fumihito.

1992-01-01

In PWR power plants, the primary system taking heat out of nuclear reactors and the secondary system generating steam and driving turbines are completely separated by steam generators, accordingly, by mutually independent water treatment, both systems are to be maintained in the optimal conditions. Namely, primary system is the closed water circulation circuit of simple liquid phase though under high temperature, high pressure condition, therefore, water shows the stable physical and chemical properties, and the minute water treatment for restraining the corrosion of structural materials and reducing radioactivity can be done. Secondary system is similar to the condensate and feedwater system of thermal power plants, and is the circuit for liquid-vapor two-phase transformation, but due to the local concentration of impurities by evaporation, the strict requirement is set for secondary water quality. However, secondary system can be treated in the state without radioactivity, and this is a great merit. The outline, basic concept and execution of primary water quality control, and the outline, concept, control criteria, facilities and execution of secondary water quality control are reported. (K.I.)

The function of single containment and double containment of PWR nuclear power plant

International Nuclear Information System (INIS)

Chen Weijing.

1985-01-01

The function and structures of single containment and double containment of PWR nuclear power plant were described briefiy. The dissimilarites of diffent type of containments, which effects the impact of environment are discused. The impact of environment, effected by 'source term', containment gas leak rate and diffusion pattern of the released gas, under different operating condition is analysed. Especially, the impact of environment under LOCA accident is fully analysed

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

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.)

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)

Numerical simulation of the heating and start-up of PWR nuclear power station

International Nuclear Information System (INIS)

Faraco-Medeiros, M.A.; Leite, C.A.T.; Ramalho, F.P.

1992-01-01

The start-up of a PWR nuclear power plant must be done within safety criteria and requires a simulation. The design of some equipment, cost and time can be optimized. A computer simulator, which allows control of all the equipment and variables into the operation, has been developed and is presented in this paper. The KWU procedure and an alternative for Angra II were simulated. The results are showed up. 09 refs, 03 figs. (B.C.A.)

Optimization of reload core design for PWR and application to Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Shen Wei; Zhongsheng Xie; Banghua Yin

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. 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

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)

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

International Nuclear Information System (INIS)

Llambias, J.M.

1993-01-01

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

Application of concrete filled steel bearing wall to inner concrete structure fro PWR nuclear power plant

International Nuclear Information System (INIS)

Sekimoto, Hisashi; Tanaka, Mamoru; Inoue, Kunio; Fukihara, Masaaki; Akiyama, Hiroshi.

1992-01-01

'Concrete filled steel bearing wall', applied to the inner concrete structure for PWR nuclear power plant, was developed for rationalization of construction procedure at site. It was concluded through preliminary studies that this new type of wall, where concrete is placed between steel plates, is best suited for the strength members of the above structure, due to the high strength and ductility of surface steel plates and the confinement effect of filled concrete. To verify the behavior from the elastic range to the inelastic range, the ultimate strength and the failure mechanism, and to clarify experimentally the structural integrity of the inner concrete structure, which was composed of a concrete filled steel bearing wall, against seismic lateral loads, horizontal loading tests using a 1/10th scale model of the inner concrete structure for PWR nuclear power plant were conducted. As a result of the tests, the inner concrete structure composed of a concrete filled steel bearing wall appeared to have a larger load carrying capacity and a higher ductility as compared with that composed of a reinforced concrete wall. (author)

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.)

Environmental surveillance of PWR power stations

International Nuclear Information System (INIS)

Conti, M.

1980-01-01

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

Siemens Nuclear Power Corporation experience with BWR and PWR fuels

International Nuclear Information System (INIS)

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

1992-01-01

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

Nuclear power

International Nuclear Information System (INIS)

1987-01-01

''Nuclear Power'' describes how a reactor works and examines the different designs including Magnox, AGR, RBMK and PWR. It charts the growth of nuclear generation in the world and its contributions to world energy resources. (author)

Modeling and simulation of pressurizer dynamic process in PWR nuclear power plant

International Nuclear Information System (INIS)

Ma Jin; Liu Changliang; Li Shu'na

2010-01-01

By analysis of the actual operating characteristics of pressurizer in pressurized water reactor (PWR) nuclear power plant and based on some reasonable simplification and basic assumptions, the quality and energy conservation equations about pressurizer' s steam zone and the liquid zone are set up. The purpose of this paper is to build a pressurizer model of two imbalance districts. Water level and pressure control system of pressurizer is formed though model encapsulation. Dynamic simulation curves of main parameters are also shown. At last, comparisons between the theoretical analysis and simulation results show that the pressurizer model of two imbalance districts is reasonable. (authors)

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

International Nuclear Information System (INIS)

Hoeld, A.

1975-01-01

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

A new approach to PWR power control using intelligent techniques

International Nuclear Information System (INIS)

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

2004-01-01

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

Simulating the steam generator and the pressurizer of a PWR nuclear power plant

International Nuclear Information System (INIS)

De Greef, J.F.

1985-01-01

In a PWR nuclear power plant, considered as a power generating device, the steam generator as a subset plays an important role in the generation process, whereas the pressurizer rather acts as a control device for security purposes. Nevertheless, from a thermodynamical point of view, the two subsets behave basically in the same way, so that a common set of basic equations may be suggested to develop for each the proper mathematical simulation model. In this paper the generation of this common set of basic equations is described, from which a specific model for each device is derived. A numerical illustration of the behaviour of the two devices for typical inputs to the derived simulation model is pictured. (author)

Automatic welding processes for reactor coolant pipes used in PWR type nuclear power plant

International Nuclear Information System (INIS)

Hamada, T.; Nakamura, A.; Nagura, Y.; Sakamoto, N.

1979-01-01

The authors developed automatic welding processes (submerged arc welding process and TIG welding process) for application to the welding of reactor coolant pipes which constitute the most important part of the PWR type nuclear power plant. Submerged arc welding process is suitable for flat position welding in which pipes can be rotated, while TIG welding process is suitable for all position welding. This paper gives an outline of the two processes and the results of tests performed using these processes. (author)

Liquid radioactive waste processing improvement of PWR nuclear power plants

International Nuclear Information System (INIS)

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

2005-01-01

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

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)

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

Strength analysis of refueling machine for large PWR in nuclear power plant

International Nuclear Information System (INIS)

Jia Xiaofeng; Zhou Guofeng; Bi Xiangjun; Ji Shunying

2010-01-01

The refueling machine of PWR plays important roles in nuclear power plant operation,and the dynamic analysis and strength assessment should be carried out to check its safety. In this paper, the finite element model (FEM) was established with the software ANSYS 12 for the refueling machine structure of large 1 000 MW PWR. The dynamic computations were performed under three work conditions, i.e. normal (cart starting and braking), abnormal (OBE) and accident(SSE) conditions, respectively. The structure responses (internal force and stress) of refueling machine under earthquake response spectrum in three directions were combined with the method of square root of square sum (SRSS). Moreover, the static response under gravity was also considered to construct the most critical conditions. With the simulated results, the strength of main structure, bold and weld joint,and the stability of landing leg for additional crane were assessed based on the RCCM code. At last, the local stress analysis of finger-form hook, which function is to take fuel assemblies, was also analyzed, while its strength was also assessed. The results show that the strengths of the refueling machine under various working conditions can meet the safety requirements. (authors)

Liquid radioactive waste processing improvement of PWR nuclear power plants; Melhorias no processamento de rejeitos liquidos radioativos de usinas nucleares PWR

Energy Technology Data Exchange (ETDEWEB)

Nery, Renata Wolter dos Reis; Martinez, Aquilino Senra; Monteiro, Jose Luiz Fontes [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: wolter@eletronuclear.gov.br; monteiro@peq.coppe.ufrj.br; aquilinosenra@lmp.ufrj.br

2005-07-01

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

Investigation of modeling and simulation on a PWR power conversion system with RELAP5

International Nuclear Information System (INIS)

Rui Gao; Yang Yanhua; Lin Meng; Yuan Minghao; Xie Zhengrui

2007-01-01

Based on the power conversion system of nuclear and conventional islands of Dayabay nuclear power station, this paper models the thermal-hydraulic systems for PWR by using the best-estimate program, RELAP5. To simulate the full-scope power conversion system, not only the reactor coolant system (RCP) of nuclear island, but also the main steam system (VVP), turbine steam and drain system (GPV), bypass system (GCT), feedwater system (FW), condensate extraction system (CEX), moisture separator reheater system (GSS), turbine-driven feedwater pump (APP), low-pressure and high-pressure feedwater heater systems (ABP and AHP) of conventional island are considered and modeled. A comparison between the simulated results and the actual data of reactor under full-power demonstrates a fine match for Dayabay, and also manifests the feasibility in simulating full-scope power conversion system of PWR with RELAP5. (author)

Power ramp testing method for PWR fuel rod at research reactor

International Nuclear Information System (INIS)

Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

2003-01-01

A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

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.)

Chemical and radiochemical specifications - PWR power plants

International Nuclear Information System (INIS)

Stutzmann, A.

1997-01-01

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

Pre design processing of waste of ex-resin without materials matrix from nuclear power plant type PWR 1000 MW

International Nuclear Information System (INIS)

Cerdas Tarigan

2010-01-01

Have been done pre design processing of waste ex-resin without capacities matrix materials from nuclear power plant type PWR 1000 MW During the time radioactive waste of ex-resin processed to use process of immobilization use matrix materials like mixture cement and epoxy resin and then conditioning. This process is not effective and efficient because end result volume of end product bigger than volume early operation system and maintenance of its installation more difficult. To overcome this created a design of technology processing of waste of ex- resin without matrix materials through process of strainer, drying and conditioning represent technological innovation newly processing of radioactive waste of ex-resin. Besides this process more effective and efficient, volume of end product waste much more small from volume early and operation system and maintenance of its easier installation. Pre design is expected to be used as a basis to make conceptual of pre design installation of strainer, drying and conditioning for the processing of waste of ex-resin from nuclear power plant type PWR 1000 MW. (author)

Periodical inspection in nuclear power stations

International Nuclear Information System (INIS)

1986-01-01

Periodical inspection is presently being made of eight nuclear power plants in nuclear power stations. Up to the present time, in three of them, failures as follows have been observed. (1) Unit 3 (PWR) of the Mihama Power Station in The Kansai Electric Power Co., Inc. Nineteen heat-transfer tubes of the steam generators were plugged up due to failure. A fuel assembly with a failed spring fixture and in another the control-rod cluster with a failed control rod fixture were replaced. (2) Unit 2 (PWR) of the Oi Power Station in The Kansai Electric Power Co., Inc. Eight heat-transfer tubes of the heat exchangers were plugged up due to failure. (3) Unit 6 (BWR) of the Fukushima Nuclear Power Station I in The Tokyo Electric Power Co., Inc. A fuel assembly with leakage was replaced. (Mori, K.)

Horizontal loading test by whole model specimen simulating inner concrete structure of PWR type nuclear power plant

International Nuclear Information System (INIS)

Furuya, Noriyuki; Sekine, Masataka; Kimura, Kozo; Yamaguchi, Yoshihiro; Yamaguchi, Tsuneo; Takeda, Toshikazu

1985-01-01

The Nuclear Power Engineering Test Center has performed a horizontal loading test by a whole model specimen simulating the inner concrete structure of a PWR type nuclear power plant in order to investigate restoring force characteristics of reactor buildings. This report describes the results of examination of applicability to the test results of analysis methods based on elastic theory. The analysis results of elastic stiffness, concrete cracking load, rebar yielding loads and ultimate strength were compared with the test results. According to this examination, it is recognized that even these analysis methods based on elastic theory are comparatively effective for analysis of an inner concrete structure of fairly complex configuration, although there are limits of the scope of applicability. (author)

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

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

Model-based fault detection and isolation of a PWR nuclear power plant using neural networks

International Nuclear Information System (INIS)

Far, R.R.; Davilu, H.; Lucas, C.

2008-01-01

The proper and timely fault detection and isolation of industrial plant is of premier importance to guarantee the safe and reliable operation of industrial plants. The paper presents application of a neural networks-based scheme for fault detection and isolation, for the pressurizer of a PWR nuclear power plant. The scheme is constituted by 2 components: residual generation and fault isolation. The first component generates residuals via the discrepancy between measurements coming from the plant and a nominal model. The neutral network estimator is trained with healthy data collected from a full-scale simulator. For the second component detection thresholds are used to encode the residuals as bipolar vectors which represent fault patterns. These patterns are stored in an associative memory based on a recurrent neutral network. The proposed fault diagnosis tool is evaluated on-line via a full-scale simulator detected and isolate the main faults appearing in the pressurizer of a PWR. (orig.)

Electricity supplies in a French nuclear power station

International Nuclear Information System (INIS)

2011-01-01

As the operation of a nuclear power station requires a power supply system enabling this operation as well as the installation safety, this document describes how such systems are designed in the different French nuclear power stations to meet the requirements during a normal operation (when the station produces electricity) or when it is stopped, but also to ensure power supply to equipment ensuring safety functions during an incident or an accident occurring on the installation. More precisely, these safety functions are provided by two independent systems in the French nuclear power stations. Their operation is briefly described. Two different types of nuclear reactors are addressed: pressurised water reactors (PWR) of second generation, EPR (or PWR of third generation)

Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

International Nuclear Information System (INIS)

Wati, Nurokhim

2008-01-01

Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

Seismic analysis with FEM for fuel transfer system of PWR nuclear power plant

International Nuclear Information System (INIS)

Jia Xiaofeng; Liu Pengliang; Bi Xiangjun; Ji Shunying

2012-01-01

In the PWR nuclear power plant, the function of the fuel transfer system (FTS) is to transfer the fuel assembly between the reactor building and the fuel building. The seismic analysis of the transfer system structure should be carried out to ensure the safety under OBE and SSE. Therefore, the ANASYS 12.0 software is adopted to construct the finite element analysis model for the fuel transfer system in a million kilowatt nuclear power plant. For the various configurations of FTS in the operating process, the stresses of the main structures, such as the transfer tube, fuel assembly container, fuel conveyor car, lifting frame in the reactor building, lifting frame in the fuel building, support and guide structure of conveyor car and the lifting frame in both buildings, are computed. The stresses are combined with the method of square root of square sum (SRSS) and assessed under various seismic conditions based on RCCM code, the results of the assessment satisfy the code. The results show that the stresses of the fuel transfer system structure meet the strength requirement, meanwhile, it can withstand the earthquake well. (authors)

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

International Nuclear Information System (INIS)

Zhang Weiguo; Gao Fengqin; Zhou Hongyi

1992-03-01

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

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

International Nuclear Information System (INIS)

Zhang Weiguo; Gao Fengqin; Zhou Hongyi

1993-01-01

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

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)

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)

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.)

Power generation by nuclear power plants

International Nuclear Information System (INIS)

Bacher, P.

2004-01-01

Nuclear power plays an important role in the world, European (33%) and French (75%) power generation. This article aims at presenting in a synthetic way the main reactor types with their respective advantages with respect to the objectives foreseen (power generation, resources valorization, waste management). It makes a fast review of 50 years of nuclear development, thanks to which the nuclear industry has become one of the safest and less environmentally harmful industry which allows to produce low cost electricity: 1 - simplified description of a nuclear power generation plant: nuclear reactor, heat transfer system, power generation system, interface with the power distribution grid; 2 - first historical developments of nuclear power; 3 - industrial development and experience feedback (1965-1995): water reactors (PWR, BWR, Candu), RBMK, fast neutron reactors, high temperature demonstration reactors, costs of industrial reactors; 4 - service life of nuclear power plants and replacement: technical, regulatory and economical lifetime, problems linked with the replacement; 5 - conclusion. (J.S.)

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

International Nuclear Information System (INIS)

Huang Xueqing; Zhang Senru

1999-01-01

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

HVDC transmission from nuclear power plant

International Nuclear Information System (INIS)

Yoshida, Yukio; Takenaka, Kiyoshi; Taniguchi, Haruto; Ueda, Kiyotaka

1980-01-01

HVDC transmission directly from a nuclear power plant is expected as one of the bulk power transmission systems from distant power generating area. Successively from the analysis of HVDC transmission from BWR-type nuclear power plant, this report discusses dynamic response characteristics of HVDC transmission (double poles, two circuits) from PWR type nuclear power plant due to dc-line faults (DC-1LG, 2LG) and ac-line faults (3LG) near inverter station. (author)

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

International Nuclear Information System (INIS)

2003-11-01

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

The capital investment and electricity cost of 2 x 600 MW PWR nuclear power plant in China

International Nuclear Information System (INIS)

Li Zhihua; Xing Leiming

1990-01-01

The capital investment and electricity cost of 2 x 600 MW PWR nuclear power plant in China are studied. If the rate of interest R 1 and of escalation R 2 are 7.2% and 10.0% respectively for RMB and the rate of interest R 1 and of escalation R 2 are 6.5% and 2.0% respectively for MK, the total investment is 9270 M RMB Yuan, the Specific investment is 7320 RMB Yuan/kW, the average selling electricity cost is 0.16 RMB Yuan/(kW·h). If the selling electricity price is 0.24 RMB Yuan/(kW·h), the rate of inner return is 7.7%, the dynamic return period is 13 years, the national income is 15800 M RMB Yuan, the profit of nuclear power plant after taxation is 6800 M RMB Yuan

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)

Effect of operating conditions and environment on properties of materials of PWR type nuclear power plant components

International Nuclear Information System (INIS)

Vacek, M.

1987-01-01

Operating reliability and service life of PWR type nuclear power plants are discussed with respect to the material properties of the plant components. The effects of the operating environment on the material properties and the methods of their determination are characterized. Discussed are core materials, such as fuel, its cladding and regulating rod materials, and the materials of pipes, steam generators and condensers. The advances in the production of pressure vessel materials and their degradation during operation are treated in great detail. (Z.M.)

Training and qualification of nuclear power plant operators

International Nuclear Information System (INIS)

Ohsuga, Y.

2008-01-01

Based on training experiences of the nuclear power plant operators of pressurized water reactors (PWR) at the Nuclear Power Training Center Ltd. (NTC) in Japan, training programs were reviewed referring to US training programs. A systematic approach is deployed to them, which mainly consist of on-the-job training and the NTC training courses to meet the needs of all operators from beginners to experienced veterans according to their experiences and objectives. The NTC training is conducted using the simulators that simulate the nuclear power plant dynamics through the use of computers. The operators trained at the NTC work in the central control room of every PWR power plant. The NTC also carries out the qualification examinations for the shift managers. (T. Tanaka)

China’s Nuclear Power Plants in Operation

Institute of Scientific and Technical Information of China (English)

无

2011-01-01

Qinshan Plant Phase I Located in Haiyan,Zhejiang Province,Qinshan Nuclear Power Plant Phase I is t he first 300-megawatt pressurized water reactor (PWR) nuclear power plant independently designed,constructed,operated and managed by China.The plant came into commercial operation in April 1994.

Water cooled type nuclear power plant

International Nuclear Information System (INIS)

Arai, Shigeki.

1981-01-01

Purpose: To construct high efficiency a PWR type nuclear power plant with a simple structure by preparing high temperature and pressure water by a PWR type nuclear reactor and a pressurizer, converting the high temperature and high pressure water into steam with a pressure reducing valve and introducing the steam into a turbine, thereby generating electricity. Constitution: A pressurizer is connected downstream of a PWR type nuclear reactor, thereby maintaining the reactor at high pressure. A pressure-reducing valve is provided downstream of the pressurizer, the high temperature and pressure water is reduced in pressure, thereby producing steam. The steam is fed to a turbine, and electric power is generated by a generator connected to the turbine. The steam exhausted from the turbine is condensed by a condenser into water, and the water is returned through a feedwater heater to the reactor. Since the high temperature and pressure water in thus reduced in pressure thereby evaporating it, the steam can be more efficiently produced than by a steam generator. (Sekiya, K.)

Optimization of the decontamination in EDF PWR power plants

International Nuclear Information System (INIS)

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

2002-01-01

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

Lethal aspects of nuclear power

International Nuclear Information System (INIS)

Callar, Achiles del

1986-01-01

This paper presents the author's researches on the disadvantages of nuclear power plants and particularly cites specific instances of PWR reactors that malfunctioned and caused the release of deadly radioactive effluents. (ELC)

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.)

HVDC transmission from isorated nuclear power plant

International Nuclear Information System (INIS)

Takenaka, Kiyoshi; Takasaki, Masahiro; Ichikawa, Tatemi; Hayashi, Toshiyuki

1985-01-01

HVDC transmission directly from nuclear power plant is considered as one of the patterns of long distance and large capacity transmission system. This reports considers two route HVDC transmission from PWR type nuclear power plant, and analyzes dynamic response characteristics due to bus fault, main protection failure and etc. using the AC-DC Power System Simulator. (author)

The reliability data acquisition system in PWR nuclear power plants

International Nuclear Information System (INIS)

Lienart, P.

1984-01-01

In April 1978, Electricite de France put a reliability data acquisition system (SRDF) into operation at its two nuclear power plant sites: Fessenheim and Bugey. In the light of the experience acquired and the advantages offered by such a data bank, this system has been progressively extended since 1982 to cover the entire PWR network. The SRDF was originally designed for the follow-up of 4000 items of equipment per pair of units. However, the various difficulties encountered in gathering data made it necessary - in order to safeguard the quality of the information - to reduce this number initially to 800 major mechanical or electromechanical items of equipment designed to ensure the safety or availability of the units. Subsequently, an increase to 1100 was possible. The SRDF consists of a centralized information bank linked by telephone to the various nuclear sites. The software enables the data-acquisition cards to be introduced, modified or deleted. Any user can gain access to the bank by simply making queries in real time. The quality of the acquisition and processing of the data depend on a list of equipment confined to essential operational systems and on a card design combining, as far as possible, the precision and accessibility of the data. A method of logical failure analysis has also been devised, its main purposes being to provide the following: (1) aid to card instruction; (2) an easier way of checking the uniformity of information concerning a failure; and (3) compatibility between the instructions and analysis of data, thereby facilitating development of the data-processing program. (author)

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)

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

International Nuclear Information System (INIS)

Martin Lopez, L.A.N.

1989-01-01

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

Modeling of hydrogen behaviour in a PWR nuclear power plant containment with the CONTAIN code

International Nuclear Information System (INIS)

Bobovnik, G.; Kljenak, I.

2001-01-01

Hydrogen behavior in the containment during a severe accident in a two-loop Westinghouse-type PWR nuclear power plant was simulated with the CONTAIN code. The accident was initiated with a cold-leg break of the reactor coolant system in a steam generator compartment. In the input model, the containment is represented with 34 cells. Beside hydrogen concentration, the containment atmosphere temperature and pressure and the carbon monoxide concentration were observed as well. Simulations were carried out for two different scenarios: with and without successful actuation of the containment spray system. The highest hydrogen concentration occurs in the containment dome and near the hydrogen release location in the early stages of the accident. Containment sprays do not have a significant effect on hydrogen stratification.(author)

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

International Nuclear Information System (INIS)

1979-01-01

The results of safety analysis on PWR type nuclear power plants with two and three loops are presented, which was conducted by the Resource and Energy Agency, in June, 1979. This analysis was made simulating the phenomenon relating to the pressurizer level gauge at the time of the TMI accident. The model plants were the Ikata nuclear power plant with two loops and the Takahama No. 1 nuclear power plant with three loops. The premise conditions for this safety analysis were as follows: 1) the main feed water flow is totally lost suddenly at the full power operation of the plants, and the feed water pump is started manually 15 minutes after the accident initiation, 2) the relief valve on the pressurizer is kept open even after the pressure drop in the primary cooling system, and the primary cooling water flows out into the containment vessel through the rupture disc of the pressurizer relief tank, and 3) the electric circuit, which sends out the signal of safety injection at the abnormal low pressure in the reactor vessel, is added from the view-point of starting the operation of the emergency core cooling system as early as possible. Relating to the analytical results, the pressure in the reactor vessels changes less, the water level in the pressurizers can be regulated, and the water level in the steam generators is recovered safely in both two and three-loop plants. It is recognized that the plants with both two- and three loops show the safe transient phenomena, and the integrity of the cores is kept under the premise conditions. The evaluation for each analyzed result was conducted in detail. (Nakai, Y.)

LWR nuclear power plant component failures

International Nuclear Information System (INIS)

Schmidt, W.H.

1980-10-01

An analysis of the most significant light water reactor (LWR) nuclear power plant component failures, from information in the computerized Nuclear Safety Information Center (NSIC) data bank, shows that for both pressurized water reactor (PWR) and boiling water reactor (BWR) plants the component category most responsible for reactor shutdowns is valves. Next in importance for PWR shutdowns is steam generators followed by seals of all kinds. For BWR plants, seals, and pipes and pipe fittings are the second and third most important component failure categories which lead to reactor shutdown. The data are for records extending from early 1972 through September 1978. A list of the most significant component categories and a breakdown of the number of component citations for both PWR and BWR reactor types are presented

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

International Nuclear Information System (INIS)

Sadek, Sinisa; Amizic, Milan; Grgic, Davor

2013-01-01

The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

Operation results of the secondary circuits of the French PWR type power plant park

International Nuclear Information System (INIS)

Mercier, J.P.

1984-01-01

Global results of performances realized since 1981 by the French PWR 900 MW power plants (installed power, availability, casual or planned shutdowns); analysis of the behaviour (casual unavailability) comparing together the performances of the different components in the secondary circuit; behaviour of the principal materials of the secondary circuit and their weight in the unavailabilities of the whole French nuclear park [fr

RENAP-MP: national PWR - medium nuclear power plant

International Nuclear Information System (INIS)

Meldonian, N.L.; Yamaguchi, M.; Kosaka, N.; Moreira, J.M.L.

1994-01-01

In this work are presented the main characteristics of a medium nuclear power plant, named RENAP - MP which has been developed by the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN - SP) and Coordenadoria para Projetos Especiais (COPESP) from de Brazilian Navy. The study aims at determining the necessary conditions for such a plant complete, in Brazil, technically and economically with other means of thermal power generation, such as coal, gas and oil. Results show that the energy average cost should be around 59 dollars/MWh; the construction time should be less than 5 years in order to reduce financial costs; and the design should emphasize the use of inherently safe technology, in order to reduce substantially the risks of nuclear accidents. It should be pointed out that the nuclear power technology has some environmental and social advantages when compared to hydro and other thermal power technologies, since it does not require flooding of large areas, reduces green house and acid rain effects, and can be located near consumption areas. (author). 11 refs, 5 tabs

Reactors of different types in the world nuclear power

International Nuclear Information System (INIS)

Simonov, K.V.

1991-01-01

The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

Nuclear power plant equipment design and construction rules

International Nuclear Information System (INIS)

Boiron, P.

1983-03-01

Presentation of the AFCEN (French association for nuclear power plant equipment design and construction rules) working, of its edition activity and of somes of its edited documents such as RCC-C (design and construction rules for PWR power plant fuel assemblies) and RCC-E (design and construction rules for nuclear facility electrical equipments) [fr

78 FR 4477 - Review of Safety Analysis Reports for Nuclear Power Plants, Introduction

Science.gov (United States)

2013-01-22

... Analysis Reports for Nuclear Power Plants: LWR Edition.'' The new subsection is the Standard Review Plan... Nuclear Power Plants: Integral Pressurized Water Reactor (iPWR) Edition.'' DATES: Comments must be filed... NUCLEAR REGULATORY COMMISSION [NRC-2012-0268] Review of Safety Analysis Reports for Nuclear Power...

Development of nuclear power in Sweden

International Nuclear Information System (INIS)

Sokolowski, E.

1994-01-01

The history and development of nuclear energetics in Sweden has been described. Up to now 12 units (9 BWR and 3 PWR nuclear reactors) with total 70 TWh electric power is being working in Sweden. If the same power will be installed as a coal-fired units the emissions of 65 million tons of CO 2 , 240,000 tons of SO 2 and 140 tons of NO x would be expected. 1 tab

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

International Nuclear Information System (INIS)

Silva, Alonso F.O.

2007-01-01

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

Aging effects in PWR power plants components

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

Nuclear power plant diagnostic system

International Nuclear Information System (INIS)

Prokop, K.; Volavy, J.

1982-01-01

Basic information is presented on diagnostic systems used at nuclear power plants with PWR reactors. They include systems used at the Novovoronezh nuclear power plant in the USSR, at the Nord power plant in the GDR, the system developed at the Hungarian VEIKI institute, the system used at the V-1 nuclear power plant at Jaslovske Bohunice in Czechoslovakia and systems of the Rockwell International company used in US nuclear power plants. These diagnostic systems are basically founded on monitoring vibrations and noise, loose parts, pressure pulsations, neutron noise, coolant leaks and acoustic emissions. The Rockwell International system represents a complex unit whose advantage is the on-line evaluation of signals which gives certain instructions for the given situation directly to the operator. The other described systems process signals using similar methods. Digitized signals only serve off-line computer analyses. (Z.M.)

Elecnuc. Nuclear power plants in the world

International Nuclear Information System (INIS)

1998-01-01

This small booklet summarizes in tables all the numerical data relative to the nuclear power plants worldwide. These data come from the French CEA/DSE/SEE Elecnuc database. The following aspects are reviewed: 1997 highlights; main characteristics of the reactor types in operation, under construction or on order; map of the French nuclear power plants; worldwide status of nuclear power plants at the end of 1997; nuclear power plants in operation, under construction and on order; capacity of nuclear power plants in operation; net and gross capacity of nuclear power plants on the grid and in commercial operation; forecasts; first power generation of nuclear origin per country, achieved or expected; performance indicator of PWR units in France; worldwide trend of the power generation indicator; nuclear power plants in operation, under construction, on order, planned, cancelled, shutdown, and exported; planning of steam generators replacement; MOX fuel program for plutonium recycling. (J.S.)

Risk of nuclear power generation as business (continued)

International Nuclear Information System (INIS)

Sato, Satoshi

2017-01-01

This paper described the following: (1) fleet formation of power companies that operate nuclear power plants in the U.S., (2) collaboration, competition, and merger between plant makers, (3) stress corrosion cracking of stream generators for PWR and their thin heat transfer tubes, especially stress corrosion cracking under primary cooling water environment (PWSCC), and (4) replacement project from Inconel 600 MA to Inconel 600 TT or 690 TT of steam generator thin heat transfer tubes of PWR plants in the U.S. and others. In addition, it described the troubles at San Onofre Nuclear Power Station in California: wear of steam generator thin tubes of Units 2 and 3, and leakage from primary system to secondary system of Unit 3, and permanent shutdown. It also described the detail of damages compensation talks between South California Edison Company that operates San Onofre nuclear power plant and Mitsubishi Heavy Industries Ltd. which supplied the steam generator. Although the operation of the 1.7 million kW plant became impossible due to the bud shedding of nuclear power renaissance, these troubles might have saved the nightmare of drifting on the way. (A.O.)

Nuclear power in Spain

International Nuclear Information System (INIS)

Koryakin, Yu.I.

1977-01-01

The present states of nuclear power in Spain is shortly surveyed. Data are provided on NPPs currently in operation, under construction, designed and planned. In line with the 10-year ''National programme of electricity supply'' a major and all increasing part of the electricity generation growth is to be ensured by NPPs and to account for more than 50% by the end of the period (1987). Out of the 7 units of NPPs now under construction, 6 units utilize PWR reactors and only 1 unit- a BWR reactor. The roles of private and public sectors are noted. Main characteristics of the ''ENSA'' plant now under construction are provided where components of NPPs with PWR and BWR reactors will be fabricated. Major developments in the fields of mining, milling and extraction of U from lignites, U enrichment, fuel fabrication and spent fuel reprocessing are considered. Measures now taken to improve the licensing procedure, surveillance of NPPs and personnel training are to advance the nuclear power development programme in the country

Nuclear power plants and the environment

Energy Technology Data Exchange (ETDEWEB)

Barabas, K [Ceskoslovenska Komise pro Atomovou Energii, Prague

1978-05-01

The environmental impacts are compared of conventional coal-fired and oil-fired power plants and of nuclear power plants. The values are compared of SO/sub 2/, NO/sub 2/, ash and soot emissions with /sup 133/Xe and /sup 85/Kr fission products release and the requirement for air for diluting these emissions in the atmosphere is assessed. Also compared are thermal pollution from an oil-fired power plant and from PWR and fast reactor power plants. The conclusion is arrived at that nuclear energy can solve the problem of increasing demand for electric and heat power while reducing negative environmental impacts.

Nuclear power plants and the environment

International Nuclear Information System (INIS)

Barabas, K.

1978-01-01

The environmental impacts are compared of conventional coal-fired and oil-fired power plants and of nuclear power plants. The values are compared of SO 2 , NO 2 , ash and soot emmisions with 133 Xe and 85 Kr fission products release and the requirement for air for diluting these emissions in the atmosphere is assessed. Also compared are thermal pollution from an oil-fired power plant and from PWR and fast reactor power plants. The conclusion is arrived at that nuclear energy can solve the problem of increasing demand for electric and heat power while reducing negative environmental impacts. (O.K.)

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)

Valves and fittings for nuclear power stations

International Nuclear Information System (INIS)

1976-01-01

The standard specifies technical requirements for valves and pipe fittings in nuclear power stations with PWR type reactors. Details of appropriate materials, welding, surface treatment for corrosion protection, painting, and complementary supply are given

State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

Energy Technology Data Exchange (ETDEWEB)

Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

2017-02-15

A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Introduction to nuclear power reactors and their health physics systems

International Nuclear Information System (INIS)

Brtis, J.S.

1982-01-01

This paper provides an introduction to: (1) the major systems of Boiling Water Reactors (BWR's) and Pressurized Water Reactors (PWR's), (2) the production and distribution of radiation sources in BWR's and PWR's, (3) the regulatory and functional requirements for nuclear power reactor design from a health physics standpoint, (4) the health physics systems provided to meet such requirements, and (5) a bibliography of documents germane to power reactor health physics design

Data list of nuclear power plants in Japan

International Nuclear Information System (INIS)

Izumi, Fumio; Nakamura, Jinichi

1982-10-01

This report has collected and compiled the data by December in 1981 concerning performances, equipments and installations of the nuclear power plants in Japan. The data have been modified according to the changes produced after previous publication of 1979 edition including BWR and PWR (JAERI-M 8947) and 1980 edition including PWR (JAERI-M 9629), and extended to cover the new plants developed thereafter. All data have been processed and tabulated with a data processing computer program FREP. Besides this report, user also can refer to 'Data List of Nuclear Power Plant in Japan' through terminals equipped at various places in JAERI using TSS (Time Shearing System) network of FACOM M-200, and the explanation of the usage is given in the Appendix. (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

The use of EDI to reduce the ammonia concentration in steam generators blowdown of PWR nuclear power plants

International Nuclear Information System (INIS)

Calay, J.C.; Goffin, C.

2000-01-01

To be recycled, PWR steam generator blowdown must be purified by mechanical filters, followed by ion exchangers (mixed bed preceded by a cationic ion exchange resin). The cationic ion exchange resin eliminates the conditioning agent ammonia in order to lengthen the cycles of the mixed bed. In the Doel nuclear power plant, Laborelec performed tests on a pilot plant for continuous electrodeionization that might replace the cation exchanger. The test campaign lasted six months. It is concluded that ammonia is removed well (1,000 μg/kg in the feed vs. 3 - 4 μg/kg in the product). The electrodeionization removes also other impurities; the conductivity of the treated water amounts to nearly 0.07 μs/cm

Nuclear power reactor safety research activities in CIAE

International Nuclear Information System (INIS)

Pu Shendi; Huang Yucai; Xu Hanming; Zhang Zhongyue

1994-01-01

The power reactor safety research activities in CIAE are briefly reviewed. The research work performed in 1980's and 1990's is mainly emphasised, which is closely related to the design, construction and licensing review of Qinshan Nuclear Power Plant and the safety review of Guangdong Nuclear Power Station. Major achievements in the area of thermohydraulics, nuclear fuel, probabilistic safety assessment and severe accident researches are summarized. The foreseeable research plan for the near future, relating to the design and construction of 600 MWe PWR NPP at Qinshan Site (phase II development) is outlined

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

Relation between source term and emergency planning for nuclear power plants

International Nuclear Information System (INIS)

Shi Zhongqi; Yang Ling

1992-01-01

Some background information of the severe accidents and source terms related to the nuclear power plant emergency planning are presented. The new source term information in NUREG-0956 and NUREG-1150, and possible changes in emergency planning requirements in U.S.A. are briefly provided. It is suggested that a principle is used in selecting source terms for establishing the emergency planning policy and a method is used in determining the Emergency Planning Zone (EPZ) size in China. Based on the research results of (1) EPZ size of PWR nuclear power plants being built in China, and (2) impact of reactor size and selected source terms on the EPZ size, it is concluded that the suggested principle and the method are suitable and feasible for PWR nuclear power plants in China

Elecnuc. Nuclear power plants in the world

International Nuclear Information System (INIS)

2000-01-01

This small booklet summarizes in tables all the numerical data relative to the nuclear power plants worldwide. These data come from the French CEA/DSE/SEE Elecnuc database. The following aspects are reviewed: 1999 highlights; main characteristics of the reactor types in operation, under construction or on order; map of the French nuclear power plants; worldwide status of nuclear power plants at the end of 1999; nuclear power plants in operation, under construction and on order; capacity of nuclear power plants in operation; net and gross capacity of nuclear power plants on the grid and in commercial operation; grid connection forecasts; world electric power market; electronuclear owners and share holders in EU, capacity and load factor; first power generation of nuclear origin per country, achieved or expected; performance indicator of PWR units in France; worldwide trend of the power generation indicator; 1999 gross load factor by operator; nuclear power plants in operation, under construction, on order, planned, cancelled, shutdown, and exported; planning of steam generators replacement; MOX fuel program for plutonium recycling. (J.S.)

Criteria for safety-related nuclear-power-plant operator actions: 1982 pressurized-water-reactor (PWR) simulator exercises

International Nuclear Information System (INIS)

Crowe, D.S.; Beare, A.N.; Kozinsky, E.J.; Haas, P.M.

1983-06-01

The primary objective of the Safety-Related Operator Action (SROA) Program at Oak Ridge National Laboratory is to provide a data base to support development of criteria for safety-related actions by nuclear power plant operators. When compared to field data collected on similar events, a base of operator performance data developed from the simulator experiments can then be used to establish safety-related operator action design evaluation criteria, evaluate the effects of performance shaping factors, and support safety/risk assessment analyses. This report presents data obtained from refresher training exercises conducted in a pressurized water reactor (PWR) power plant control room simulator. The 14 exercises were performed by 24 teams of licensed operators from one utility, and operator performance was recorded by an automatic Performance Measurement System. Data tapes were analyzed to extract operator response times (RTs) and error rate information. Demographic and subjective data were collected by means of brief questionnaires and analyzed in an attempt to evaluate the effects of selected performance shaping factors on operator performance

3D thermal-hydraulic analysis on core of PWR nuclear power station

International Nuclear Information System (INIS)

Yao Zhaohui; Wang Xuefang; Shen Mengyu

1997-01-01

Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

Utilization of nuclear energy for generating electric power in the FRG, with special regard to LWR-type reactors

International Nuclear Information System (INIS)

Vollradt, J.

1977-01-01

Comments on interdependencies in energy industry and energy generation as seen by energy supply utilities, stating that the generation of electric power in Germany can only be based on coal and nuclear energy in the long run, are followed by the most important, fundamental, nuclear-physical, technological and in part political interdependencies prevailing in the starting situation of 1955/58 when the construction of nuclear power plant reactors began. Then the development ranging to the 28000 MW nuclear power output to be expected in 1985 is outlined, totalling in 115000 MW electric power in the FRG. Finally, using the respectively latest order, the technical set up of each of the reactor types with 1300 MWe unit power offered by German manufacturers are described: BBC/BBR PWR-type reactor Neupotz, KWU-PWR-type reactor Hamm and KWU PWR-type reactor double unit B+C Gundremmingen. (orig.) [de

Elecnuc. Nuclear power plants in the world

International Nuclear Information System (INIS)

2003-01-01

This 2003 version of Elecnuc contents information, data and charts on the nuclear power plants in the world and general information on the national perspectives concerning the electric power industry. The following topics are presented: 2002 highlights; characteristics of main reactor types and on order; map of the French nuclear power plants; the worldwide status of nuclear power plants on 2002/12/3; units distributed by countries; nuclear power plants connected to the Grid by reactor type groups; nuclear power plants under construction; capacity of the nuclear power plants on the grid; first electric generations supplied by a nuclear unit; electrical generation from nuclear plants by country at the end 2002; performance indicator of french PWR units; trends of the generation indicator worldwide from 1960 to 2002; 2002 cumulative Load Factor by owners; nuclear power plants connected to the grid by countries; status of license renewal applications in Usa; nuclear power plants under construction; Shutdown nuclear power plants; exported nuclear power plants by type; exported nuclear power plants by countries; nuclear power plants under construction or order; steam generator replacements; recycling of Plutonium in LWR; projects of MOX fuel use in reactors; electricity needs of Germany, Belgium, Spain, Finland, United Kingdom; electricity indicators of the five countries. (A.L.B.)

Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

International Nuclear Information System (INIS)

Rubini, L.A.

1978-02-01

An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

Decommissioning of Swedish nuclear power reactors. Technology and costs

International Nuclear Information System (INIS)

1994-06-01

The main topics discussed are planning, technology and costs of decommissioning nuclear power reactors. Oskarshamn-3 (BWR) and Ringhals-4 (PWR) have been used as reference reactors. 29 refs, figs, tabs

Corrosion problems in nuclear power industry

International Nuclear Information System (INIS)

Flis, J.; Janik-Czachor, M.

1980-01-01

Characteristics were given of steels and alloys used in the PWR nuclear power plants and of water used in the primary and secondary systems. Corrosion damages of materials and installations were described. It was indicated that the damages were due mainly to stress corrosion cracking. Main preventive methods were listed. (author)

Pressurised water reactor. A critique of the Government's nuclear power programme

Energy Technology Data Exchange (ETDEWEB)

Flood, M; Chudleigh, R; Conroy, C

1981-01-01

The subject is covered in sections, entitled: energy forecasts - their decline and fall (Department of Energy; electricity supply industry; the writing on the wall); nuclear economics - how the dream dissolved (Britain's programme - off to a bad start; the government's programme -a catastrophe about to be repeated; the cost of a British PWR; a new choice - nuclear power versus insulation); PWR safety - the unresolvable problem (one in a million (query); fundamental weaknesses; consequences of an accident; unresolved safety problems; the select committee's findings); a new energy strategy - catering for our needs (energy efficiency; district heating/combined heat and power; solar heating; liquid fuels; the international context); conclusions.

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

Chemical process measurements in PWR-type nuclear power plants

International Nuclear Information System (INIS)

Glaeser, E.

1978-01-01

In order to achieve high levels of availability of nuclear power plants equipped with pressurized water reactors, strict standards have to be applied to the purity of coolant and of other media. Chemical process measurements can meet these requirements only if programmes are established giving maximum information with minimum expenditure and if these programmes are realized with effective analytical methods. Analysis programmes known from literature are proved for their usefulness, and hints are given for establishing rational programmes. Analytical techniques are compared with each other taking into consideration both methods which have already been introduced into nuclear power plant practice and methods not yet generally used in practice, such as atomic absorption spectrophotometry, gas chromatography, etc. Finally, based on the state of the art of chemical process measurements in nuclear power plants, the trends of future development are pointed out. (author)

Availability estimation of international nuclear power plants

International Nuclear Information System (INIS)

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

1978-11-01

Results are presented of investigation on the factors influencing the availability of nuclear power plants of the PWR type; an estimation of expected values for the availability factor and the probability of its having lower values than a certain specified value are given. (Author) [pt

Principle simulator for a PWR nuclear power station

International Nuclear Information System (INIS)

Wahlstroem, B.

1975-05-01

A report is given on a simulator developed for the training of operational and planning staff for the Lovisa nuclear power station in Finland. All main components of the power station are illustrated and trainees can operate the simulator in the power range 3-100 %. The model was originally developed for planning the control system of Lovisa I, for which reason the simulator project could be carried out on a relatively limited budget. (author)

Present state of nuclear power business in China

International Nuclear Information System (INIS)

Morokuzu, Muneo

2011-01-01

This article presented present state of nuclear power business in China based on latest information obtained at visit at nuclear power related facilities in December 2010. China Atomic Energy Authority (CAEA) promoted nuclear power, while National Nuclear Safety Administration (NNSA) was an independent regulatory body of nuclear power. Construction of nuclear power was promoted by three national nuclear engineering development corporations: China National Nuclear Corporation (CNNC), China Guangdon Nuclear Power Corporation (CGNPC) and State Nuclear Power Technology Corporation (SNPTC). In China, 13 nuclear power reactors were in operation and 27 under construction. Shortage of nuclear engineers became evident with rapid growth of nuclear power, which forced delay of nuclear power construction schedule. Future strategies of reactor type varied domestic, French and US ones respectively dependent on CNNC, CGNPC and SNPTC, CNNC seemed to change from third generation reactor (CNP 1000) to second one (CP 1000) due to regulatory licensing difficulty of NNSA. As for advanced reactor development, large scale PWR project, HTR project and FBR development project were proceeding. As HTR project was selected as high-priority project, an experimental reactor (HTR-10) was critical in 2000 and construction of demonstration reactor started in 2009. (T. Tanaka)

Nuclear safeguards control in nuclear power stations

International Nuclear Information System (INIS)

Boedege, R.; Braatz, U.; Heger, H.

1976-01-01

The execution of the Non-Proliferation Treaty (NPT) has initiated a third phase in the efforts taken to ensure peace by limiting the number of atomic powers. In this phase it is important, above all, to turn into workable systems the conditions imposed upon technology by the different provisions of the Verification Agreement of the NPT. This is achieved mainly by elaborating annexes to the Agreement specifically geared to certain model plants, typical representatives selected for LWR power stations being the plants at Garigliano, Italy (BWR), and Stade, Federal Republic of Germany (PWR). The surveillance measures taken to prevent any diversion of special nuclear material for purposes of nuclear weapons manufacture must be effective in achieving their specific objective and must not impede the circumspect management of operations of the plants concerned. A VDEW working party has studied the technical details of the planned surveillance measures in nuclear power stations in the Federal Republic of Germany and now presents a concept of material balancing by units which meets the conditions imposed by the inspection authority and could also be accepted by the operators of nuclear power stations. The concept provides for uninterrupted control of the material balance areas of the nuclear power stations concerned, allows continuous control of the whole nuclear fuel cycle, is based exclusively on existing methods and facilities, and can be implemented at low cost. (orig.) [de

Conference/debate on nuclear power. Press file

International Nuclear Information System (INIS)

1998-01-01

This press dossier presents in a digest way the advantages of nuclear power with respect to other energy sources. After a presentation of the worldwide energy resources (electric power demand, fossil-fuel reserves, renewable energies, environmental constraints), the dossier describes the functioning principle of a PWR reactor as a presentation of the French nuclear program (historical aspects, policy, organization of activities, trade, EPR project). In the last part, the sociological aspects of nuclear power are approached: environmental aspects (natural radioactivity, reactors safety, radioactive wastes, environment protection, carbon dioxide and energy production), and public health aspects (principles of radioactivity, dose effects, industrial and scientific applications of radionuclides, nuclear controversy, reactor accidents, legal aspects of radioprotection, safety standards and controls). (J.S.)

Elecnuc. Nuclear power plants in the world

International Nuclear Information System (INIS)

2005-01-01

This 2005 edition of the Elecnuc booklet summarizes in tables all numerical data relative to the nuclear power plants worldwide. These data come from the PRIS database managed by the IAEA. The following aspects are reviewed: 2004 highlights; main characteristics of reactor types; map of the French nuclear power plants on 2005/01/01; worldwide status of nuclear power plants at the end of 2004; units distributed by countries; nuclear power plants connected to the grid by reactor-type group; nuclear power plants under construction on 2004; evolution of nuclear power plant capacities connected to the grid; first electric generations supplied by a nuclear unit; electrical generation from nuclear power plants by country at the end 2004; performance indicator of PWR units in France; trend of the generation indicator worldwide; 2004 load factor by owners; units connected to the grid by countries at 12/31/2004; status of licence renewal applications in USA; nuclear power plants under construction at 12/31/2004; shutdown reactors; exported nuclear capacity in net MWe; exported and national nuclear capacity connected to the grid; exported nuclear power plants under construction or order; exported and national nuclear capacity under construction or order; recycling of plutonium in LWR; Mox licence plant projects; Appendix - historical development; acronyms, glossary

The Paks Nuclear Power Station

International Nuclear Information System (INIS)

Erdosi, N.; Szabo, L.

1978-01-01

As the first stage in the construction of the Paks Nuclear Power Station, two units of 440 MW(e) each will be built. They are operated with two coolant loops each. The reactor units are VVER 440 type water-moderated PWR type heterogeneous power reactors designed in the Soviet Union and manufactured in Czechoslovakia. Each unit operates two Soviet-made K-220-44 steam turbines and Hungarian-made generators of an effective output of 220 MW. The output of the transformer units - also of Hungarian made - is 270 MVA. The radiation protection system of the nuclear power station is described. Protection against system failures is accomplished by specially designed equipment and security measures especially within the primary circuit. Some data on the power station under construction are given. (R.P.)

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

Elecnuc. Nuclear power plants in the world. 1997

International Nuclear Information System (INIS)

Maubacq, F.; Tailland, C.

1997-04-01

This small booklet provides information about all type of nuclear power plants worldwide. It is based on the data taken from the CEA/DSE/SEE Elecnuc database. The content comprises: the 1996 highlights, the main characteristics of the different type of reactors in operation or under construction, the map of the French nuclear power plant sites, the worldwide status of nuclear power plants at the end of 1996, the nuclear power plants in operation, under construction or on order (by groups of reactor-types), the power capacity evolution of power plants in operation, the net and gross capacity of the power plants on the grid, the commercial operation and grid connection forecasts, the first achieved or expected power generation supplied by a nuclear reactor for each country and the power generation from nuclear reactors, the performance indicator of the PWR units in France, the trends of the power generation indicator worldwide, the nuclear power plants in operation, under construction, on order, planned, cancelled, decommissioned and exported worldwide, the schedule of steam generator replacements, and the MOX fuel plutonium recycling programme. (J.S.)

New generation nuclear power units of PWR type integral reactors

International Nuclear Information System (INIS)

Mitenkov, F.M.; Kurachen Kov, A.V.; Malamud, V.A.; Panov, Yu.K.; Runov, B.I.; Flerov, L.N.

1997-01-01

Design bases of new generation nuclear power units (nuclear power plants - NPP, nuclear co-generation plants - NCP, nuclear distract heating plants - NDHP), using integral type PWPS, developed in OKBM, Nizhny Novgorod and trends of design decisions optimization are considered in this report. The problems of diagnostics, servicing and repair of the integral reactor components in course of operation are discussed. The results of safety analysis, including the problems of several accident localization with postulated core melting and keeping corium in the reactor vessel and guard vessel are presented. Information on experimental substantiation of the suggested plant design decisions is presented. (author)

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

Dura Seal recommendations for nuclear power plants

International Nuclear Information System (INIS)

Anon.

1980-01-01

Reactor systems (BWR, PWR and Candu) are briefly reviewed with reference to the pumping services encountered in each system, to indicate the conditions imposed on mechanical seals for nuclear power plant liquid handling equipment. A description of the Dura Seals used in each service is included. (U.K.)

The plutonium recycle for PWR reactors from brazilian nuclear program

International Nuclear Information System (INIS)

Rubini, L.A.

1978-01-01

The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U 3 O 8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

Simulation of a pressurized-water nuclear power station

International Nuclear Information System (INIS)

Larminaux, Robert; Ourmann, Michel

1978-01-01

Faced with the large programme of fitting out PWR nuclear power stations, Electricite de France have undertaken a series of studies with a view to ensuring the best possible adaptation of the secondary part -particularly the feed water heating section- to the nuclear boiler. In order to undertake such studies it has been necessary to finalize simulation models of the entire power station. So as to verify the validity of the models, experiment-calculation comparisons were made during transient operating states recorded at the Ardennes power station as well as during starting up trials at the Tihange I power station [fr

The basic planning for the environmental relationship of improved nuclear power plant

International Nuclear Information System (INIS)

Choi, Seungho; Jung, Hoonseok; Lee, Dongheup

2003-01-01

The Improved Nuclear Power Plant (INPP) which design has been emphasized on the safety, technology and environmental relationship is about to be realized after hard works of design improvements, largely changes made from existing PWR nuclear power plants for three years. It will undoubtedly play a leading role of nuclear power generation industry as a main electric power source and secure a place as a successful issue between development and environment throughout nation-wide understanding and support. (author)

Nuclear power: 2004 world report - evaluation

International Nuclear Information System (INIS)

Anon.

2005-01-01

Last year, 2004, 441 nuclear power plants were available for power supply in 31 countries of the world. Nuclear generating capacity attained its highest level so far at an aggregate gross power of 385,854 MWe and an aggregate net power of 366,682 MWe, respectively. Nine different reactor lines are operated in commercial nuclear power plants. Light water reactors (PWR and BWR) again are in the lead with 362 plants. At year's end, 22 nuclear power plants with an aggregate gross power of 18,553 MWe and an aggregate net power, respectively, of 17,591 MWe were under construction in nine countries. Of these, twelve are light water reactors, nine are CANDU-type reactors, and one is a fast breeder reactor. So far, 104 commercial reactors with powers in excess of 5 MWe have been decommissioned in eighteen countries, most of them low-power prototype plants. 228 nuclear power plants of those in operation, i.e. slightly more than half, were commissioned in the 1980es. Nuclear power plant availabilities in terms of capacity and time again reached record levels. Capacity availability was 84.30%, availability in terms of time, 85.60%. The four nuclear power plants in Finland continue to be world champions in this respect with a cumulated average capacity availability of 90.30%. (orig.)

Application of fatigue monitoring system in PWR nuclear power plant

International Nuclear Information System (INIS)

Piao Lei

2014-01-01

Fatigue failure is one form of equipment failure of nuclear power plant, influencing equipment lifetime and lifetime extension. Fatigue monitoring system can track real thermal transient at fatigue sensitive components, establish a basis for fatigue analyses based on realistic operating loads, identify unexpected operational transients, optimize the plant behavior by improved operating modes, provide supporting data for lifetime management, enhance security of plant and reduce economical loss. Fatigue monitoring system has been applied in many plants and is required to be applied in Generation-III nuclear power plant. It is necessary to develop the fatigue monitoring system with independent intellectual property rights and improve the competitiveness of domestic Generation-III nuclear power technology. (author)

Neural network of Gaussian radial basis functions applied to the problem of identification of nuclear accidents in a PWR nuclear power plant

International Nuclear Information System (INIS)

Gomes, Carla Regina; Canedo Medeiros, Jose Antonio Carlos

2015-01-01

Highlights: • It is presented a new method based on Artificial Neural Network (ANN) developed to deal with accident identification in PWR nuclear power plants. • Obtained results have shown the efficiency of the referred technique. • Results obtained with this method are as good as or even better to similar optimization tools available in the literature. - Abstract: The task of monitoring a nuclear power plant consists on determining, continuously and in real time, the state of the plant’s systems in such a way to give indications of abnormalities to the operators and enable them to recognize anomalies in system behavior. The monitoring is based on readings of a large number of meters and alarm indicators which are located in the main control room of the facility. On the occurrence of a transient or of an accident on the nuclear power plant, even the most experienced operators can be confronted with conflicting indications due to the interactions between the various components of the plant systems; since a disturbance of a system can cause disturbances on another plant system, thus the operator may not be able to distinguish what is cause and what is the effect. This cognitive overload, to which operators are submitted, causes a difficulty in understanding clearly the indication of an abnormality in its initial phase of development and in taking the appropriate and immediate corrective actions to face the system failure. With this in mind, computerized monitoring systems based on artificial intelligence that could help the operators to detect and diagnose these failures have been devised and have been the subject of research. Among the techniques that can be used in such development, radial basis functions (RBFs) neural networks play an important role due to the fact that they are able to provide good approximations to functions of a finite number of real variables. This paper aims to present an application of a neural network of Gaussian radial basis

Knowledge of ageing phenomenons of materials used in the PWR power plants

International Nuclear Information System (INIS)

Vancon, D.; Meyzaud, Y.; Soulat, P.

1996-01-01

The nuclear power plants with PWR type reactors are planned to work during forty years and are the subject of studies aiming to check their integrity during all their life. The materials used to the fabrication of the components can be submitted different stress. The temperature, the mechanical constraints, the irradiation are examples of stress which can make the materials getting old. This text presents three themes: the ageing by irradiation, the thermal ageing and the corrosion, and their principle industrial consequences. (N.C.)

Judith Perera: 'Nuclear Power in Europe'-A review

International Nuclear Information System (INIS)

Pavel, Eugeniu

2000-01-01

The book published by Financial Time Energy, London 1999, presents the history of nuclear power and its current status in both Eastern and Western Europe. The book extends on 370 pages and is structured in 8 chapters, annexes, a glossary of terms and references. Romanian nuclear power is represented by the Cernavoda NPP Unit 1 commissioned in December 1996. Till September 1999 the production of Cernavoda NPP amounted up to 15,271 TWh with a internal consumption lower than 8%, which covered 10-14% of the total electric energy demand of Romania, at a 12 to 14 USD/MWh production cost. The statistical data show that: - with the highest nuclear energy quota appears US, namely with 28.1% of world nuclear power production; - in a classification on countries, Lithuania and France appear on the first place, namely with 77.2% and 75.8%, respectively, as share of nuclear energy in the total national energy production. A review of the reactor types mostly used world wide show that the PWR is the most used technology in France, Japan, USA and several other countries. The chapter two presents the performances of different reactor types among which AGR, BWR and PWR, the load factor of which appears to be the highest. The problems of fuel cycles, fuel conversion, enrichment and production are treated in Chapter 3. The following chapter presents the economic aspects of nuclear power, the energy costs being a pro in the public debates on the future of nuclear energy. A detailed analysis on the environmental impact of nuclear power production is given in Chapter 6. The situation of nuclear sector throughout Europe is shown in the following chapter while the last one is devoted to the main European power companies implied in the nuclear industry. In conclusion, the prospects of nuclear energy in Europe are reviewed by taking into account factors like: public acceptance, costs, competition, waste management, fuel market, safety environmental issues, in-service life, personnel

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

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)

Framatome advanced nuclear power-benefits for our clients from the new company

International Nuclear Information System (INIS)

Weber, P.

2001-01-01

Framatome ANP (Advanced Nuclear Power) merges the complementary strengths of two global nuclear industry leaders Framatome and Siemens - offering clients the best technological solutions for safe, reliable and economical plant performance. With a combined workforce of 13,300 skilled individuals, Framatome ANP is now the nuclear industry's leading supplier. Serving as Original Equipment Manufacturer (OEM) for more than 90 reactors that provide about 30% of the world's total installed nuclear power capacity, our experienced resources remain focused on the local needs of individual clients, wherever in the world they may be. The Company main business used to be turnkey construction of complete Nuclear Power plants, BWR and PWR capabilities, heavy equipment manufacturing, comprehensive I and C capabilities, and also expertise and knowledge of VVER. Framatome ANP will benefit in all of its fields of activity of the experience gained through Framatome and Siemens' collaboration on the next generation reactor, the EPR, as well as on steam generators replacements and or modernization of VVER. Framatome ANP nuclear fuel designs for both PWR and BWR plants provide innovative features and world-leading performance. Framatome ANP is organized according a matrix organization with: - 4 Business Groups (Project and Engineering, Service, Nuclear Fuel, Mechanical Equipment) - 3 Regional Divisions (Framatome Advanced Nuclear Power S.A.S., France; Framatome Advanced Nuclear Power GmbH, Germany; Framatome Advanced Nuclear Power Inc., USA). By 30th January 2001 Siemens Nuclear Power GmbH, founded in 2000 as successor of the Nuclear Division of Siemens Power Generation Group (KWU), was renamed to Framatome Advanced Nuclear Power GmbH forming the German part of the world wide acting company. Over the past 40 years 23 nuclear power plants all around the world - not only pressurized and boiling water reactors, but also two heavy-watermoderated reactors have been designed, constructed and

Nuclear power now and in the future

Energy Technology Data Exchange (ETDEWEB)

Collier, J G [Nuclear Electric (United Kingdom)

1991-08-01

The future of the nuclear industry in the United Kingdom is considered from the perspective of the new public sector utility, Nuclear Electric, set up to retain control of nuclear power stations on the privatization of the rest of the electricity supply industry. Two major objectives are the increased nuclear generation of electricity and the cutting of costs. These are discussed in terms of life extension programmes for the magnox reactors, improved performance of AGR reactors and expectations for the Sizewell B PWR station now under construction; waste management, reactor decommissioning and fuel-cycle costs are also considered. Economic, environmental and political criteria are outlined which need to be addressed in relation to the government's review of nuclear power in 1991. Because of the marginal economic advantages of nuclear power in the United Kingdom, it will be important to quantify the environmental and diversity benefits of this source. (UK).

The AMEBA PWR, a new concept in the technology of nuclear reactor safety

Energy Technology Data Exchange (ETDEWEB)

Novelli, A

2000-05-01

AMEBA is an Italian acronym which stands for 'alta moderazione e basso arricchimento' (high moderation and low enrichment). The AMEBA reactor is nothing more than a PWR which possesses very unusual values of both volumetric ratio moderator/fuel and U-235 enrichment of UO{sub 2}. The possibility is shown of the technical realisation of a nuclear power plant equipped with an AMEBA PWR reactor. Among the most enticing properties of AMEBA are the following: self-shut-down in any abnormal condition, elimination of all need for control rods and boric acid dissolution in the water, absolute impossibility of reaching values of reactivity greater than a fraction of a dollar, intrinsic subcriticality, attaining to several dollars, in non-operative condition when the water is at ambient temperature, normal operation with a very small-sized pressurizer, self-start-up.

The AMEBA PWR, a new concept in the technology of nuclear reactor safety

International Nuclear Information System (INIS)

Novelli, A.

2000-01-01

AMEBA is an Italian acronym which stands for 'alta moderazione e basso arricchimento' (high moderation and low enrichment). The AMEBA reactor is nothing more than a PWR which possesses very unusual values of both volumetric ratio moderator/fuel and U-235 enrichment of UO 2 . The possibility is shown of the technical realisation of a nuclear power plant equipped with an AMEBA PWR reactor. Among the most enticing properties of AMEBA are the following: self-shut-down in any abnormal condition, elimination of all need for control rods and boric acid dissolution in the water, absolute impossibility of reaching values of reactivity greater than a fraction of a dollar, intrinsic subcriticality, attaining to several dollars, in non-operative condition when the water is at ambient temperature, normal operation with a very small-sized pressurizer, self-start-up

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

Assessment and management of ageing of major nuclear power plant components important to safety: PWR pressure vessels. 2007 update

International Nuclear Information System (INIS)

2007-06-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that effective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) is one of the most important issues for plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. IAEA-TECDOC-1120 documented ageing assessment and management practices for pressurized water reactor (PWR) reactor pressure vessels (RPVs) that were current at the time of its finalization in 1997-1998. Safety significant operating events have occurred since the finalization of the TECDOC, e.g. primary water stress corrosion cracking (PWSCC) of Alloy 600 control rod drive mechanism (CRDM) penetrations and boric acid corrosion/wastage of RPV heads, which threatened the integrity of the RPV heads. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1120 should be updated by incorporating those new events and their countermeasures. The objective of this report is to update IAEA-TECDOC-1120 in order to provide current ageing management guidance for PWR RPVs to all involved in the operation and regulation of PWRs and thus to help ensure PWR RPV integrity in IAEA Member States throughout their entire service life

Analysis and planning of the utilization of nuclear power plants

International Nuclear Information System (INIS)

Skvarka, P.

1985-01-01

The utilization coefficient as one of the characteristics of availability of nuclear power plants and the operation results (like maximum power, block number, and electric energy generation) are investigated by different statistic methods for several nuclear power plants with PWR type reactors and compared with those of WWER 440-type reactors. By means of linear many-parameter regression analysis the utilization coefficient is studied in dependence on block power and time after reactor commissioning. Forecastings of mean utilization coefficients are presented for the power of WWER 1000-type reactors

Resfria - a computational routine for thermal-hydraulic analysis of a cooldown in the PWR

International Nuclear Information System (INIS)

Silva Neto, A.J. da; Maciel Filho, L.A.

1989-01-01

This paper presents the computer code RESFRIA, designed to calculate the process parameters in a PWR nuclear power plant during a cooldown normal procedure. The procedure is described and some of the models developed to the simulation of systems and equipments are presented. A simplified flowchart of the computational routine and the results in the form of a diagram, for a typical PWR nuclear power plant, are also presented. (author)

The future for nuclear power

International Nuclear Information System (INIS)

Marshall of Goring, Lord.

1989-01-01

Lord Marshall explains how the situation for nuclear power in late 1989 in the United Kingdom had come about. Despite warnings that for a successful nuclear programme a large generator which has the obligation to supply in a defined geographical area should operate, this is not what will happen under the plans to privatise the electricity supply industry in the UK. Under these no body will have the obligation to supply electricity and the distribution Company will not have the obligation to supply after the first few years. Other problems with the privatisation plans are discussed. The implications of the government's decisions on nuclear power, first to maintain the Magnox stations in the government sector, second to abandon the full PWR construction programme and thirdly not to transfer the AGRs to the private sector, are discussed. (UK)

Improving automated load flexibility of nuclear power plants with ALFC

Energy Technology Data Exchange (ETDEWEB)

Kuhn, Andreas [AREVA GmbH, Karlstein (Germany). Plant Control/Training; Klaus, Peter [E.ON NPP Isar 2, Essenbach (Germany). Plant Operation/Production Engineering

2016-07-01

In several German and Swiss Nuclear Power Plants with Pressurized Water Reactor (PWR) the control of the reactor power was and will be improved in order to be able to support the energy transition with increasing volatile renewable energy in the grid by flexible load operation according to the need of the load dispatcher (power system stability). Especially regarding the mentioned German NPPs with a nominal electric power of approx. 1,500 MW, the general objectives are the main automated grid relevant operation modes. The new possibilities of digital I and C (as TELEPERM {sup registered} XS) enable the automation of the operating modes provided that manual support is no longer necessary. These possibilities were and will be implemented by AREVA within the ALFC-projects. Manifold adaption algorithms to the reactor physical variations during the nuclear load cycle enable a precise control of the axial power density distribution and of the reactivity management in the reactor core. Finally this is the basis for a highly automated load flexibility with the parallel respect and surveillance of the operational limits of a PWR.

Improving automated load flexibility of nuclear power plants with ALFC

Energy Technology Data Exchange (ETDEWEB)

Kuhn, Andreas [AREVA GmbH, Karlstein (Germany). Section Manager Training; Klaus, Peter [Preussenelektra NPP, Essenbach (Germany). Production Engineering

2017-03-15

In several German and Swiss Nuclear Power Plants with Pressurized Water Reactor (PWR) the control of the reactor power was and will be improved in order to be able to support the energy transition with increasing volatile renewable energy in the grid by flexible load operation according to the need of the load dispatcher (power system stability). Especially regarding the mentioned German NPPs with a nominal electric power of approx. 1500 MW, the general objectives are several automated grid relevant operation modes. The new possibilities of digital I and C (as TELEPERM {sup registered} XS) enable the automation of this operating modes provided that manual support is no longer necessary. These possibilities were and will be implemented by AREVA within the ALFC-projects. Manifold adaption algorithms to the reactor physical variations during the nuclear load cycle enable a precise control of the axial power density distribution and of the reactivity manage - ment in the reactor core. Finally this is the basis for a highly automated load flexibility with the parallel respect and surveillance of the operational limits of a PWR.

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

Problems of nuclear power plant safety evaluation

International Nuclear Information System (INIS)

Suchomel, J.

1977-01-01

Nuclear power plant safety is discussed with regard to external effects on the containment and to the human factor. As for external effects, attention is focused on shock waves which may be due to explosions or accidents in flammable material transport and storage, to missiles, and to earthquake effects. The criteria for evaluating nuclear power plant safety in different countries are shown. Factors are discussed affecting the reliability of man with regard to his behaviour in a loss-of-coolant accident in the power plant. Different types of PWR containments and their functions are analyzed, mainly in case of accident. Views are discussed on the role of destructive accidents in the overall evaluation of fast reactor safety. Experiences are summed up gained with the operation of WWER reactors with respect to the environmental impact of the nuclear power plants. (Z.M.)

ELECNUC. Nuclear power plants in the world - 2014 edition. Status on 12-31-2013, Draft using the IAEA's PRIS database and specific I-tese studies

International Nuclear Information System (INIS)

2014-01-01

Japan situation has not significantly changed since 2012, almost all its nuclear power plants are still stopped. The process for restarting 48 of them is under way but expected to take some years. As last year no order was registered, but 2013 met an important number of building starts concerning 10 new PWR (11.3 GWe): 1 - Belarus: First building site in eastern Europe (excluded Russia) for more than 25 years. 2 - USA: Four PWR units (4.5 GWe) that represent the first new buildings in this country since 1977. 3 - China: Three new PWR, that leads to a total of 27 plants under construction in this country. 4 - United Arab Emirates and South Korea: The sister-ships of Barakah-2 and Shin-Hanul-2 started right after the No.1 units in 2012. These 10 new plants are all PWR units, which is coherent with the worldwide situation of 58 PWR constructions for only 4 PHWR in India (and 4 BWR - 2 in Japan and 2 in Taiwan - all delayed for different reasons). Worldwide capacity balance Despite the commercial start of 3 GWe (2 PWR in China, 1 in Iran), the planned shutdown of 6 units (5.4 GWe: Fukushima-Daishi 5 and 6 in Japan and 4 PWR in USA) leads to a decrease of -2.4 GWe of the nuclear capacity. Thanks to a better average load factor the overall nuclear electricity production still increases of 13 TWh. But as 4.1 GWe (3 PWR in China, 1 in India) were connected to the grid and the constructions represent more than 50 units with about 10 annual commercial operations for the next four years, this loss of capacity should be exceptional. Contents: 1) 2013 highlights; 2) Main characteristics of reactor types; 3) Map of the French nuclear power plants on 01/01/2013; 4) Worldwide status of nuclear power plants (12/31/2013); 5) Units distributed by countries; 6) Nuclear power plants connected to the Grid- by reactor type groups (12/31/2013); 7) Nuclear power plants under construction on 2013; 8) Evolution of nuclear power plants capacities connected to the grid; 9) First electric

Risk evaluation for motor operated valves in an Inservice Testing Program at a PWR nuclear power plant in Taiwan

International Nuclear Information System (INIS)

Li, Y.C.; Chen, K.T.; Su, Y.L.; Ting, K.; Chien, F.T.; Li, G.D.; Huang, S.H.

2012-01-01

Safety related valves such as Motor Operated Valves (MOV), Air Operated Valves (AOV) or Check Valves (CV) play an important role in nuclear power plant. Functioning of these valves mainly aim at emergency reactivity control, post-accident residue heat removal, post-accident radioactivity removal and containment isolation when a design basis accident occurred. In order to maintain these valves under operable conditions, an Inservice Testing Program (IST) is defined for routine testing tasks based on the ASME Boiler and Pressure Vessel Code section XI code requirements. Risk based Inservice Testing Programs have been studied and developed extensively in the nuclear energy industry since the 1990s. Risk Based evaluations of IST can bring positive advantages to the licensee such as identifying the vulnerability of the system, reducing unnecessary testing burden, concentrating testing resources on the critical pass oriented valves and saving plant’s personnel dose exposure. This risk evaluation is incorporated with quantitative and qualitative analyses to the Motor Operated Valves under current Inservice Testing Program for PWR nuclear power plant in Taiwan. With the outcome of the risk classifications for the safety related MOVs through numerical or deterministic analyses, a risk based testing frequency relief is suggested to demonstrate the benefits received from the risk based Inservice Testing Program. The goal made of this study, it could be as a reference and cornerstone for the licensee to perform overall scope Risk-Informed Inservice Testing Program (RI-IST) evaluation by referring relevant methodologies established in this study.

Projects of Modifications of design for mitigation of accidents outside the design Bases on nuclear Central PWR Siemens-KWU and Westinghouse; Proyectos de Modificaciones de Sieno para Mitigacion de Accidentes fuera de la Bases de Diseno en Centrales Nucleares PWR Siemens-KWU y Westinghouse

Energy Technology Data Exchange (ETDEWEB)

Dominguez Gonzalez, G.; Cano Rodriguez, L. A.; Arguello Tara, A.

2014-07-01

Following the accident at the Japanese Fukushima-Daiichi NPP, the different regulators of nuclear power generation have required numerous reports regarding the evaluation and modification of the capacity of the plants to face accidents with severities beyond that established in their Design Bases. Under this new scenario, with multiple new demands and commitments, EA has carried out the required works for the implementation of strategies to mitigate the consequences of beyond Design Basis accidents for utilities owning Siemens-KWU and Westinghouse PWR nuclear power plants. (Author)

Power variation and frequency regulation. Adaptation of PWR plant possibilities to the network needs

International Nuclear Information System (INIS)

Baboulin, J.P.; Burger, M.

1980-01-01

When the PWR are an important part of the power installed on a network, and that will be the case of the EDF network in the coming years, the participation of those plants to the power regulating becomes a necessity for the operating staff. This load regulating includes: daily variations of high amplitude; a permanent frequency - power regulating. The first part of the communication shows the network exploitation principles, and the resulting power variations concerning the existing nuclear power plants. Such transients are leading to stresses on fuel. The second part of the communication reports about the test program engaged by EDF in collaboration with the CEA and FRAMATOME, in order to study the fuel behaviour in real power conditions and power cycles, and that, just to the operational burn up of this fuel. (author)

2006 nuclear power world report

International Nuclear Information System (INIS)

Anon.

2007-01-01

At the turn of 2006/2007, 437 nuclear power plants were available for energy supply, or were being commissioned, in 31 countries of the world. This is seven plants less than at the turn of 2005/2006. The aggregate gross power of the plants amounted to approx. 389.5 GWe, the aggregate net power, to 370.5 GWe. This indicates a slight decrease of gross power by some 0.15 GWe compared to the level the year before, while the available net power increased, also slightly, by approx. 0.2 GWe. The Tarapur 3 nuclear generating unit in India, a D 2 O PWR of 540 MWe gross power, was newly commissioned. In 2006, 8 nuclear power plants in Europe (4 in the United Kingdom, 2 in Bulgaria, 1 each in the Slovak Republic and in Spain) discontinued power operation for good. 29 nuclear generating units, i.e. 6 plants more than at the end of 2005, were under construction in late 2006 in 9 countries with an aggregate gross power of approx. 25.5 GWe. Worldwide, some 40 new nuclear power plants are in the concrete project design, planning, and licensing phases; in some of these cases, contracts have already been signed. Net electricity generation in nuclear power plants worldwide in 2006 achieved another top ranking level of approx. 2,660 billion kWh (2005: approx. 2,750 billion kWh). Since the first generation of electricity in a nuclear power plant in the EBR-1 fast breeder (USA) on December 20, 1951, cumulated gross production has reached approx. 56,875 billion kWh, and operating experience has grown to some 12,399 reactor years. (orig.)

Optimization of thermal efficiency of nuclear central power like as PWR

International Nuclear Information System (INIS)

Lapa, Nelbia da Silva

2005-10-01

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

Recent Advances in Ocean Nuclear Power Plants

Directory of Open Access Journals (Sweden)

Kang-Heon Lee

2015-10-01

Full Text Available In this paper, recent advances in Ocean Nuclear Power Plants (ONPPs are reviewed, including their general arrangement, design parameters, and safety features. The development of ONPP concepts have continued due to initiatives taking place in France, Russia, South Korea, and the United States. Russia’s first floating nuclear power stations utilizing the PWR technology (KLT-40S and the spar-type offshore floating nuclear power plant designed by a research group in United States are considered herein. The APR1400 and SMART mounted Gravity Based Structure (GBS-type ONPPs proposed by a research group in South Korea are also considered. In addition, a submerged-type ONPP designed by DCNS of France is taken into account. Last, issues and challenges related to ONPPs are discussed and summarized.

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

Korean experiences on nuclear power technology

International Nuclear Information System (INIS)

Kim, H.; Yang, H.

1994-01-01

This paper describes the outstanding performance of the indigenous development program of nuclear power technology such as the design and fabrication of both CANDU and PWR fuel and in the design and construction of nuclear steam supply system in Korea. The success has been accomplished through the successful technology transfer from foreign suppliers and efficient utilization of R and D manpower in the design and engineering of nuclear power projects. In order to implement the technology transfer successfully, the joint design concept has been introduced along with effective on-the-job training and the transfer of design documents and computer codes. Korea's successful development of nuclear power program has resulted in rapid expansion of nuclear power generation capacity in a short time, and the nuclear power has contributed to the national economy through lowering electricity price by about 50 % as well as stabilizing electricity supply in 1980s. The nuclear power is expected to play a key role in the future electricity supply in Korea. Now Korea is under way of taking a step toward advanced nuclear technology. The national electricity system expansion plan includes 18 more units of NPPs to be constructed by the year 2006. In this circumstance, the country has fixed the national long-term nuclear R and D program (lgg2-2001) to enhance the national capability of nuclear technology. This paper also briefly describes future prospects of nuclear technology development program in Korea

Nuclear power and the economic interests of consumers

International Nuclear Information System (INIS)

MacKerron, G.

1982-06-01

This study examines the main economic issues surrounding the use of nuclear power from a broadly consumer perspective. It sets out to answer four related questions: has nuclear investment been good value in the past; are current plans for nuclear investment (especially at Heysham II and beyond) likely to be good value in the future; what are the likely economic consequences if there is an explicit or de facto moratorium on the building of more nuclear power plants; and what strategic factors influence decision-making about nuclear investment and what are their consequences for consumers. (It should be noted that the CEGB's 'Statement of Case' for a PWR at Sizewell was published too late to be taken into account in this report). (author)

Analysis of the return to power scenario following a LBLOCA in a PWR

Energy Technology Data Exchange (ETDEWEB)

Macian, R.; Tyler, T.N.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)

1995-09-01

The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus, the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.

Experiences of operation for Ikata Nuclear Power Station

International Nuclear Information System (INIS)

Kashimoto, Shigeyuki

1979-01-01

No. 1 plant in the Ikata Nuclear Power Station, Shikoku Electric Power Co., Inc., is a two-loop PWR unit with electric output of 566 MW, and it began the commercial operation on September 30, 1977, as the first nuclear power station in Shikoku. It is the 13th LWR and 7th PWR in Japan. The period of construction was 52 months since it had been started in June, 1973. During the period, it became the object of the first administrative litigation to seek the cancellation of permission to install the reactor, and it was subjected to the influence of the violent economical variation due to the oil shock, but it was completed as scheduled. After the start of operation, it continued the satisfactory operation, and generated about 2.35 billion KWh for 4300 operation hours. It achieved the rate of utilization of 96.7%. Since March 28, 1978, the first periodical inspection was carried out, and abnormality was not found in the reactor, the steam generator and the fuel at all. The period of inspection was 79 days and shorter than expected. The commercial operation was started again on June 14. The outline of the Ikata Nuclear Power Station, its state of operation, and the periodical inspection are reported. Very good results were able to be reported on the operation for one year, thanks to the valuable experiences offered by other electric power companies. (Kako, I.)

Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals: 2007 update

International Nuclear Information System (INIS)

2007-06-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that effective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) is one of the most important issues for plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. IAEA-TECDOC-1119 documents ageing assessment and management practices for PWR Reactor Vessel Internals (RVIs) that were current at the time of its finalization in 1997-1998. Safety significant operating events have occurred since the finalization of the TECDOC, e.g. irradiation assisted stress corrosion cracking (IASCC) of baffle-former bolts, which threatened the integrity of the vessel internals. In addition, concern of fretting wear of control rod guide tubes has been raised in Japan. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1119 should be updated by incorporating those new events and their countermeasures. The objective of this report is to update relevant sections of the existing IAEA-TECDOC- 1119 in order to provide current ageing management guidance for PWR RVIs to all involved in the operation and regulation of PWRs and thus to help ensure PWR safety in IAEA Member States throughout their entire service life

Sizewell B: consent application for Britain's first PWR power station

International Nuclear Information System (INIS)

1981-02-01

The Central Electricity Generating Board has applied to the Secretary of State for Energy for consent and for other necessary permissions to construct a nuclear power station of about 1200 MW output capacity based on the pressurised water reactor (PWR) system on the Board's existing site at Sizewell (near Leiston) in Suffolk to be known as Sizewell B. Application has also been made to the Health and Safety Executive to extend the existing nuclear site licence to permit the use of the site for a pressurised water reactor. The Secretary of State for Energy has already stated that a Public Inquiry will be held into the application and this is expected to take place in 1982. The Board is making these applications now to give ample time for public discussion and consultation. Construction of the station could not begin until the outcome of the Public Inquiry is known and the necessary consents, nuclear licence and clearances have been given. The text of the application is presented. Some background information is given. (author)

Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

Science.gov (United States)

Liu, Ting; Qu, Yunhuan; Meng, De; Zhang, Qiaoer; Lu, Xinhua

2018-01-01

China’s spent fuel storage in the pressurized water reactors(PWR) is stored with wet storage way. With the rapid development of nuclear power industry, China’s NPPs(NPPs) will not be able to meet the problem of the production of spent fuel. Currently the world’s major nuclear power countries use dry storage as a way of spent fuel storage, so in recent years, China study on additional spent fuel dry storage system mainly. Part of the PWR NPP is ready to apply for additional spent fuel dry storage system. It also need to safety classificate to spent fuel dry storage facilities in PWR, but there is no standard for safety classification of spent fuel dry storage facilities in China. Because the storage facilities of the spent fuel dry storage are not part of the NPP, the classification standard of China’s NPPs is not applicable. This paper proposes the safety classification suggestion of the spent fuel dry storage for China’s PWR NPP, through to the study on China’s safety classification principles of PWR NPP in “Classification for the items of pressurized water reactor nuclear power plants (GB/T 17569-2013)”, and safety classification about spent fuel dry storage system in NUREG/CR - 6407 in the United States.

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

Energy Technology Data Exchange (ETDEWEB)

Lapa, Nelbia da Silva

2005-10-15

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

Axial power deviation control strategy and computer simulation for Daya Bay Nuclear Power Station

International Nuclear Information System (INIS)

Liao Yehong; Zhou Xiaoling, Xiao Min

2004-01-01

Daya Bay Nuclear Power Station has very tight operation diagram especially at its right side. Therefore the successful control of axial power deviation for PWR is crucial to nuclear safety. After analyzing various core characters' effect on axial power distribution, several axial power deviation control strategies has been proposed to comply with different power varying operation scenario. Application and computer simulation of the strategies has shown that our prediction of axial power deviation evolution are comparable to the measurement values, and that our control strategies are effective. Engineering experience shows that the application of our methodology can predict accurately the transient of axial power deviation, and therefore has become a useful tool for reactor operation and safety control. This paper presents the axial power control characteristics, reactor operation strategy research, computer simulation, and comparison to measurement results in Daya Bay Nuclear Power Station. (author)

Safety device and machine system of nuclear power plant

International Nuclear Information System (INIS)

1978-10-01

It introduces principle and kinds of heat power including heat balance and nuclear power. It explains a lot of technical terms about the nuclear power system, which are primary loop, reactor, steam generator, primary coolant pump and pressurizer in PWR, chemical and volume control system, component cooling system, safety injection system, and spent fuel cooling and storage system in auxiliary system, liquid solid and gaseous waste disposal system in radwaste disposal, gland sealing system, turbine instrumentation, turning gear, hydrogen cooling system, condenser, feedwater heater, degenerate heater, auxiliary heat exchanger, centrifugal pump, rotary reciprocating and tank and pressure vessel.

Nuclear power plants in the world - 2010 edition

International Nuclear Information System (INIS)

2010-01-01

This small booklet summarizes in tables all data relative to the nuclear power plants worldwide. These data come from the IAEA's PRIS and AREVA-CEA's GAIA databases. The following aspects are reviewed: 2009 highlights, Main characteristics of reactor types, Map of the French nuclear power plants on 2010/01/01, Worldwide status of nuclear power plants (12/31/2009), Units distributed by countries, Nuclear power plants connected to the Grid- by reactor type groups, Nuclear power plants under construction on 2009, Evolution of nuclear power plants capacities connected to the grid, First electric generations supplied by a nuclear unit in each country, Electrical generation from nuclear power plants by country at the end 2009, Performance indicator of french PWR units, Evolution of the generation indicators worldwide by type, Nuclear operator ranking according to their installed capacity, Units connected to the grid by countries at 12/31/2009, Status of licence renewal applications in USA, Nuclear power plants under construction at 12/31/2009, Shutdown reactors, Exported nuclear capacity in net MWe, Exported and national nuclear capacity connected to the grid, Exported nuclear power plants under construction, Exported and national nuclear capacity under construction, Nuclear power plants ordered at 12/31/2009, Long term shutdown units at 12/31/2009, COL applications in the USA, Recycling of Plutonium in reactors and experiences, Mox licence plants projects, Appendix - historical development, Meaning of the used acronyms, Glossary

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.)

Structural integrity evaluation of PWR nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Cruz, Julio R.B.; Mattar Neto, Miguel

1999-01-01

The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

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

Nuclear power: consolidation or change

International Nuclear Information System (INIS)

1986-02-01

Currently about 45% of the electricity consumed in Scotland comes from nuclear sources and when the power station at Torness, near Edinburgh, is fully commissioned it will be about 60%. Torness is an AGR type reactor and the Scottish Electricty Board (SSEB) consider that there is no case for changing to a PWR type design for any future reactor built in the UK. The economics of 'a next AGR' and a PWR reactor are compared. The construction times for the Torness programme, the electrical output, station life, load factor, fuel costs and additional support costs are also considered, and compared with costs estimated by the Central Elecricity Generating Board (CEGB) for the proposed Sizewell-B reactor. Exposure of station staff to radiation, safety, public acceptibility and development potential, impact on industry, job creation and export potential are also discussed briefly. (UK)

A full scope nuclear power plant simulator for multiple reactor types with virtual control panels

International Nuclear Information System (INIS)

Yonezawa, Hisanori; Ueda, Hiroki; Kato, Takahisa

2017-01-01

This paper summarizes a full scope nuclear power plant simulator for multiple reactor types with virtual control panels which Toshiba developed and delivered. After the Fukushima DAIICHI nuclear power plants accident, it is required that all the people who are engaged in the design, manufacturing, operation, maintenance, management and regulation for the nuclear power plant should learn the wide and deep knowledge about the nuclear power plant design including the severe accident. For this purpose, the training with a full scope simulator is one of the most suitable ways. However the existing full scope simulators which are consist of the control panels replica of the referenced plants are costly and they are hard to remodel to fit to the real plant of the latest condition. That's why Toshiba developed and delivered the new concept simulator system which covers multiple referenced plants even though they have different design like BWR and PWR. The control panels of the simulator are made by combining 69 large Liquid Crystal Display (LCD) panels with touch screen instead of a control panel replica of referenced plant. The screen size of the each panel is 42 inches and 3 displays are arranged in tandem for one unit and 23 units are connected together. Each panel displays switches, indicators, recorders and lamps with the Computer Graphics (CG) and trainees operate them with touch operations. The simulator includes a BWR and a PWR simulator model, which enable trainees to learn the wide and deep knowledge about the nuclear power plant of BWR and PWR reactor types. (author)

2002 Nuclear Power World Report - Evaluation

International Nuclear Information System (INIS)

Anon.

2003-01-01

Last year, in 2002, 441 nuclear power plants were available for power supply in 31 countries in the world. With an aggregate gross power of 377,359 MWe, and an aggregate net power of 359,429 MWe, respectively, the nuclear generating capacity reached its highest level so far. Nine different reactor lines are used in commercial facilities. Light water reactors (PWR and BWR) contribute 355 plants, which makes them the most common reactor line. In twelve countries, 32 nuclear power plants with an aggregate gross power of 26,842 MWe and an aggregate net power of 25,546 MWe, respectively, are under construction. Of these, 25 units are light water reactors while eight are CANDU-type plants. In eighteen countries, 94 commercial reactors with more than 5 MWe power have been decommissioned so far. Most of these plants are prototypes with low powers. 228 of the nuclear power plants currently in operation, i.e. slightly more than half of them, were commissioned in the eighties. The oldest commercial nuclear power plant, Calder Hall unit 1, supplied power into the public grid in its 47th year of operation in 2002. The availability in terms of time and capacity of nuclear power plants rose from 74.23% in 1991 to 83.40% in 2001. A continued rise to approx. 85% is expected for 2002. In the same way, the non-availability in terms of time (unscheduled) dropped from 6.90% to 3.48%. The four nuclear power plants in Finland are the world's leaders with a cumulated average capacity availability of 90.00%. (orig.) [de

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

Adaptive control of a PWR core power using neural networks

International Nuclear Information System (INIS)

Arab-Alibeik, H.; Setayeshi, S.

2005-01-01

Reactor power control is important because of safety concerns and the call for regular and appropriate operation of nuclear power plants. It seems that the load-follow operation of these plants will be unavoidable in the future. Discrepancies between the real plant and the model used in controller design for load-follow operation encourage one to use auto-tuning and (or) adaptive techniques. Neural network technology shows great promise for addressing many problems in non-model-based adaptive control methods. Also, there has been a great attention to inverse control especially in the neural and fuzzy control context. Fortunately, online adaptation eliminates some limitations of inverse control and its shortcomings for real world applications. We use a neural adaptive inverse controller to control the power of a PWR reactor. The stability of the system and convergence of the controller parameters are guaranteed during online adaptation phase provided the controller is near the plant's real inverse after offline training period. The performance of the controller is verified using nonlinear simulations in diverse operating conditions

Nuclear power - a reliable future

International Nuclear Information System (INIS)

Valeca, Serban

2002-01-01

development of advanced fuels based on slightly enriched uranium recovered from enriched fuel treatment as well as on fuel cycles using the spent fuel from PWR reactors in CANDU reactors. The paper addresses also legal aspects of nuclear power, international conventions and agreements and international cooperation in the nuclear field

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Study of the concrete tensile creep: application for the containment vessel of the nuclear power plants (PWR)

International Nuclear Information System (INIS)

Reviron, Nanthilde

2009-01-01

The aim of this work is to study experimentally and to conduct numerical simulations on the creep of concrete subjected to tensile stresses. The main purpose is to predict the behaviour of containment vessels of nuclear power plants (PWR) in the case of decennial test or accident. In order to satisfy to these industrial needs, it is necessary to characterize the behaviour of concrete under uniaxial tension. Thus, an important experimental study of tensile creep in concrete has been performed for different loading levels (50%, 70% and 90% of the tensile strength). In these tests, load was kept constant during 3 days. Several tests were performed: measurements of elastic properties and strength (in tension and in compression), monitoring of drying, shrinkage, basic creep and drying creep strains. Moreover, compressive creep tests were also performed and showed a difference with tensile creep. Furthermore, decrease of tensile strength and failure under tensile creep for large loading levels were observed. A numerical model has been proposed and developed in Cast3m finite element code. (author)

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)

Improvement of availability of PWR nuclear plants through the reduction of the time required for refueling/maintenance outages

International Nuclear Information System (INIS)

Mayers, J.B.; Soth, L.G.

1978-04-01

The objective of the project, conducted by Commonwealth Research Corporation and Westinghouse Electric Corporation, is to identify improvements in procedures and equipment which will reduce the time required for refueling/maintenance outages at PWR nuclear power plants. The outage of Commonwealth Edison Zion Station Unit 1 in March through May of 1976 was evaluated to identify those items which caused delays and those work activities that offer the potential for significant improvements that could reduce the overall duration of the outage and achieve an improvement in the plant's availability for power production. Modifications in procedures have been developed and were evaluated during one or more outages in 1977. Conceptual designs have been developed for equipment modifications to the refueling system that could reduce the time required for the refueling portion of the outage. The purpose of the interim report is to describe those conceptual designs and to assess their impact upon future outages. Recommendations are included for the implementation of these equipment improvements in a continuation of this program as a demonstration of plant availability benefits that can be realized in PWR nuclear plants already in operation or under construction

Analysis of adverse events occurred at overseas nuclear power plants in 2003

International Nuclear Information System (INIS)

Miyazaki, Takamasa; Sato, Masahiro; Takagawa, Kenichi; Fushimi, Yasuyuki; Shimada, Hiroki; Shimada, Yoshio

2004-01-01

The adverse events that have occurred in the overseas nuclear power plants can be studied to provide an indication of how to improve the safety and the reliability of nuclear power plants in Japan. The Institute of Nuclear Safety Systems (INSS) obtains information related to overseas adverse events and incidents, and by evaluating them proposes improvements to prevent similar occurrences in Japanese PWR plants. In 2003, INSS obtained approximately 2800 pieces of information and, by evaluating them, proposed nine recommendations to Japanese utilities. This report shows a summary of the evaluation activity and of the tendency analysis based on individual event analyzed in 2003. The tendency analysis was undertaken on about 1600 analyzed events, from the view point of Mechanics, Electrics, Instruments and Controls and Operations, about the causes, countermeasures, troubled equipments and the possible of lessons learnt from overseas events. This report is to show the whole tendency of overseas events and incidents for the improvement of the safety and reliability of domestic PWR plants. (author)

The new operating conditions of French nuclear power plants

International Nuclear Information System (INIS)

Leclercq, J.

1986-01-01

Six themes are examined: France's unique position in view of the size of its nuclear operating plant, the role of nuclear power in matching electricity supply to demand, the excellent flexibility provided by PWR facilities in operation, the approaches used in the field of automatic operational control systems, the systematic use of data processing for maintenance and generation and the gains in productivity that can be gained as a result of improving fuel use [fr

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

Computations of steam flow and heat transfer in nuclear power plant condensers

International Nuclear Information System (INIS)

Yuan, A.

1997-01-01

To improve performance of its PWR nuclear power plants, Electricite de France has developed a performance monitoring system that checks simultaneously the operation of the components of the secondary system. The performance monitoring system is based on a computational software CITER for steady state runs. A one-dimensional condenser model has been developed. Application of this code to a nuclear power plant condenser shows that predicted values in good agreement with the design values

Factors analysis of water hammer in FLOWMASTER for main feedwater systems of PWR nuclear power plants

International Nuclear Information System (INIS)

Wang Xin; Han Weishi

2010-01-01

The main feedwater system of a nuclear power plant (NPP) is an important part in ensuring the cooling of a steam generator. It is the main pipe section where water hammers frequently occur. Studying the regulator patterns of water hammers in the main feedwater systems is significant to the stable operation of the system. This article focuses on a parametric study to avoid the consequences of water hammer effect in PWR by employing a general purpose fluid dynamic simulation software-FLOWMASTER. Through FLOWMASTER's transient calculating functions, a mathematical model is established with boundary conditions such as feedwater pumps, control valves, etc., calculations of water hammer pressure when feedwater pumps and control valves shut down, and simulations during instantaneous changes in water hammer pressure. Combining a plethora of engineering practical examples, this research verified the viability of calculating water hammer pressure through FLOWMASTER's transient functions and we found out that, increasing the periods of closure of control valves and feedwater pumps control water hammers effectively. We also found out that changing the intervals of closing signals to feedwater pumps and control valves aid to relieve hydraulic impact. This could be a guideline for practical engineering design and system optimization. (author)

ELECNUC Nuclear power plants in the world - 2013 edition

International Nuclear Information System (INIS)

2013-01-01

This small booklet summarizes in a series of tables the figures relative to the nuclear power plants worldwide. Data come from the IAEA's PRIS database and from specific I-tese studies. The following aspects are reviewed: 2012 highlights; Main characteristics of reactor types; Map of the French nuclear power plants on 2012/01/01; Worldwide status of nuclear power plants (12/31/2012); Units distributed by countries; Nuclear power plants connected to the Grid- by reactor type groups; Nuclear power plants under construction on 2012; Evolution of nuclear power plants capacities connected to the grid; First electric generations supplied by a nuclear unit in each country; Electrical generation from nuclear power plants by country at the end 2012; Performance indicator of french PWR units; Evolution of the generation indicators worldwide by type; Nuclear operator ranking according to their installed capacity; Units connected to the grid by countries at 12/31/2012; Status of licence renewal applications in USA; Nuclear power plants under construction at 12/31/2012; Shutdown reactors; Exported nuclear capacity in net MWe; Exported and national nuclear capacity connected to the grid; Exported nuclear power plants under construction; Exported and national nuclear capacity under construction; Nuclear power plants ordered at 12/31/2012; Long term shutdown units at 12/31/2012; COL (Combined Licence) applications in the USA; Recycling of Plutonium in reactors and experiences; Mox licence plants projects; Appendix - historical development; Meaning of the used acronyms; Glossary

13th International conference on environmental degradation of materials in nuclear power systems

International Nuclear Information System (INIS)

2007-01-01

The 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems was held on August 19-23, 2007 in Whistler, British Columbia, Canada. More of a scientific meeting than a convention, this conference series is the premier nuclear industry corrosion meeting where the 225 registrations consisted of world experts of the field from utilities, engineering and service organizations, manufacturers, research establishments and universities gathered to listen to 144 technical papers on new work and to explore new insights into corrosion mechanisms in the many water cooled systems in nuclear power plants. Over 225 delegates attended the conference, over 144 technical papers were presented in the following sessions: IASCC; Waste; PWR Secondary; Ni-Base Welds; Operating Experience; Low Alloy Steels; Alloy 800 Steam Generator Tubing; Zirconium Alloys; Crack Growth; SCWR; PWR Primary; BWR SCC; Irradiation Effects; Flow Accelerated Corrosion; and, Nobel Metal

Simulation model of a PWR power plant

International Nuclear Information System (INIS)

Larsen, N.

1987-03-01

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

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

International Nuclear Information System (INIS)

Schoemacker, M.

1984-01-01

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

Elecnuc - Nuclear power plants in the world - 2009 edition

International Nuclear Information System (INIS)

2009-01-01

This small booklet summarizes in tables all data relative to the nuclear power plants worldwide. These data come from the IAEA's PRIS and AREVA-CEA's GAIA databases. The following aspects are reviewed: 2008 highlights, Main characteristics of reactor types, Map of the French nuclear power plants on 2008/01/01, Worldwide status of nuclear power plants (12/31/2008), Units distributed by countries, Nuclear power plants connected to the Grid- by reactor type groups, Nuclear power plants under construction on 2008, Evolution of nuclear power plants capacities connected to the grid, First electric generations supplied by a nuclear unit in each country, Electrical generation from nuclear powe plants by country at the end 2008, Performance indicator of french PWR units, Evolution of the generation indicators worldwide by type, Nuclear operator ranking according to their installed capacity, Units connected to the grid by countries at 12/31/2008, Status of licence renewal applications in USA, Nuclear power plants under construction at 12/31/2008, Shutdown reactors, Exported nuclear capacity in net MWe, Exported and national nuclear capacity connected to the grid, Exported nuclear power plants under construction, Exported and national nuclear capacity under construction, Nuclear power plants ordered at 12/31/2008, Long term shutdown units at 12/31/2008, COL applications in the USA, Recycling of Plutonium in reactors and experiences, Mox licence plants projects, Appendix - historical development, Meaning of the used acronyms, Glossary

Experience gained in the training of nuclear power plant operating personnel with nuclear power plant simulators

International Nuclear Information System (INIS)

Buettner, J.; Fueg, J.; Schlegel, G.

1980-01-01

The simulator of a PWR-type reactor with 1.200 MW was accomplished in September 1977. In January 1978, the simulator of a BWR-type reactor with 800 MW started operation. The American company Singer/Link supplied computer hardware and software; Kraftwerk Union AG supplied control room equipment, power plant data and acted as consulting engineers for the construction and acceptance of the simulators. This way it is ensured that the simulated process reflects the state of German nuclear engineering. (orig./DG) [de

Changes in the structure and relations of property in the Great Britain nuclear power

International Nuclear Information System (INIS)

Rumyantsev, V.V.

1990-01-01

Analysis of nuclear power development in the Great Britain is given. Currently it is reoriented to a wide application of PWR reactors instead of gas cooled ones, their economic indices being incapable of providing sufficient competitiveness on the British market. Reasons and state of affairs in denationalizing nuclear power, economic and financial difficulties connected with the realization of this politico-economic procedure are reported

Dispositions taken in France to limit gaseous releases from PWR power plants in abnormal operating conditions

International Nuclear Information System (INIS)

Collinet, J.; Guieu, S.; Mulcey, P.

1989-12-01

The implementation of France's major nuclear programme - 56 PWR units in service or under construction - has gone hand in hand with the development of an original philosophy in the field of nuclear safety. From an initial core of deterministic safety philosophy current in the seventies, which has been wholly retained and, in some instances, refined, a range of complements has been made to include consideration of a number of additional situations based on a probabilistic approach. This has resulted in a better coherence for safety and a reduction of the severe accident probability. Furthermore, the establishment of emergency plans has enabled the Safety Authorities and the utility to adopt a coherent and logical approach to severe accidents, with the aim of better achieving defence in depth. This has resulted in the provision of certain additional measures intended to further reduce the consequences of severe accidents. In a accordance with the safety philosophy, adopted in France for nuclear PWR power stations, filtration systems have been specified and installed to limit the radiological consequences of consecutive gaseous emissions, on the one hand, in accidents taken into account in the design and, on the other hand, in accidents liable to jeopardize the integrity of the containment

2004 world nuclear power report - evaluation

International Nuclear Information System (INIS)

Anon.

2004-01-01

Last year, 2003, 439 nuclear power plants were available for electricity generation in 31 countries of the world. With an aggregate gross capacity of 380,489 MWe and an aggregate net capacity of 361,476MWe, nuclear generating capacity reached its highest level so far. Nine different reactor lines are operated in the commercial nuclear power plants. Light water reactors (PWR and BWR) continue to be in the lead with 355 plants. Twenty-nine nuclear power plants with an aggregate gross capacity of 24,222 MWe and an aggregate net capacity of 23,066 MWe were under construction in eleven countries. Of these, twenty are light water reactors, and seven are CANDU-type reactors. Ninety-nine commercial reactors with a capacity in excess of 5 MWe have so far been decommissioned in eighteen countries, most of them prototype plants of low power. 228 plants, i. e. slightly more than half of the number of plants currently in operation, were commissioned in the 1980s. The oldest commercial nuclear power plant in the world, Calder Hall unit 1, was disconnected from the power grid for good in its 48th year of operation in 2003. For the first time in ten years, the availability in terms of time and capacity of nuclear power plants has decreased from 83,80% in 2002 to 80.50%, and from 84.60% to 81.50%, respectively, in 2003. The main causes are prolonged outages of high-capacity plants in Japan as a consequence of administrative restrictions. The four nuclear power plants in Finland continue to be at the top of the list worldwide with a cumulated average availability of capacity of 90.30%. (orig.)

Nuclear power in the UK electricity market

International Nuclear Information System (INIS)

Coffey, J.M.

1995-01-01

Nuclear Electric was formed in the public sector to operate only nuclear power plant, and the Company has been foremost in developing the UK's capability for PWR design and construction. It is now obliged to compete on equal terms with privately-owned generators, and we have made it clear that we would invest in further nuclear plant only if the terms were commercially attractive to the company. The competitive environment in which we now operate has led us to recognise that the priority for the Company in the Nuclear Review is to seek the commercial flexibility which accompanies privatisation. Accordingly, our evidence to the Government in the Nuclear Review has shown that the problems of confidence which surrounded nuclear power in 1989 have been substantially resolved. The improved accounting costs and low avoidable costs of the existing stations make the commercial case for their continued operation. The completion of Szewell B has not only given us a gist class new, profitable power plant, but given confidence in the costs and performance of any follow-on PWRs. In the longer term, a greater recognition of the external environmental costs of fossil-fuel generation may swing the market in favour of nuclar power construction. (orig.) [de

Design of nuclear power plants

International Nuclear Information System (INIS)

Lobo, C.G.

1987-01-01

The criteria of design and safety, applied internationally to systems and components of PWR type reactors, are described. The main criteria of the design analysed are: thermohydraulic optimization; optimized arrangement of buildings and components; low costs of energy generation; high level of standardization; application of specific safety criteria for nuclear power plants. The safety criteria aim to: assure the safe reactor shutdown; remove the residual heat and; avoid the release of radioactive elements for environment. Some exemples of safety criteria are given for Angra-2 and Angra-3 reactors. (M.C.K.) [pt

Future efforts on safety security at nuclear power stations

International Nuclear Information System (INIS)

Kondo, Shunsuke

2003-01-01

As operation management of nuclear power generation in Japan was at the highest level in the world at beginning of 1990s, Japan has gradually been left behind by foreign countries at indices such as its operation ratio, its employees' exposure, and so on, and is at a general level. In special, as PWR showed 89% in its operation ratio corresponding to international level on PWR at last year, BWR was concentrated to countermeasure of stress corrosion cracking (SCC) caused by storage on consideration of stress relaxation at machining of parts made of SUS-316LC at a number of nuclear reactors, and all of units in the Tokyo Electric Power Co., Ltd. had an accident on feasibility to cease them by finding out incorrect deeds at past periodical and self inspections. The minor Committee on Nuclear Regulation Rule Investigation of the Nuclear Security Party of the Advisory Committee for Energy judged this accident formed by neglecting tense feelings on inspection based on shortage of recognition on necessity to do administrative explanation obligation for natives and preparation of quality assurance system expressible on validity of safety management for local society of customers by management center in electric business companies, to propose a countermeasure to be done by government and private companies. Here was expressed future important subjects under their outlines. (G.K.)

Technical diagnostics - equipment monitoring for increasing safety and availability of nuclear power plants

International Nuclear Information System (INIS)

Sturm, A.; Foerster, R.

1977-01-01

Utilization of technical diagnostics in equipment monitoring of nuclear power plants for ensuring nuclear safety, economic availability, and for decision making on necessary maintenance is reviewed. Technical diagnostics is subdivided into inspection and early detection of malfunctions. Moreover, combination of technical diagnostics and equipment monitoring, integration of technical diagnostics into maintenance strategy, and problems of introducing early detection of malfunctions into maintenance management of nuclear power plants are also discussed. In addition, a compilation of measuring techniques used in technical diagnostics has been made. The international state of the art of equipment monitoring in PWR nuclear power plants is illustrated by description of the sound and vibration measuring techniques. (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

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)

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

International Nuclear Information System (INIS)

Affonso, Renato Raoni Werneck; Costa, Danilo Leite; Borges, Diogo da Silva; Lava, Deise Diana; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes

2014-01-01

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

State of the art and prospects of nuclear power development

International Nuclear Information System (INIS)

Egorov, Yu.V.; Glushenkova, S.Z.

1984-01-01

State of the art and prospects of nuclear power development abroad except the COMECON countries and Yugoslavia, are considered. Both average electric power of power units and load factor (LF) are shown to increase. Data on both the total generating capacity and nuclear power generation in certain countries are given. The number of commissioned NPPs in the USA decreases but terms of their construction and licensing are reduced, program of fast breeder reactor construction is being realized, prohibition of nuclear fuel reprocessing in cancelled. France came to the second place in the world as regards the operating NPPs. The nuclear power generation in Japan makes up 13% of the total generating capacity in the country. The LF of the Japan NPPs with BWR and PWR in 1982 made up 70.7 and 69.8%, respectively. A higher reliability of NPPs, decrease in the time for periodical inspections and prolongation of the operating cycle have promoted an increase in the LF

Nuclear Power and Safety Division activity

International Nuclear Information System (INIS)

Pazdera, F.

1991-01-01

History of the Division is briefly described. Present research is centered on reliability analyses and thermal hydraulic analyses of transients and accidents. Some results of the safety analyses have been applied at nuclear power plants. A characterization is presented of computer codes for analyzing the behavior of fuel in normal and accident conditions. Research activities in the field of water chemistry and corrosion are oriented to the corrosion process at high temperatures and high pressures, and the related mass and radioactivity transfer; the effect of some chemical processes on primary coolant circuit materials; optimization of PWR filtration systems; and the development of the requisite monitoring instrumentation. A computerized operator support system has been developed, and at present it is tested at the Dukovany nuclear power plant. A program of nuclear fuel cycle strategy and economy has been worked out for nuclear fuel performance evaluation. Various options for better fuel exploitation, alternatives for advanced fuelling, and fuel cycle costs are assessed, and out-of-reactor fuel cycle options are compared. (M.D.). 7 refs., 32 refs

Waste generation comparison: Coal-fired versus nuclear power plants

International Nuclear Information System (INIS)

LaGuardia, T.S.

1998-01-01

Low-level radioactive waste generation and disposal attract a great deal of attention whenever the nuclear industry is scrutinized by concerned parties, be it the media, the public, or political interests. It is therefore important to the nuclear industry that this issue be put into perspective relative to other current forms of energy production. Most of the country's fossil-fueled power comes from coal-fired plants, with oil and gas as other fuel sources. Most of the generated waste also comes from coal plants. This paper, therefore, compares waste quantities generated by a typical (1150-MW(electric)) pressurized water reactor (PWR) to that of a comparably sized coal-fired power plant

The development of nuclear power as an alternative energy

International Nuclear Information System (INIS)

Zheng Mingguang; Ye Cheng; Han Xu

2010-01-01

In this paper, we discuss the world's recent development of nuclear power, which does not seem to achieve what has been anticipated by the groups of experts. In this regard, there is before us a heavy responsibility and a long way to go. Nuclear power is very important for China's environmental protection and economic development. Suggestions are made for longer-term demand for the electricity and long-term nuclear power plan in China.According to the nuclear power strategy adopted by developed nations, we believe that China's nuclear power strategy and nuclear techniques shall be based on advanced PWR, such as AP1000, CAP1000, CAP1400 and CAP1700, which will be the main NPP types in a long time in China, and the latter two shall be highly competitive in the world.For generation IV reactors, the SCWR (Supercritical Water Reactor) is more suitable than others, in terms of the industrial development in PWRs and fossil fuel super critical generator systems in China. In addition,investment should be increased in fast reactors to achieve the fuel closed-cycle, and attentions should be paid to the TWR (Traveling-wave Reactor), and small and medium-sized reactor as well.Finally, conclusions are given for large scale development of nuclear power in China. (authors)

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

Degradation of fastener in reactor internal of PWR

Energy Technology Data Exchange (ETDEWEB)

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

2000-03-01

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

Analysis of design of auxiliary system of Booshehr Nuclear Power Plant

International Nuclear Information System (INIS)

Naseh Hasanzadeh, M.

1999-01-01

Power plant's internal auxiliary system has an important role in its safety operation. Because of the decay heat and safety aspects in the nuclear power plants, this role is more important. In this thesis, operation of the nuclear power plant with PWR reactor is studied and deferent nuclear systems described. In the next section all electrical loads in the Booshehr Nuclear Power Plant identified and feeding methods of each load is determined. by use of the single line diagram of the internal auxiliary system, the nominal rating of all electrical devices as transformers, inverters, Ups, diesel generators and etc. is determined. In the following, short circuit calculations performed and by above conclusion, rating values of circuit breakers is determined. At last the starting problems of electrical motors is studied and the results of motor's behavior at starting moment is discussed

Plant Design Nuclear Fuel Element Production Capacity Optimization to Support Nuclear Power Plant in Indonesia

International Nuclear Information System (INIS)

Bambang Galung Susanto

2007-01-01

The optimization production capacity for designing nuclear fuel element fabrication plant in Indonesia to support the nuclear power plant has been done. From calculation and by assuming that nuclear power plant to be built in Indonesia as much as 12 NPP and having capacity each 1000 MW, the optimum capacity for nuclear fuel element fabrication plant is 710 ton UO 2 /year. The optimum capacity production selected, has considered some aspects such as fraction batch (cycle, n = 3), length of cycle (18 months), discharge burn-up value (Bd) 35,000 up 50,000 MWD/ton U, enriched uranium to be used in the NPP (3.22 % to 4.51 %), future market development for fuel element, and the trend of capacity production selected by advances country to built nuclear fuel element fabrication plant type of PWR. (author)

Methods and benefits of experimental seismic evaluation of nuclear power plants. Final report

International Nuclear Information System (INIS)

1979-07-01

This study reviews experimental techniques, instrumentation requirements, safety considerations, and benefits of performing vibration tests on nuclear power plant containments and internal components. The emphasis is on testing to improve seismic structural models. Techniques for identification of resonant frequencies, damping, and mode shapes, are discussed. The benefits of testing with regard to increased damping and more accurate computer models are oulined. A test plan, schedule and budget are presented for a typical PWR nuclear power plant

Methods and benefits of experimental seismic evaluation of nuclear power plants. Final report

Energy Technology Data Exchange (ETDEWEB)

1979-07-01

This study reviews experimental techniques, instrumentation requirements, safety considerations, and benefits of performing vibration tests on nuclear power plant containments and internal components. The emphasis is on testing to improve seismic structural models. Techniques for identification of resonant frequencies, damping, and mode shapes, are discussed. The benefits of testing with regard to increased damping and more accurate computer models are oulined. A test plan, schedule and budget are presented for a typical PWR nuclear power plant.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Chemistry Optimitation of a PWR nuclear power plant

International Nuclear Information System (INIS)

Frandrich, J.; Ramminger, U.

2015-01-01

One of the main objectives for the plant operator of a Nuclear Power Plant is to protect the Steam Generators (SG) during the lifetime of the plant by ensuring a safe and reliable operation. The SGs serve as an important barrier to prevent the spread of contamination out of the primary circuit. One the other hand impurities are accumulated within the SGs leading to extreme chemical and physical conditions. The application of an optimized water chemistry treatment of the secondary side is essential to ensure a good performance of the steam generators. (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)

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

International Nuclear Information System (INIS)

Thorlaksen, B.

1981-05-01

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

The modernization of the process computer of the Trillo Nuclear Power Plant

International Nuclear Information System (INIS)

Martin Aparicio, J.; Atanasio, J.

2011-01-01

The paper describes the modernization of the Process computer of the Trillo Nuclear Power Plant. The process computer functions, have been incorporated in the non Safety I and C platform selected in Trillo NPP: the Siemens SPPA-T2000 OM690 (formerly known as Teleperm XP). The upgrade of the Human Machine Interface of the control room has been included in the project. The modernization project has followed the same development process used in the upgrade of the process computer of PWR German nuclear power plants. (Author)

Nuclear power training programmes in Spain

International Nuclear Information System (INIS)

Tanarro, A.; Izquierdo, L.

1977-01-01

The introduction of nuclear power in Spain is developing very rapidly. At present 1.1GW(e) are installed in Spain and this is expected to increase to 8GW(e) in 1980 and to 28GW(e) in 1990. Spanish industry and technology are also rapidly increasing their participation in building nuclear stations, in manufacturing the necessary components and in the activities related to the nuclear fuel cycle. All of this requires properly trained personnel, which is estimated to become approximately 1200 high-level technicians, 1100 medium-level technicians and 1500 technical assistants by 1980. This personnel is trained: (a) in engineering schools; (b) in the Nuclear Studies Institute; (c) in the electric companies with nuclear programmes. The majority of the high-level engineering schools in the country include physics and basic nuclear technology courses in their programmes. Some of them have an experimental low-power nuclear reactor. The Nuclear Studies Institute is an official organism dependent on the Nuclear Energy Commission and responsible, among other subjects, for training personnel for the peaceful use and development of nuclear energy in the country. The electric companies also participate in training personnel for future nuclear stations and they plan to have advanced simulators of PWR and BWR type stations for operator training. The report deals with the personnel requirement forecasts and describes the training programmes. (author)

The outlook for nuclear power in Europe by 2030

International Nuclear Information System (INIS)

Leclercq, J.

2008-01-01

A 57% increase in the world consumption of electricity is expected between 2004 and 2030. According to the alternative policy scenario of the world energy outlook 2006, the contribution of nuclear power would be +300 GW for a total increase of +4600 GW in power production. The economic outlook for nuclear power appears to be favorable over a long period. Between 2006 and 2030, about 30 countries will order the construction of nuclear power plants but only 5 countries (Usa, China, Japan, Russia and India) will concentrate the 2/3 of this demand. This demand will be met mostly with 10 commercial offers representing reactors of third generation (6 PWR-types + 3 BWR-types + 1 Candu-type). The existing resources of natural uranium (about 15*10 6 tonnes) are sufficient to ensure in 2040 a global nuclear power as high as 3 to 4 times the today's nuclear power. As for Europe, 2 scenarios are considered: an evolution of -60 GW in case of no decision concerning the construction of new nuclear plants and a likely +120 GW scenario including the replacement of 64 GW. The second scenario will lead to an installed capacity of 229 GW in 2030 compared to today's 172 GW. (A.C.)

Generic nuclear power plant component failure data bank

International Nuclear Information System (INIS)

Araujo Goes, A.G. de; Gibelli, S.M.O.

1988-11-01

This report consist in the development of a generic nuclear power plant component failure data bank. This data bank was implemented in a PC-XT microcomputer, IBM compatible, using the Open Access II program. Generic failure data tables for Westinghouse nuclear power plants and for general PWR power plants are presented. They are the final product of a research which included a preselection and a selection of data collected from the available sources in the library of CNEN (National Nuclear Energy Commission) and from the CIN/CNEN (Neclear Information Center). Futhermore, a proposal of evaluating models of average failure rates of pumps and valves are also presented. Through the electronic data bank one can easily have a generic view of failure rate ranges as well as failure models foe a certain component. It is very importante to develop procedures to collect and store generic failure data that can be quickly accessed, in order to update the Probabilistic Safety Study of Angra-1 and to used in studies which may have component failures of nuclear power plant safety systems. In the future, data specialization can be achieved by means of statistical calculations involving specific data collected from the operational experience of Angra-1 nuclear power plant and the generic data bank. (author) [pt

Research and design of distributed intelligence fault diagnosis system in nuclear power plant

International Nuclear Information System (INIS)

Liu Yongkuo; Xie Chunli; Cheng Shouyu; Xia Hong

2011-01-01

In order to further reduce the misoperation after the faults occurring of nuclear power plant, according to the function distribution of nuclear power equipment and the distributed control features of digital instrument control system, a nuclear power plant distributed condition monitoring and fault diagnosis system was researched and designed. Based on decomposition-integrated diagnostic thinking, a fuzzy neural network and RBF neural network was presented to do the distributed local diagnosis and multi-source information fusion technology for the global integrated diagnosis. Simulation results show that the developed distributed status monitoring and fault diagnosis system can diagnose more typical accidents of PWR to provide effective diagnosis and operation information. (authors)

The French nuclear programme

International Nuclear Information System (INIS)

Bacher, Pierre

1987-01-01

France has a civil nuclear power generation programme second only to the USA with 49 nuclear units in operation and 13 under construction. The units in service are described. These include 33 PWR 900 MW and 9 PWR 1300 MW units. The electricity consumption and generation in France is illustrated. The absence of a powerful anti-nuclear lobby and two main technical options have contributed to the success of the French nuclear programme. These are the PWR design and the plant standardization policy which allows the setting up of an effective industrial complex (eg for analysis of operating conditions and of safety and reliability information). The programme and the reasons for its success are reviewed. Research programmes and future plans are also discussed. (UK)

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

International Nuclear Information System (INIS)

Cao Xiaoning; Li Yunde; Li Jinghong; Lin Fangliang

1997-01-01

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

Application of the neutron noise analysis technique in nuclear power plants

International Nuclear Information System (INIS)

Lescano, Victor H.; Wentzeis, Luis M.

1999-01-01

Using the neutron noise analysis in nuclear power plants, and without producing any perturbation in the normal operation of the plant, information of the vibration state of the reactor internals and the behavior of the operating conditions of the reactor primary circuit can be obtained. In Argentina, the neutron noise analysis technique is applied in customary way in the nuclear power plants Atucha I and Embalse. A database was constructed and vibration frequencies corresponding to different reactor internals were characterized. Reactor internals with particular mechanical vibrations have been detected and localized. In the framing of a cooperation project between Argentina and Germany, we participated in the measurements, analysis and modelisation, using the neutron noise technique, in the Obrigheim and Gundremmingen nuclear power plants. In the nuclear power plant Obrigheim (PWR, 350 M We), correlations between the signals measured from self-power neutron detectors and accelerometers located inside the reactor core, were made. In the nuclear power plant Gundremmingen (BWR, 1200 M We) we participated in the study of a particular mechanical vibration detected in one of the instrumentation tube. (author)

Modernization of the Almaraz, AscO & VandellOs non-1E Control systems during the last decade the Spanish PWR nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Fuente Arias, E. de la; Serrano Jimenez, J.; Madroñal Rodriguez, E.

2016-07-01

During the last decade the Spanish PWR nuclear power plants designed by Westinghouse have planned and implemented the modernization of the non-1E Control systems. The driving forces behind the modernization of the original Control Systems are the management of the obsolescence of these systems and the implementation of functional improvements in the plants to increase the Control System reliability and availability. Westinghouse Ovation platform has been used in the modernization of the Reactor Control System, Turbine Control System, Plant Computer and Feedwater Heaters Level and MSR s Drains tanks Level control. Modernizations have been spread through the years in such a way that there is not impact on the outages and the different organizations on the customer and estinghouse can have dedicated teams to work in these projects. (Author)

Elecnuc. Nuclear power plants in the world; Elecnuc. Les centrales nucleaires dans le monde

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

This small booklet summarizes in tables all the numerical data relative to the nuclear power plants worldwide. These data come from the French CEA/DSE/SEE Elecnuc database. The following aspects are reviewed: 1999 highlights; main characteristics of the reactor types in operation, under construction or on order; map of the French nuclear power plants; worldwide status of nuclear power plants at the end of 1999; nuclear power plants in operation, under construction and on order; capacity of nuclear power plants in operation; net and gross capacity of nuclear power plants on the grid and in commercial operation; grid connection forecasts; world electric power market; electronuclear owners and share holders in EU, capacity and load factor; first power generation of nuclear origin per country, achieved or expected; performance indicator of PWR units in France; worldwide trend of the power generation indicator; 1999 gross load factor by operator; nuclear power plants in operation, under construction, on order, planned, cancelled, shutdown, and exported; planning of steam generators replacement; MOX fuel program for plutonium recycling. (J.S.)

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.

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

International Nuclear Information System (INIS)

Schreiner, L.A.

1991-01-01

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

Energy and exergy analyses of Angra-2 nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Marques, João G.O.; Costa, Antonella L.; Pereira, Claubia; Fortini, Ângela, E-mail: jgabrieloliveira2010@bol.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: fortini@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Nuclear Power Plants (NPPs) based on Pressurized Water Reactors (PWRs) technology are considered an alternative to fossil fuels plants due to their reliability with low operational cost and low CO{sub 2} emissions. An example of PWR plant is Angra-2 built in Brazil. This NPP has a nominal electric power output of 1300 MW and made it possible for the country save its water resources during electricity generation from hydraulic plants, and improved Brazilian knowledge and technology in nuclear research area. Despite all these benefits, PWR plants generally have a relatively low thermal efficiency combined with a large amount of irreversibility generation or exergy destruction in their components, reducing their capacity to produce work. Because of that, it is important to assess such systems to understand how each component impacts on system efficiency. Based on that, the aim of this work is to evaluate Angra-2 by performing energy and exergy analyses to quantify the thermodynamic performance of this PWR plant and its components. The methodology consists in the development of a mathematical model in EES (Engineering Equation Solver) software based on thermodynamic states in addition to energy and exergy balance equations. According to the results, Angra 2 has energy efficiency of 36.18% and exergy efficiency of 49.24%. Reactor core is the most inefficient device in the NPP; it has exergy efficiency of 67.16% and is responsible for 63.88% of all exergy destroyed in Angra-2. (author)

PWR control rod ejection analysis with the numerical nuclear reactor

International Nuclear Information System (INIS)

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

2008-01-01

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

Preliminary study of the nuclear power option in Belarus

International Nuclear Information System (INIS)

Grusha, N.M.; Kazazyan, V.T.; Malykhin, A.P.; Mikhalevich, A.A.; Yakushau, A.P.; Yaroshevich, O.I.

1999-01-01

The Republic of Belarus possesses an economy with many energy intensive branches. At the same time the share of domestic energy resources is about 15% of total energy demand. The share of the payment for primary energy resources reaches 60% or USD 2 billion of the total energy import. That is comparable with the annual state budget. In addition to that, about half of the installed capacities have reached their operation life and 90% of the units have to be retrofitted or replaced until 2010. Thus, the problem of energy supply is one of the most important ones for Belarus' economy. The nuclear power appears to be one of the possible ways for solving the energy demand problem in Belarus which has, as in case of many countries of Central and South-Eastern Europe, limited energy resources. In 1992 - 1994 the works for studying the possibility of NPP siting were recommenced and six relatively competitive sites have been chosen out from 54 possible locations for NPP siting. Parallely, works on assessment of environmental NPP effect in these sites were carried out. As concerning the reactors to be purchased and installed in the sites selected, the following options were taken into consideration: PWR of American Company WESTINGHOUSE; PWR N4 of France Company FRAMATOME; PWR KONVOI of German Company SIEMENS. Also promising are the new generation of Russian Reactor NPP, namely NPP - 91, NPP - 92 and NPP with NGWWER - 640 reactors. Preliminary assessment having in view the feasibility characteristics, safety, reliability as well as the degree of completion shows the Russian projects NPP - 92 and NGWWER - 640 as more preferably at present. Concerning the radioactive waste management, sites for storing low and medium active waste have been determined as well as regions for high active waste disposal. At present Belarus Republic disposes of a definite production, engineering and scientific potential, which can be used when the nuclear power program will be launched. Construction

Fuzzy control applied to nuclear power plant pressurizer system

International Nuclear Information System (INIS)

Oliveira, Mauro V.; Almeida, Jose C.S.

2011-01-01

In a pressurized water reactor (PWR) nuclear power plants (NPPs) the pressure control in the primary loop is very important for keeping the reactor in a safety condition and improve the generation process efficiency. The main component responsible for this task is the pressurizer. The pressurizer pressure control system (PPCS) utilizes heaters and spray valves to maintain the pressure within an operating band during steady state conditions, and limits the pressure changes, during transient conditions. Relief and safety valves provide overpressure protection for the reactor coolant system (RCS) to ensure system integrity. Various protective reactor trips are generated if the system parameters exceed safe bounds. Historically, a proportional-integral derivative (PID) controller is used in PWRs to keep the pressure in the set point, during those operation conditions. The purpose of this study has two main goals: first is to develop a pressurizer model based on artificial neural networks (ANNs); second is to develop a fuzzy controller for the PWR pressurizer pressure, and compare its performance with the P controller. Data from a simulator PWR plant was used to test the ANN and the controllers as well. The reference simulator is a Westinghouse 3-loop PWR plant with a total thermal output of 2785 MWth. The simulation results show that the pressurizer ANN model response are in reasonable agreement with the simulated power plant, and the fuzzy controller built in this study has better performance compared to the P controller. (author)

Fuzzy control applied to nuclear power plant pressurizer system

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Mauro V.; Almeida, Jose C.S., E-mail: mvitor@ien.gov.b, E-mail: jcsa@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2011-07-01

In a pressurized water reactor (PWR) nuclear power plants (NPPs) the pressure control in the primary loop is very important for keeping the reactor in a safety condition and improve the generation process efficiency. The main component responsible for this task is the pressurizer. The pressurizer pressure control system (PPCS) utilizes heaters and spray valves to maintain the pressure within an operating band during steady state conditions, and limits the pressure changes, during transient conditions. Relief and safety valves provide overpressure protection for the reactor coolant system (RCS) to ensure system integrity. Various protective reactor trips are generated if the system parameters exceed safe bounds. Historically, a proportional-integral derivative (PID) controller is used in PWRs to keep the pressure in the set point, during those operation conditions. The purpose of this study has two main goals: first is to develop a pressurizer model based on artificial neural networks (ANNs); second is to develop a fuzzy controller for the PWR pressurizer pressure, and compare its performance with the P controller. Data from a simulator PWR plant was used to test the ANN and the controllers as well. The reference simulator is a Westinghouse 3-loop PWR plant with a total thermal output of 2785 MWth. The simulation results show that the pressurizer ANN model response are in reasonable agreement with the simulated power plant, and the fuzzy controller built in this study has better performance compared to the P controller. (author)

Load follow operation in nuclear power plants and its influence on PWR fuel behaviour

International Nuclear Information System (INIS)

Nagino, Y.; Miyazaki, Y.

1980-01-01

The contribution of nuclear power generation to our company's grid system is becoming greater each year, which makes it necessary to operate nuclear power plants with load follow mode in the near future. (author)

Development of the vacuum drying process for the PWR spent nuclear fuel dry storage

Energy Technology Data Exchange (ETDEWEB)

Baeg, Chagn Yeal; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

2016-12-15

This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP's storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process)

French PWR safety philosophy

International Nuclear Information System (INIS)

Conte, M.

1986-05-01

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

A reliability centered maintenance model applied to the auxiliary feedwater system of a nuclear power plant

International Nuclear Information System (INIS)

Araujo, Jefferson Borges

1998-01-01

The main objective of maintenance in a nuclear power plant is to assure that structures, systems and components will perform their design functions with reliability and availability in order to obtain a safety and economic electric power generation. Reliability Centered Maintenance (RCM) is a method of systematic review to develop or optimize Preventive Maintenance Programs. This study presents the objectives, concepts, organization and methods used in the development of RCM application to nuclear power plants. Some examples of this application are included, considering the Auxiliary Feedwater System of a generic two loops PWR nuclear power plant of Westinghouse design. (author)

Experience in NPP operation and prospects of nuclear power development in France

International Nuclear Information System (INIS)

Lekler, Zh.

1986-01-01

Main results and perspectives of nuclear power development in France are analysed. Data characterizing operation of NPPs with PWR reactors in France and leading capitalist countries are given. Problems related to the organization of works on control of NPP operation and its maintenance are discussed. Power utilization factor for french power units amounted to 75% in 1984. NPP share of all electric power generated in the country constituted 59%. Emergency reactor shut-downs took 5% of calendar time in 1984

Behavior of four PWR rods subjected to a simulated loss-of-coolant accient in the power burst facility

International Nuclear Information System (INIS)

Cook, T.F.; Hagrman, D.L.; Sepold, L.K.

1978-01-01

Cladding deformation characteristics resulting from the first nuclear blowdown tests (LOC-11) conducted in the Power Burst Facility (PBF) are emphasized in this paper. The objective of the LOC-11 tests was to obtain data on the thermal, mechanical, and materials behavior of pressurized and unpressurized fuel rods when exposed to a blowdown similiar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The test hardware consisted of four separately shrouded fresh fuel rods of PWR 15 x 15 design. Initial plenum pressures ranged from atmospheric to 4.8 MPa (representative of end-of-life). During LOC-11C, the four fuel rods were subjected to 6.5 hours of nuclear operation at approximately 67 kW/m average rod power to cause decay heat build-up. Just before the start of blowdown, cladding surface temperatures were about 620 K and fuel centerline temperatures were in the 2500 to 2600 K range. During the 30-second blowdown transient, CHF occurred 2 seconds after initiation. Fuel centerline temperature dropped continuously, while cladding surface temperatures increased. Maximum cladding temperatures of 1030 to 1050 K occurred 15 seconds into the transient. Posttest destructive examination revealed cladding microstructures and oxide thicknesses consistent with the measured cladding temperatures. The cladding surface thermocouples did not appreciably affect cladding temperature distributuion (fin cooling effect) in the vicinity of the thermocouples

A realistic approach to nuclear power for tomorrow

International Nuclear Information System (INIS)

Knapp, R.W.; Pedersen, T.J.

1993-01-01

The future global role of nuclear power is determined by a puzzling set of considerations. Like any other substantial contributor to the world's energy needs, nuclear power must be generated at an acceptable cost and with a negligible effect on the environment. Furthermore, it must achieve and maintain a socially reasonable level of public acceptance, which in turn is not necessarily governed by rational assessments of the true safety and environmental impact of nuclear power plants. ABB's approach to this situation can be best characterized as a open-quotes cautious evolutionclose quotes-step by step improvements rather dramatic design changes. For the next decade(s), ABB will therefore largely base its offerings to the market on open-quotes evolutionsclose quotes of its successful versions of the two reactor types that represent the overwhelming portion of installed generating capacity world-wide, namely with BWR and the PWR. In addition to these, and owing to the extensive lead time that relate to the development, verification and licensing of distinctly new reactor concepts, ABB continues to develop designs of the open-quotes passiveclose quotes type in preparation of future generations. Accordingly, the paper will elaborate on ABB's current reactor designs-the BWR 90 and the System 80+ TM PWR plants. Both can be designed, licensed and constructed in accordance with any safety regulations now in force or envisaged in the Western world. Emphasis has been, and will be, placed on features that aim at ensuring predictability with regard to licensing and, hence, construction time and corresponding electricity costs. An overview of the status of ABB's open-quotes passiveclose quotes PIUS system will also be presented

A realistic approach to nuclear power for tomorrow

International Nuclear Information System (INIS)

Knapp, R.W.; Pedersen, T.J.

1993-01-01

The future global role of nuclear power is determined by a puzzling set of considerations. Like any other substantial contributor to the world's energy needs, nuclear power must be generated at an acceptable cost and with a negligible effect on the environment. Furthermore, it must achieve and maintain a socially reasonable level of public acceptance, which in turn is not necessarily governed by rational assessments of the true safety and environmental impact of nuclear power plants. ABB's approach to this situation can be best characterized as a open-quotes cautious evolutionclose quotes-step by step improvements rather dramatic design changes. For the next decade(s), ABB will therefore largely base its offerings to the market on open-quotes evolutionclose quotesof its successful versions of the two reactor types that represent the overwhelming portion of installed generating capacity world-wide, namely the BWR and PWR. In addition to these, and owing to the extensive lead time that relate to the development, verification and licensing of distinctly new reactor concepts, ABB continues to develop designs of the open-quotes passiveclose quotes type in preparation for future generations. Accordingly, the paper will elaborate on ABB's current reactor designs the BWR 90 and the System 80*TM PWR plants. Both can be designed, licensed and constructed in accordance with any safety regulations now in force or envisaged in the Western world. Emphasis has been, and will be, placed on features that aim at ensuring predictability with regard to licensing and, hence, construction time and corresponding electricity generation costs. An overview of the status of ABB's open-quotes passiveclose quotes PIUS system will also be presented

Elecnuc. Nuclear power plants in the world; Elecnuc. Les centrales nucleaires dans le monde

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

This 2003 version of Elecnuc contents information, data and charts on the nuclear power plants in the world and general information on the national perspectives concerning the electric power industry. The following topics are presented: 2002 highlights; characteristics of main reactor types and on order; map of the French nuclear power plants; the worldwide status of nuclear power plants on 2002/12/3; units distributed by countries; nuclear power plants connected to the Grid by reactor type groups; nuclear power plants under construction; capacity of the nuclear power plants on the grid; first electric generations supplied by a nuclear unit; electrical generation from nuclear plants by country at the end 2002; performance indicator of french PWR units; trends of the generation indicator worldwide from 1960 to 2002; 2002 cumulative Load Factor by owners; nuclear power plants connected to the grid by countries; status of license renewal applications in Usa; nuclear power plants under construction; Shutdown nuclear power plants; exported nuclear power plants by type; exported nuclear power plants by countries; nuclear power plants under construction or order; steam generator replacements; recycling of Plutonium in LWR; projects of MOX fuel use in reactors; electricity needs of Germany, Belgium, Spain, Finland, United Kingdom; electricity indicators of the five countries. (A.L.B.)

Transient evaluation using EMTP at single open phase with the offsite power transformer for the emergency power supply systems of nuclear power plants

International Nuclear Information System (INIS)

Shimada, Yoshio

2017-01-01

The emergency power supply systems of nuclear power plants, as the objects of this research, are critical in supplying stable electric power to such systems as the emergency core cooling system (ECCS), and in maintaining safety of the nuclear power reactor; this was apparent in the accident at the Fukushima Daiichi Nuclear Power Station. The Nuclear Regulatory Commission (USNRC) issued regulatory documents (BL 2012-011, IN 2012-032), and has commenced evaluations on newly discovered vulnerability in the design of power supply systems which cannot be detected with under-voltage protection relays, with certain kinds of configuration of coils and iron core structures, such as when the offsite power supply side is a Y-connection and the load side is a ⊿-connection etc., when the detection of single open phase fault with the circuit of a transformer which is without a ground fault connected to the offsite power supply system. This report uses simulation by the electro magnetic transients program (EMTP) and clearly describe the response at the time of the power supply single open phase without ground fault for various configuration of coils and various iron core structures of the three-phase transformer, and identify the important issues in the response of emergency power supply systems and the safety related components of representative domestic PWR plants when the single open phase fault occurred without ground fault. This report describes the results of the simulations of operations of the protection relays of the emergency power supply systems and the safety related components of representative a domestic PWR plant with EMTP. This report explains the method to detect open-phase when the transformer is no-load which United States Electric Power Research Institute (EPRI) developed. As the detailed analyses data from EPRI related to the detection method concerned have not been disclosed officially yet, in this paper, the quantitative and detailed verification results

Corrective maintenance support system for nuclear power plants

International Nuclear Information System (INIS)

Kakiuchi, Tetsuo

1996-01-01

With increase of share of nuclear power generation in electric power supply in Japan, requirement for further safe operation and improvement of economics for the nuclear power plants is promoting. The pressed water type (PWR) nuclear power plant in operation in Japan reaches to 22 sets, application rate of the instruments is 74% as mean value for 7 years since 1989 and in high level, which is due to a result of preventive maintenance in ordinary and periodical inspections. The present state of maintenance at the nuclear power plant is mainly preventive maintenance, which is mainly conducted in a shape of time planning maintenance but partially in a shape of state monitoring maintenance for partial rotating appliances. Concretely speaking, the periodical inspection was planned on a base of daily inspection and a long term program on maintenance, and executed on a base of feedback function to think of the long term program again by evaluating the periodical inspection results. Here were introduced on the monitoring diagnosis and periodical inspection regionalization equipment, fatigue monitoring system, automatic supersonic wave damage inspection equipment for reactor, steam evaporator heat conductive tube inspection equipment, automatic testing equipment for measuring controller, air working valve property testing equipment, as maintaining support system in the PW generation plant. (G.K.)

Clean nuclear power (2. part)

International Nuclear Information System (INIS)

Rocherolles, R.

1998-01-01

The 450 nuclear power plants which produce 24% of world electricity do not generate greenhouse gas effects, but 8,000 tonnes per year of irradiated, radioactive fuel. The first article which was published in the July-August 1997 issue of this journal, described the composition and management of these fuels. This article wish to show the advantage of 'advanced re-processing', which would separate fission products from actinides, in order to incinerate them separately in dedicated fuels and reactors, which, from an ecological point of view, seems more efficient than burying them underground in deep, geological layers. To rid the planet of waste which is continuing to build up, the first step is to build 'incinerators' which will eliminate fission products by slow neutron assisted neutronic capture, and actinides by fast neutron assisted fission. Various projects have been set up, in particular, in Los Alamos, Japan and the CERN. The Carlo Rubbia hybrid machine operating on the well-known thorium cycle is the most advanced project. An incinerator connected up to standard PWR reactor produces no actinide, and reduces the existing stock of plutonium. However, the proper solution, obviously, is to no longer produce waste along with power; second generation nuclear fission will do this. The CERN team bas studied a clean reactor, producing practically no actinides, or fission products, more or less. Thus, the solution to the problem of waste is at hand, and nuclear power will be cleaner that all other types of power. The world market opening up to clean nuclear power is about 1,300 Gigawatts, or 1,300 plants of 1,000 Megawatts. Remarkable progress is taking place under our very eyes; soon we will have clean power in sufficient quantities, at a lower cost than that of other forms of power. (authors)

On the performance of an artificial bee colony optimization algorithm applied to the accident diagnosis in a PWR nuclear power plant

International Nuclear Information System (INIS)

Oliveira, Iona Maghali S. de; Schirru, Roberto; Medeiros, Jose A.C.C.

2009-01-01

The swarm-based algorithm described in this paper is a new search algorithm capable of locating good solutions efficiently and within a reasonable running time. The work presents a population-based search algorithm that mimics the food foraging behavior of honey bee swarms and can be regarded as belonging to the category of intelligent optimization tools. In its basic version, the algorithm performs a kind of random search combined with neighborhood search and can be used for solving multi-dimensional numeric problems. Following a description of the algorithm, this paper presents a new event classification system based exclusively on the ability of the algorithm to find the best centroid positions that correctly identifies an accident in a PWR nuclear power plant, thus maximizing the number of correct classification of transients. The simulation results show that the performance of the proposed algorithm is comparable to other population-based algorithms when applied to the same problem, with the advantage of employing fewer control parameters. (author)

On the performance of an artificial bee colony optimization algorithm applied to the accident diagnosis in a PWR nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Iona Maghali S. de; Schirru, Roberto; Medeiros, Jose A.C.C., E-mail: maghali@lmp.ufrj.b, E-mail: schirru@lmp.ufrj.b, E-mail: canedo@lmp.ufrj.b [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

2009-07-01

The swarm-based algorithm described in this paper is a new search algorithm capable of locating good solutions efficiently and within a reasonable running time. The work presents a population-based search algorithm that mimics the food foraging behavior of honey bee swarms and can be regarded as belonging to the category of intelligent optimization tools. In its basic version, the algorithm performs a kind of random search combined with neighborhood search and can be used for solving multi-dimensional numeric problems. Following a description of the algorithm, this paper presents a new event classification system based exclusively on the ability of the algorithm to find the best centroid positions that correctly identifies an accident in a PWR nuclear power plant, thus maximizing the number of correct classification of transients. The simulation results show that the performance of the proposed algorithm is comparable to other population-based algorithms when applied to the same problem, with the advantage of employing fewer control parameters. (author)

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)

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

International Nuclear Information System (INIS)

Schwirian, R.; Robison, G.

1994-07-01

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

Investigation on the use digital controls instead of PID analog controls in the level control of steam generators of nuclear power PWR

International Nuclear Information System (INIS)

Alvarenga, Marco Antonio Bayout

2012-01-01

The aim of this study is to identify current alternatives for the implementation of digital controllers in the level control of steam generators of nuclear power PWR (Pressurized Water Reaetor). It is intended to identify the types of digital controls that are available from the theoretical and conceptual viewpoints for this purpose. We investigate the advantages and disadvantages of each controller model. From this assessment are pointed the most suitable models in hierarchical scale. This evaluation also serves to suggest possible types of control installation as a whole, where the level control of the steam generators becomes just one of many controls that are part of the plant. In this case, the use of digital controls allows the non-linear and multivariable treatment which is characteristic of complex systems, such as the nuclear power generation. The treatment of nonlinearities and multivariable aspects allows a more detailed study of the stability of these plants when they are subject to transients or several accidents, such as the case of losing external power of transients. In the specific case of steam generators, the instabilities result from the emergence of the shrink and swell phenomenas, depending on the load variations of thermonuclear plant. The application of several types and digital controllers, considering these inherent characteristics of the level control of steam generators, allows to infer which types of controllers are more appropriate to treat instabilities of this type and to make conjectures in its use for the cases of more complex instabilities, considering the integration of all nucleus-plant controls.

Seismic re-evaluation of French nuclear power plants

International Nuclear Information System (INIS)

Andrieu, R.

1995-01-01

After a presentation of the seismic inputs which have been taken into account in the design of the French Nuclear Power Plants, the re-assessed values of these inputs are shown. Some considerations about the specificity of the French PWR program with regard to the standardisation of plants are given together with the present objectives of seismic re-evaluations. Finally the main results of the seismic re-analysis being performed for the Phenix Fast Reactor are considered. (author)

An engineer-constructor's view of nuclear power plant safety

International Nuclear Information System (INIS)

Landis, J.W.; Jacobs, S.B.

1984-01-01

At SWEC we have been involved in the development of safety features of nuclear power plants ever since we served as the engineer-constructur for the first commerical nuclear power station at Shippingport, Pennsylvania, in the 1950s. Our personnel have pioneered a number of safety innovations and improvements. Among these innovations is the subatmospheric containment for pressurized water reactor (PWR) power plants. This type of containment is designed so that leakage will terminate within 1 to 2 hours of the worst postulated loss of coolant accident. Other notable contributions include first use of reinforced-concrete atmospheric containments for PWR power plants and of reinforced-concrete, vapor-suppression containments for boiling water reactor (BWR) power plants. Both concepts meet rigorous U.S. safety requirements. SWEC has performed a substantial amount of work on developing standardized plant designs and has developed standardized engineering and construction techniques and procedures. Standardization concepts are being developed in Canada, France, USSR, and Germany, as well as in the United States. The West German convoy concept, which involves developing a number of standardized plants in a common effort, has been quite successful. We believe standardization contributes to safety in a number of ways. Use of standardized designs, procedures, techniques, equipment, and methods increases efficiency and results in higher quality. Standardization also reduces the design variations with which plant operators, emergency teams, and regulatory personnel must be familiar, thus increasing operator capability, and permits specialized talents to be focused on important safety considerations. (orig./RW)

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

Examination of turbine discs from nuclear power plants

International Nuclear Information System (INIS)

Czajkowski, C.J.; Weeks, J.R.

1982-01-01

Investigations were performed on a cracked turbine disc from the Cooper Nuclear Power Station, and on two failed turbine discs (governor and generator ends) from the Yankee-Rowe Nuclear Power Station. Cooper is a boiling water reactor (BWR) which went into commercial operation in July 1974, and Yankee-Rowe is a pressurized water reactor (PWR) which went into commercial operation in June 1961. Cracks were identified in the bore of the Cooper disc after 41,913 hours of operation, and the disc removed for repair. At Yankee-Rowe two discs failed after 100,000 hours of operation. Samples of the Cooper disc and both Yankee-Rowe disc (one from the governor and one from the generator end of the LP turbine) were sent to Brookhaven National Laboratory (BNL) for failure analysis

Analytical technical of lightning surges induced on grounding mesh of PWR nuclear power plant

International Nuclear Information System (INIS)

Ikeda, I.; Tani, M.; Yonezawa, T.

1990-01-01

An analytical lightning surge technique is needed to make a qualitative and predictive evaluation of transient voltages induced on local grounding meshes and instrumentation cables by a lightning strike on a lightning rod in a PWR plant. This paper discusses an experiment with lightning surge impulses in a PWR plant which was setup to observe lightning caused transient voltages. Experimental data when compared with EMTP simulation results improved the simulation method. The improved method provides a good estimation of induced voltages on grounding meshes and instrumentation cables

Development of halogen-free flame-retardant cable for nuclear power plant. 2

International Nuclear Information System (INIS)

Matsumoto, Tetsuo; Kimura, Hitoshi; Ishii, Nobuhisa

1997-01-01

Halogen-free flame-retardant cables were developed for PWR nuclear power stations. It was confirmed that the developed cables possess flame retardant property, corrosion resistance, low toxicity and low smoke generation, and withstand the normal operation in the environment in PWR containment vessels for 60 years and loss of coolant accident. In the advancement of LWR technology, it is important to improve the reliability of machinery and equipment, to extend the period of continuous operation, to optimize the operation cycle and to improve the maintenance of plants. By improving halogen-free flame-retardant material and applying it to the cables for nuclear power stations, it can contribute to the above purposes. The required characteristics of these cables are explained, and the targets of development are power cables, control cables, instrumentation cables and insulated wires which do not contain halogen. The basic material is polyolefin, in which flame retardant magnesium hydroxide and the agent for improving radiation resistance are mixed. The corrosive property and toxicity of gases, smoke generation and the prevention of spread of flame when the cables burn and the durability in environment were evaluated. (K.I.)

The fourth nuclear power plant in Argentina

Energy Technology Data Exchange (ETDEWEB)

Nogarin, Mauro

2017-01-15

Since 2006 the nuclear sector in Argentina has aimed at recovering and strengthening its capabilities and facilities. Part of the challenge posed by this revival has been to also accompany the development of activities with a higher level of responsibility for safety and the environment. Among the strategic decisions taken in recent years, one main highlight is the construction of the nuclear power plant CAREM25 entirely with Argentine technology and design under the responsibility of the National Atomic Energy Commission. On February 4, 2015, the Ministry of Federal Planning and the National Energy Administration (NEA) signed the agreement for cooperation and construction of pressurized water reactor (PWR) with ACP-1000 technology, developed in the Peoples Republic of China.

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)

Methodology for safety classification of PWR type nuclear power plants items

International Nuclear Information System (INIS)

Oliveira, Patricia Pagetti de

1995-01-01

This paper contains the criteria and methodology which define a classification system of structures, systems and components in safety classes according to their importance to nuclear safety. The use of this classification system will provide a set of basic safety requirements associated with each safety class specified. These requirements, when available and applicable, shall be utilized in the design, fabrication and installation of structures, systems and components of Pressurized Water Reactor Nuclear Power Plants. (author). 13 refs, 1 tab

The role of NDT in nuclear power development in Pakistan

International Nuclear Information System (INIS)

Asghar Ali Khan; Sabir Choudhary, M.; Arif Iftikhar, M.; Afaque, A. S.; Yousaf Raza Zaidi, S.

2003-01-01

Pakistan has two operating nuclear power plants namely, Karachi Nuclear Power Plant (KANUPP) which is 137 MW Candu type Canadian reactor using natural uranium fuel and the Chashma Nuclear Power Plant (CHASNUPP) which is a 300 MW PWR type Chinese built reactor using 3% enriched uranium fuel. A third nuclear power plant is being negotiated for construction. This would most probably be the twin unit of CHASNUPP and the construction might begin early next year.Non destructive testing (NDT) has an important role in the development and safe operation of the nuclear power plants by providing the Pre-Service Inspection (PSI) services during the manufacturing and installation phase, and the In-Service Inspection (ISI) services during the operation and maintenance phase. ISI of various components of nuclear power plants is an essential activity which has to be carried out either on emergency basis on as and when required basis or periodically at regular intervals described in the quality assurance QA manuals of the plant. There are numerous components and systems in the nuclear power plants working together. The failure of one system affects the performance of the whole plant. There are two main divisions, called the Nuclear Island and Conventional Island. Main components of Nuclear Island are reactor pressure vessel, reactor core, steam generators, pressurizer, primary coolant pumps and primary piping, etc. and the main components in Conventional Island are turbine, condensers, pre-heaters, moisture separators, secondary heat treatment system and piping etc. (Author)

Operator training simulator for nuclear power plant

International Nuclear Information System (INIS)

Shiozuka, Hiromi

1977-01-01

In nuclear power plants, training of the operators is important. In Japan, presently there are two training centers, one is BWR operation training center at Okuma-cho, Fukushima Prefecture, and another the nuclear power generation training center in Tsuruga City, Fukui Prefecture, where the operators of PWR nuclear power plants are trained. This report describes the BWR operation training center briefly. Operation of a nuclear power plant is divided into three stages of start-up, steady state operation, and shut down. Start-up is divided into the cold-state start-up after the shut down for prolonged period due to periodical inspection or others and the hot-state start-up from stand-by condition after the shut down for a short time. In the cold-state start-up, the correction of reactivity change and the heating-up control to avoid excessive thermal stress to the primary system components are important. The BWR operation training center offers the next three courses, namely beginner's course, retraining course and specific training course. The training period is 12 weeks and the number of trainees is eight/course in the beginner's course. The simulator was manufactured by modeling No. 3 plant of Fukushima First Nuclear Power Station, Tokyo Electric Power Co. The simulator is composed of the mimic central control panel and the digital computer. The software system comprises the monitor to supervise the whole program execution, the logic model simulating the plant interlock system and the dynamic model simulating the plant physical phenomena. (Wakatsuki, Y.)

1994 - Starting a new era for nuclear power in Great Britain

International Nuclear Information System (INIS)

Collier, J.G.

2004-01-01

A gap is opening up between what the world would like to consume in terms of energy and what the world's environment can sustain. To bridge this gap a much greater contribution from nuclear and renewables is needed to limit the emission of CO 2 and other harmful gases. Plans already laid in some Pacific Rim countries recognise this need and nuclear energy is expected to grow rapidly in this region over the next 20-30 years. For Pacific Rim countries without plentiful indigenous fossil fuel resources, nuclear power is the most economic form of electricity generation, minimising the need for expensive imports and strengthening security of energy supply. Nuclear power already makes a significant contribution to reducing emissions in other regions, especially Europe. Well over a third of electricity in Europe comes from nuclear power. In the UK the figure is nearer a quarter and rising. If nuclear power is to fulfil its role as a key part of UK energy supply into the 21st century, we need to maintain our existing capacity as our older plant retires by building further, modern PWRs. Central to our future is Sizewell B power station - the UK's first PWR. Construction is complete and commissioning is well under way. The project is nearing completion - ahead of time and under budget. The success of Sizewell B is the foundation for the future of nuclear power in the UK. Stations based on the Sizewell design will become the workhorses of our utility. Already Nuclear Electric has submitted a planning application to construct Sizewell C. This is a technical replica of Sizewell B in twin reactor form with a net output of around 2600MW. And this design also has strong export potential. Westinghouse and Nuclear Electric have submitted a joint-vender to build a PWR here in Taiwan. Our bid is based on the Sizewell design; an innovative, safe and proven nuclear power plant design for supplying reliable low-cost electricity for the Republic of China's growing economy. The Sizewell

Field vibration test of principal equipment of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Shiraki, Kazuhiro; Fujita, Katsuhisa; Kajimura, Motohiko; Ikegami, Yasuhiko; Hanzawa, Katsumi; Sakai, Yoshiyuki; Kokubo, Eiji; Igarashi, Shigeru

1984-09-01

Japan is one of the most earthquake-stricken countries in the world, and demands for aseismic design have become severer recently. In a nuclear power plant in particular, consisting of a reactor vessel and other facilities dealing with a radioactive substance in some form or other, it is essential from the standpoint of safety to eliminate any possibility of radioactive hazards for the local public, and the employees at the plant as well, if these facilities are struck by an earthquake. This paper is related to the reactor vessel, reactor primary cooling equipment and piping system and important general piping as examples of important facilities of a nuclear power plant, and discusses vibration tests of an actual plant in the field from the standpoint of enhancing the aseismic safety of the Mitsubishi PWR nuclear power plant. Especially concerning vibration test technology, the effects in the evaluation of aseismic safety and its limits are studied to prove how it contributes to the enhancement of the reliability of aseismic design of nuclear power plants.

Definition of thermal-hydraulics parameters of a naval PWR via energy balance of a Westinghouse PWR

Energy Technology Data Exchange (ETDEWEB)

Chaves, Luiz C.; Curi, Marcos F., E-mail: marcos.curi@cefet-rj.br [Centro Federal de Educação Tecnológica Celso Suckow da Fonseca (CEFET-RJ), Rio de Janeiro, RJ (Brazil). Department of Mechanical Engineering

2017-07-01

In this work, we used the operational parameters of the Angra 1 nuclear power plant, designed by Westinghouse, to estimate the thermal-hydraulic parameters for naval nuclear propulsion, focusing on the analysis of the reactor and steam generator. A thermodynamics analysis was made to reach the operational parameters of primary circuit such as pressure, temperature, power generated among others. Previous studies available in literature of 2-loop Westinghouse Nuclear Power Plants, which is based on a PWR and similar to Angra-1, support this analysis in the sense of a correct procedure to deal with many complex processes to energy generation from a nuclear source. Temperature profiles in reactor and steam generator were studied with concepts of heat transfer, fluid mechanics and also some concepts of nuclear systems, showing the behavior into them. In this simulation, the Angra 1 primary circuit was reduced on a scale of 1: 3.5 to fit in a Scorpène-class submarine. The reactor generates 85.7 MW of total thermal power. The maximum power and temperatures reached were lower than the operational safe limits established by Westinghouse. The number of tubes of the steam generator was determined in 990 U-tubes with 6.3 m of average length. (author)

Fire protection devices in the controlled region of GKN nuclear power station

International Nuclear Information System (INIS)

Bernhardt, S.; Grauf, E.

1976-01-01

In the GKN nuclear power station ('Neckar reactor'), an 805 MW PWR reactor whose start-up is scheduled for the near future, fire protection measures have been realized that go far beyond those realized in other German nuclear power stations until now. One of the main reasons is that the authorities have been sensibilized by a fire in the refuelling cavity during construction and by the Browns Ferry fire and are therefore extremely thorough in their examination. Further subsections have been added to the fire prevention sections in order to provide better quenching devices for potential fire sites. (orig./AK) [de

PWR and BWR light water reactor systems in the USA and their fuel cycle

International Nuclear Information System (INIS)

Crawford, W.D.

1977-01-01

Light water reactor operating experience in the USA can be considered to date from the choice of the pressurized water reactor (PWR) for use in the naval reactor program and the subsequent construction and operation of the nuclear power plant at Shippingport, Pennsylvania in 1957. The development of the boiling water reactor (BWR) in 1954 and its selection for the plant at Dresden, Illinois in 1959 established this concept as the other major reactor type in the US nuclear power program. The subsequent growth profile is presented, leading to 31 PWR's and 23 BWR's currently in operation as well as to plants in the planning and construction phase. A significant operating record has been accumulated concerning the availability of each of these reactor types as determined by: (1) outage for refueling, (2) component reliability, (3) maintenance requirements, and (4) retrofitting required by government regulation. In addition, the use and performance of BWR's and PWR's in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to assure effective safeguards at nuclear power installations. Safeguards measures currently in place are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring to verify that results are within the limits established in the licensing process. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. The PWR and BWR Fuel Cycle is examined in terms of: (1) fuel burnup experience and prospects for improvement, (2) the status and outlook for natural uranium resources, (3) enrichment capacity, (4) reprocessing and recycle, and the interrelationships among the latter three factors. High level waste management currently involving on-site storage of spent fuel is discussed

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)

Evaluations on power ramp data of PWR fuels by FROST and THERMOST codes

International Nuclear Information System (INIS)

Murai, K.; Ogawa, S.; Nuno, H.; Kondo, Y.

1987-01-01

An evaluation is presented of power ramp data of Mitsubishi's PWR fuel rods tested in R-2, Studsvik, which was analysed by FROST and THERMOST codes. The analyses give good predictions for measured diameter changes and on-power rod elongations. The work indicates that FROST is capable of analysing both radial and axial pellet-cladding mechanism interaction (PCMI) appropriately, and that predicted states of PCMI (i.e. stress and strain which cannot be measured directly) are considered to be reliable. The ramp data used in the present analyses were obtained in two joint programmes with five Japanese PWR utilities (KEPCO, KYEPCO, SEPCO, HEPCO, and JAPCO). (UK)

A Study on Structured Simulation Framework for Design and Evaluation of Human-Machine Interface System -Application for On-line Risk Monitoring for PWR Nuclear Power Plant-

International Nuclear Information System (INIS)

Zhan, J.; Yang, M.; Li, S.C.; Peng, M.J.; Yan, S.Y.; Zhang, Z.J.

2006-01-01

The operators in the main control room of Nuclear Power Plant (NPP) need to monitor plant condition through operation panels and understand the system problems by their experiences and skills. It is a very hard work because even a single fault will cause a large number of plant parameters abnormal and operators are required to perform trouble-shooting actions in a short time interval. It will bring potential risks if operators misunderstand the system problems or make a commission error to manipulate an irrelevant switch with their current operation. This study aims at developing an on-line risk monitoring technique based on Multilevel Flow Models (MFM) for monitoring and predicting potential risks in current plant condition by calculating plant reliability. The proposed technique can be also used for navigating operators by estimating the influence of their operations on plant condition before they take an action that will be necessary in plant operation, and therefore, can reduce human errors. This paper describes the risk monitoring technique and illustrates its application by a Steam Generator Tube Rupture (SGTR) accident in a 2-loop Pressurized Water Reactor (PWR) Marine Nuclear Power Plant (MNPP). (authors)

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.)

Seismic resistance design of nuclear power plant building structures in Japan

International Nuclear Information System (INIS)

Kitano, Takehito

1997-01-01

Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

Seismic resistance design of nuclear power plant building structures in Japan

Energy Technology Data Exchange (ETDEWEB)

Kitano, Takehito [Kansai Electric Power Co., Inc., Osaka (Japan)

1997-03-01

Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

Status report on the application of process noise technique in nuclear power plants

International Nuclear Information System (INIS)

Espefaelt, R.; Aakerhielm, F.

1979-09-01

The report gives a survey of applications of noise technique reported for nuclear power plants. The scope has been limited to areas of interest for BWR and PWR plants of the types found in Sweden and with an emphasis on cases where the practical applicability has been clearly demonstrated. (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

Workers doses in central European PWR NPPs

International Nuclear Information System (INIS)

Janzekovic, H.; Krizman, M.

2003-01-01

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

Life extension, power upgrade, and return to service work for Pickering NGS and other PWR and CANDU plants

International Nuclear Information System (INIS)

Millman, J.; Idvorian, N.; Schneider, W.

2002-01-01

Work on life extension, power upgrade and return to service has been performed and is in progress for a number of PWR and CANDU plants. For PWR plants, power upgrade work has been done for the new replacement steam generators in several cases. This work consists of redoing the formal equipment qualification analysis and reports for the uprated operating conditions to support the application for license adjustment. Life extension assessments have been performed for several CANDU plants. These are highly detailed assessments in which the particular steam generator is reassessed part by part as to the ability of each to sustain full life operation and also extended life operation. Return to service work for Pickering NGSA specifically has included this type of assessment and also specific repair, cleaning and retrofit activities including secondary side inspection, waterlancing, divider plate repair, eddy current inspection, etc. Steam generator modifications and retrofit work have been performed in a number of cases. The paper discusses various life extension, power upgrade, equipment modification and return to service activities all of which are part of the renewed drive in the industry to realise the full potential of nuclear plants by getting more and better performance from the extended service of existing plants. (author)

The prospects for nuclear power in the UK. Conclusions of the Government's nuclear review

International Nuclear Information System (INIS)

1995-05-01

The conclusions of the United Kingdom government's review of the nuclear industry in Britain were presented to Parliament in May 1995. The provision of public sector support for a new nuclear power station is deemed unwarranted against the background of the current electricity market. In reaching this conclusion the government considered possible environmental and strategic advantages, the question of diversity of fuel sources, and wider economic benefits. It is intended to privatize parts of Nuclear Electric and Scottish Nuclear in 1996 as subsidiaries of a single holding company to take over the UK's AGR and PWR stations together with a significant level of their associated liabilities. A publicly owned company will continue to run the magnox stations and retain their liabilities. British Nuclear Fuels Ltd (BNFL) will continue to offer nuclear fuel services and to receive government support in developing business in overseas markets. Since their formation in 1990, Nuclear Electric and Scottish Nuclear have improved their financial performance significantly. At privatisation the nuclear component of the fossil fuel levy will cease to be paid to Nuclear Electric and Scottish Nuclear will no longer receive the current premium price paid under the Nuclear Energy Agreement. The current regulatory regime and rigorous safety standards for nuclear power will remain substantially unchanged. (UK)

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)

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)

A new advanced software platform for nuclear power plant process information systems

International Nuclear Information System (INIS)

Sorsa, A.

1993-01-01

In the late 80s, ABB Stromberg Power Ltd. started the development of a new generation software platform for the power plant Process Information System (PIS). This development resulted in a software platform called Procontrol PMS. Procontrol PMS is a platform for fully distributed systems which provides the following features: distributed data processing, non-stop architecture, low-cost incremental expansion path, open network architecture, high functionality, effective application development environment, and advanced user interface services. A description of the structure of the Procontrol PMS software is given. ABB has received by May 1992 six orders for nuclear power plant PISs based on Procontrol PMS (4 for PWR plants, 2 for BWRs). The first Procontrol PMS based nuclear power plant PIS was commissioned in 1989 at the Loviisa nuclear power plant and has been running with 100% availability since the commissioning. (Z.S.) 2 figs

Simulation of dynamic response of nuclear power plant based on user-defined model in PSASP

International Nuclear Information System (INIS)

Zhao Jie; Liu Dichen; Xiong Li; Chen Qi; Du Zhi; Lei Qingsheng

2010-01-01

Based on the energy transformation regularity in physical process of pressurized water reactors (PWR), PWR NPP models are established in PSASP (Power System Analysis Software Package), which are applicable for calculating the dynamic process of PWR NPP and power system transient stabilization. The power dynamic characteristics of PWR NPP is simulated and analyzed, including the PWR self-stability, self-regulation and power step responses under power regulation system. The results indicate that the PWR NPP can afford certain exterior disturbances and 10%P n step under temperature negative feedbacks. The regulate speed of PWR power can reach 5%P n /min under the power regulation system, which meets the requirement of peak regulation in Power Grid. (authors)

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

Kemeny commission report. Comments by NPC (Nuclear Power Company)

Energy Technology Data Exchange (ETDEWEB)

1979-01-01

Following the accident at Three Mile island in March 1979 the Nuclear Power Company (NPC) has made evaluations of the available information and carried out preliminary investigations of the implications of the accident for all UK AGR's and for the PWR design. The results are summarized. Recommendations of the President's Commission are discussed with especial reference to those under the following headings: (1) The utility and its suppliers. (2) Training of operating personnel. (3) Technical assessment. It is concluded that while many of the relevant recommendations in the report of the President's Commission are already satisfied by UK practices for nuclear safety, and are being applied by NPC in their current programmes of work on AGR's and PWR's it is realised that it is of the utmost importance that complacent attitudes to safety are not allowed to develop. The accident at TMI and report of the Commision have provided a sharp reminder of the necessity to maintain unremitting vigilance in all aspects of design, construction and operation which are involved with the safety of the public and operators.

An estimation of population doses from a nuclear power plant during normal operation

International Nuclear Information System (INIS)

Nowicki, K.

1975-07-01

A model is presented for estimation of the potential submersion and inhalation radiation doses to people located within a distance of 1000 km from a nuclear power plant during normal operation. The model was used to calculate doses for people living 200-1000 km from hypothetical nuclear power facility sited near the geographical centre of Denmark. Two kinds of sources are considered for this situation: - unit release of 15 isotopes of noble gases and iodines, - effluent releases from two types of 1000 MWe Light Water Power Reactors: PWR and BWR. Parameter variations were made and analyzed in order to obtain a better understanding of the mechanisms of the model. (author)

Plant computer system in nuclear power station

International Nuclear Information System (INIS)

Kato, Shinji; Fukuchi, Hiroshi

1991-01-01

In nuclear power stations, centrally concentrated monitoring system has been adopted, and in central control rooms, large quantity of information and operational equipments concentrate, therefore, those become the important place of communication between plants and operators. Further recently, due to the increase of the unit capacity, the strengthening of safety, the problems of man-machine interface and so on, it has become important to concentrate information, to automate machinery and equipment and to simplify them for improving the operational environment, reliability and so on. On the relation of nuclear power stations and computer system, to which attention has been paid recently as the man-machine interface, the example in Tsuruga Power Station, Japan Atomic Power Co. is shown. No.2 plant in the Tsuruga Power Station is a PWR plant with 1160 MWe output, which is a home built standardized plant, accordingly the computer system adopted here is explained. The fundamental concept of the central control board, the process computer system, the design policy, basic system configuration, reliability and maintenance, CRT display, and the computer system for No.1 BWR 357 MW plant are reported. (K.I.)

ROX PWR

Energy Technology Data Exchange (ETDEWEB)

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

1999-12-01

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

The optimum fuel and power distribution for a PWR burnup cycle

International Nuclear Information System (INIS)

Stillman, J.A.

1989-01-01

A method was developed to determine the optimum fuel and power distributions for a PWR burnup cycle. The backward diffusion calculation [1] and the Core-wise Green's Function [2] method were used for the core model which provided analytic derivatives for solving the nonlinear optimization problem using successive linear programming [3] methods. The solution algorithm consisted of a reverse depletion strategy which begins at the end of cycle and solves simultaneously for the optimal fuel and burnable absorber distributions while the core is depleted to the beginning of cycle. The resulting optimal solutions minimize the required fissile fuel inventory and burnable absorber loading for a PWR

Influence of Egyptian electrical grid and nuclear power plants under disturbances based on PSS/E

International Nuclear Information System (INIS)

Shaat, M. K.; Kotb, S. A.; Mahmoud, H. M.

2012-12-01

The capacity of the electrical power system in Egypt will increase rapidly in the coming twenty years. In year 2018, power generation will be connecting to the Egyptian electrical grid. Consequently, the interaction of nuclear power plants and other systems become a very important issue, and a detailed nuclear power model for the medium-term and long-term power system stability should be developed. However, there is no nuclear unit model that can describe the detailed characteristics of the nuclear unit in the available commercial power system simulation software. In this paper, a detailed pressurized water reactor (PWR) nuclear unit model for medium-term and long-term power system transient stability is proposed. The model is implemented by a user defined program in PSS/E through PSS/E Mat lab Seamanlike interface. Also this paper proposes a design of power plant rector controller for the nuclear power plant. This model can be used to analyze the difference influences between the Egyptian electrical grid and nuclear power plants for examples transient fault on electrical grid and outage of nuclear power plant. The simulation results show that the proposed model is valid. (Author)

Development of control room design in French PWR nuclear power plants

International Nuclear Information System (INIS)

Guesnier, G.

1996-01-01

The layouts of the control rooms of the French nuclear power stations have undergone great development in the period 1970-1990. The control rooms, with an architecture similar to that of the oil fired power stations, were similar to those of the 1300 MW blocks in which the human factor was emphasised. For the selection of a computerised control room for the N4 series, comprehensive functional and ergonomical validation on a full simulator was required. (author) 3 figs., 7 refs

The modernization of the process computer of the Trillo Nuclear Power Plant; Modernizacion del ordenador de proceso de la Central Nuclear de Trillo

Energy Technology Data Exchange (ETDEWEB)

Martin Aparicio, J.; Atanasio, J.

2011-07-01

The paper describes the modernization of the Process computer of the Trillo Nuclear Power Plant. The process computer functions, have been incorporated in the non Safety I and C platform selected in Trillo NPP: the Siemens SPPA-T2000 OM690 (formerly known as Teleperm XP). The upgrade of the Human Machine Interface of the control room has been included in the project. The modernization project has followed the same development process used in the upgrade of the process computer of PWR German nuclear power plants. (Author)

PWR radiation fields at combustion engineering plants through mid-1985: Final report

International Nuclear Information System (INIS)

Barshay, S.S.; Beineke, T.A.; Bradshaw, R.W.

1987-01-01

This report presents the results of the initial phase of the EPRI-PWR Standard Radiation Monitoring Program (SRMP) for PWR nuclear power plants with Nuclear Steam Supply Systems supplied by Combustion Engineering, Inc. The purposes of the SRMP are to provide reliable, consistent and systematic measurements of the rate of radiation-field buildup at operating PWR's; and to use that information to identify opportunities for radiation control and the consequent reduction of occupational radiation exposure. The report includes radiation surveys from seven participating power plants. These surveys were conducted at well-defined locations on the reactor coolant loop piping and steam generators, and/or inside the steam generator channel heads. In most cases only one survey is available from each power plant, so that conclusions about the rate of radiation-field buildup are not possible. Some observations are made about the distribution pattern of radiation levels within the steam generator channel heads and around the reactor coolant loops. The report discusses the relationship between out-of-core radiation fields (as measured by the SRMP) and: the pH of the reactor coolant, the concentration of lithium hydroxide in the reactor coolant, and the frequency of changes in reactor power level. In order to provide data for possible future correlations of these parameters with the SRMP radiation-field data, the report summarizes information available from participating plants on primary coolant pH, and on the frequency of changes in reactor power level. 12 refs., 22 figs., 7 tabs

Electrical systems design applications on Japanese PWR plants in light of the Fukushima Daiichi Accident

International Nuclear Information System (INIS)

Nomoto, Tsutomu

2015-01-01

After the Fukushima Daiichi nuclear power plant (1F-NPP) accident (i.e. Station Blackout), several design enhancements have been incorporated or are under considering to Mitsubishi PWR plants' design of not only operational plants' design but also new plants' design. Especially, there are several important enhancements in the area of the electrical system design. In this presentation, design enhancements related to following electrical systems/equipment are introduced; - Offsite Power System; - Emergency Power Source; - Safety-related Battery; - Alternative AC Power Supply Systems. In addition, relevant design requirements/conditions which are or will be considered in Mitsubishi PWR plants are introduced. (authors)

Preliminary nuclear power reactor technology qualitative assessment for Malaysia

International Nuclear Information System (INIS)

Shamsul Amri Sulaiman

2011-01-01

Since the worlds first nuclear reactor major breakthrough in December 02, 1942, the nuclear power industry has undergone tremendous development and evolution for more than half a century. After surpassing moratorium of nuclear power plant construction caused by catastrophic accidents at Three-mile island (1979) and Chernobyl (1986), today, nuclear energy is back on the policy agendas of many countries, both developed and developing, signaling nuclear revival or nuclear renaissance. Selection of suitable nuclear power technology has thus been subjected to primary attention. This short paper attempts to draw preliminary technology assessment for the first nuclear power reactor technology for Malaysia. Methodology employed is qualitative analysis collating recent finding of tnb-kepco preliminary feasibility study for nuclear power program in peninsular malaysia and other published presentations and/or papers by multiple experts. The results suggested that pressurized water reactor (PWR) is the prevailing technology in terms of numbers and plant performances, and while the commercialization of generation IV reactors is remote (e.g. Not until 2030), generation III/ III+ NPP models are commercially available on the market today. Five (5) major steps involved in reactor technology selection were introduced with a focus on introducing important aspects of selection criteria. Three (3) categories for the of reactor technology selection were used for the cursory evaluation. The outcome of these analyses shall lead to deeper and full analyses of the recommended reactor technologies for a comprehensive feasibility study in the near future. Recommendations for reactor technology option were also provided for both strategic and technical recommendations. The paper shall also implore the best way to select systematically the first civilian nuclear power reactor. (Author)

Purposes and related aspects of the nuclear power plant aging Project

International Nuclear Information System (INIS)

Garaud, J.; Reynes, L.; Montardy, A. de

1989-01-01

The purposes and some aspects of the nuclear power plant Lifetime project is presented. The work is focused on the study of the aging of the PWR type power plant components. The project consists in two phases. The first one involves: the status reports and the analysis of the lifetime data of the most sensitive units, as well as the applied actions and proposals, the influence of the exploitation modes, the safety and the cost prospectives. The second phase includes: the definition, the contracts, the complementary actions and their achievement, the reasons and the surveyance improvements, the development of the decision supporting means, the summary and the conclusions. The selected units for the project first step and the project of actions, are listed. The first conclusions of the project show that most of the 900 MW PWR reactor components will have about 40 years of service life [fr

Modelling of core protection and monitoring system for PWR nuclear power plant simulator

International Nuclear Information System (INIS)

Jung Kun Lee; Byoung Sung Han

1997-01-01

A nuclear power plant simulator was developed for Younggwang units 3 and 4 nuclear power plant (YGN Nos 3 and 4) in Korea; it has been in operation on training center since November 1996. The core protection calculator (CPC) and the core operating limit supervisory system (COLSS) for the simulator were also developed. The CPC is a digital computer-based core protection system, which performs on-line calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). It initiates reactor trip when the core conditions exceed designated DNBR or LPD limitations. The COLSS is designed to assist operators by implementing the limiting conditions for operations in the technical specifications. With these systems, it is possible to increase capacity factor and safety of nuclear power plants, because the COLSS data can show accurate operation margin to plant operators and the CPC can protect reactor core. In this study, the function of CPC/COLSS is analyzed in detail, and then simulation model for CPC/COLSS is presented based on the function. Compared with the YGN Nos 3 and 4 plant operation data and CEDIPS/COLSS FORTRAN code test results, the predictions with the model show reasonable results. (Author)

Vulnerability analysis in a pwr nuclear power plant containment building

OpenAIRE

Musolas Otaño, Antoni Maria

2013-01-01

When supervising a nuclear power plant, the containment building is crucial. Its functions are guaranteeing structural integrity and avoiding leaks in case of accident. Both events are considered of high risk. Once a given overpressure is registered inside the containment building, three possible outputs are considered: serviceability, breakdown, and collapse. The aim is the study of vulnerability. The vulnerability of the containment building under an overpressure is described by the conditi...

Reactors. Nuclear propulsion ships

International Nuclear Information System (INIS)

Fribourg, Ch.

2001-01-01

This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

Total-system expertise in economically efficient operation of nuclear power plants

International Nuclear Information System (INIS)

2000-01-01

Siemens Nuclear Power GmbH can look back on well over 40 years of experience in developing and constructing nuclear power plants. 23 Power plant units of Siemens design are in operation in five countries, and in autumn this year, another one will start commercial operation, while yet another one is under construction. In comparative international power plant surveys, the Siemens-design systems usually rank in top positions when it comes to comparing systems availability and electric power generation, and Siemens have build a reputation in manufacturing power plants up to the highest safety standards worldwide. Our experience as a manufacturer of turnkey PWR and BWR type reactors, as well as our profound knowledge of international nuclear standardisation, engineering codes and safety guides, has been used and processed to the benefit of the services offered by Siemens, resulting in well-devised service packages, and enhancements and optimisation of our machinery and equipment. Siemens has of course obtained the relevant licenses and certification for all its services and products according to DIN ISO 9001, KTA and ASME standards [de

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)

Analysis of events occurred at overseas nuclear power plants in 2004

International Nuclear Information System (INIS)

Miyazaki, Takamasa; Nishioka, Hiromasa; Sato, Masahiro; Chiba, Gorou; Takagawa, Kenichi; Shimada, Hiroki

2005-01-01

The Institute of Nuclear Safety Systems (INSS) investigates the information related to events and incidents occurred at overseas nuclear power plants, and proposes recommendations for the improvement of the safety and reliability of domestic PWR plants by evaluating them. Succeeding to the 2003 report, this report shows the summary of the evaluation activity and of the tendency analysis based on about 2800 information obtained in 2004. The tendency analysis was undertaken on about 1700 analyzed events, from the view point of mechanics, electrics and operations, about the causes, troubled equipments and so on. (author)

Microcomputer simulation of PWR power plant pressurizer

International Nuclear Information System (INIS)

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

1990-01-01

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

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

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

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.

Evaluating and improving nuclear power plant operating performance

International Nuclear Information System (INIS)

1999-07-01

This report aims to provide the basis for improvements in the understanding of nuclear power plants operation and ideas for improving future productivity. The purpose of the project was to identify good practices of operating performance at a few of the world's most productive plants. This report was prepared through a series of consultants meetings, a specialists meeting and an Advisory Group meeting with participation of experts from 23 Member States. The report is based on self-assessment of half a dozen plants that have been chosen as representatives of different reactor types in as many different countries, and the views and assessment of the participants on good practices influencing plant performance. Three main areas that influence nuclear power plant availability and reliability were identified in the discussions: (1) management practices, (2) personnel characteristics, and (3) working practices. These areas cover causes influencing plant performance under plant management control. In each area the report describes factors or good practices that positively influence plant availability. The case studies, presented in annexes, contain the plant self-assessment of areas that influence their availability and reliability. Six plants are represented in the case studies: (1) Dukovany (WWER, 1760 MW) in the Czech Republic; (2) Blayais (PWR, 3640 MW) in France; (3) Paks (WWER, 1840 MW) in Hungary; (4) Wolsong 1 (PHWR, 600 MW) in the Republic of Korea; (5) Trillo 1 (PWR, 1066 MW) in Spain; and (6) Limerick (BWR, 2220 MW) in the United States of America

Great Britain 1994: Starting a new era for nuclear power

International Nuclear Information System (INIS)

Collier, J.G.

1994-01-01

In the course of the British program of privatization of the state-owned electricity generating industry the nuclear power plants in England and Wales were incorporated into the newly founded (and state-owned) NE in 1989. In addition, the government decided in favor of a moratorium in the construction of more nuclear power plants. Only Sizewell B, Britain's first pressurized water reactor (PWR), was to be completed. At that point in time, three other PWRs were in the planning stage. The government decision included a review of the future of nuclear technology to be held in 1994. At the beginning of 1994, the nuclear future of Britain appears in a much more positive light than had been thought possible in 1989. When founding NE, the British government endowed it with a startup capital and also instituted a tax levied on fossil energy sources, the revenue from which is being made available to NE in compensation for the old liabilities NE had to assume. (orig./UA) [de

Availability Improvement of German Nuclear Power Plants

International Nuclear Information System (INIS)

Wilhelm, Oliver

2008-01-01

High availability is important for the safety and economical performance of Nuclear Power Plants (NPP). The strategy for availability improvement in a typical German PWR shall be discussed here. Key parameters for strategy development are plant design, availability of safety systems, component reliability, preventive maintenance and outage organization. Plant design, availability of safety systems and component reliability are to a greater extent given parameters that can hardly be influenced after the construction of the plant. But they set the frame for maintenance and outage organisation which have shown to have a large influence on the availability of the plant. (author)

Extension of the nuclear power plant lifetime

International Nuclear Information System (INIS)

Keramsi, Alain

2011-01-01

After a presentation of the French nuclear context (history of the reactor fleet, choice of reactor type, PWR operation principle, competitiveness, environmental performance), this Power Point presentation addresses the context and challenges of the operation lifetime (average fleet age in different countries, examples of extensions, case of the United States, what is at stake with lifetime extension, decennial visits, EDF strategy), discusses the EDF's safety objectives (definition of the three main safety functions, impact of the operation duration and of the coexistence of two generations for the safety functions), discusses how to manage the ageing phenomenon for replaceable and non-replaceable components

The Role of Nuclear Power in Pakistan

International Nuclear Information System (INIS)

Parvez, A.; Iqleem, J.

2002-01-01

Although the energy and electricity demand in Pakistan have been steadily growing, the per capita electricity consumption at around 300 kWh is still rather small when compared to most countries. The current installed capacity is around 17,700 MW with fossil fuels providing nearly two-third of this capacity, hydro a little less than one-third and nuclear around 2.5%. A major fraction of the oil used in Pakistan has to be imported while hydro remains subject to seasonal changes. The next 20 year projections point to a serious electrical energy generation shortfall even when the contribution from indigenous gas, coal, and hydro is increased optimistically. It is estimated that a deficit of some 3000-5000 MW may exist which will have to be met from an alternate energy resource like nuclear. Two small nuclear power plants (KANUPP, a 137 MWe CANDU which has been operating safely for nearly three decades, and CHASNUPP, the newly built 325 MWe PWR supplied by China) are already on-line. KANUPP has essentially been operated without any vendor support thanks to a systematic self-reliance program. The experience gained through procuring, operating and maintaining these power plants, coupled with the need to meet the projected electrical energy shortfall which cannot be met through conventional resources, makes nuclear a very viable option, and Pakistan an ideal case to study the current and future role of nuclear in a developing country with medium sized grid. This paper will describe an overview of the experience of development of nuclear power in Pakistan. Future strategies, which involve negotiating a case for nuclear with the energy policy makers, interacting with the vendor on matters of obtaining new plants, and increasing self-reliance in the area of nuclear power technology, will also be discussed. (author)

Optimum fuel use in PWR reactors

International Nuclear Information System (INIS)

Neubauer, W.

1979-07-01

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

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

International Nuclear Information System (INIS)

Drab, F.; Grof, V.

1978-01-01

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

The gamma thermometer as a measuring instrument of the nuclear power of a LWR core

International Nuclear Information System (INIS)

Hantouche, C.

1984-07-01

After a presentation of the gamma thermometer and its environment, this thesis deals with the calibration of a gamma thermometer or determination of the thermal transfer ratio of the signal. Then, the acquisition systems for signals obtained from gamma thermometers installed in PWR nuclear power plants are presented. One deals also with the nuclear transfer function establishment and deconvolution. Then, one deals with the reconstruction of the axial power of an assembly provided with a gamma thermometer. Qualitative and quantitative studies of measurements obtained from this gamma thermometer are finally presented. 64 refs [fr

Elecnuc. Nuclear power plants in the world; Elecnuc. Les centrales nucleaires dans le monde

Energy Technology Data Exchange (ETDEWEB)

NONE

2008-07-01

This small booklet summarizes in tables all data relative to the nuclear power plants worldwide. These data come from the IAEA's PRIS and AREVA-CEA's GAIA databases. The following aspects are reviewed: 2007 highlights; Main characteristics of reactor types; Map of the French nuclear power plants on 2007/01/01; Worldwide status of nuclear power plants (12/31/2007); Units distributed by countries; Nuclear power plants connected to the Grid- by reactor type groups; Nuclear power plants under construction on 2007; Evolution of nuclear power plants capacities connected to the grid; First electric generations supplied by a nuclear unit in each country; Electrical generation from nuclear power plants by country at the end 2007; Performance indicator of French PWR units; Evolution of the generation indicators worldwide by type; Nuclear operator ranking according to their installed capacity; Units connected to the grid by countries at 12/31/2007; Status of licence renewal applications in USA; Nuclear power plants under construction at 12/31/2007; Shutdown reactors; Exported nuclear capacity in net MWe; Exported and national nuclear capacity connected to the grid; Exported nuclear power plants under construction; Exported and national nuclear capacity under construction; Nuclear power plants ordered at 12/31/2007; Long term shutdown units at 12/31/2007; COL (combined licences) applications in the USA; Recycling of Plutonium in reactors and experiences; Mox licence plants projects; Appendix - historical development; Meaning of the used acronyms; Glossary.

Reviewing nuclear power station achievement

International Nuclear Information System (INIS)

Howles, L.R.

1976-01-01

For measurement of nuclear power station achievement against original purchase the usual gross output figures are of little value since the term loosely covers many different definitions. An authentically designed output figure has been established which relates to net design output plus house load at full load. Based on these figures both cumulative and moving annual load factors are measured, the latter measuring the achievement over the last year, thus showing trends with time. Calculations have been carried out for all nuclear stations in the Western World with 150 MW(e) gross design output and above. From these are shown: moving annual load factor indicating relative station achievements for all the plants; cumulative load factors from which return of investment can be calculated; average moving annual load factors for the four types of system Magnox, PWR, HWR, and BWR; and a relative comparison of achievement by country in a few cases. (U.K.)

Fourteen lessons learned from the successful nuclear power program of the Republic of Korea

International Nuclear Information System (INIS)

Choi, Sungyeol; Jun, Eunju; Hwang, IlSoon; Starz, Anne; Mazour, Tom; Chang, SoonHeung; Burkart, Alex R.

2009-01-01

This paper summarized a development history and lessons of Korean nuclear power infrastructures from the beginning of the nuclear power program in 1956 to the localization of complete scope of PWR technology in 1990. The objective of this paper is to show the guideline on the issues that the development of a national infrastructure for nuclear power using the realistic experiences in order to help the developing countries newly starting nuclear power program as a long-term energy supply option. Development strategies and lessons learned from the successful Korean experience have been presented based on milestones structure of IAEA in order to help decision makers, advisers, senior managers and national planners of nuclear power program. Lessons for national nuclear power programs include considerations before launching a program, preparation and decision making, and the construction of the first nuclear power plant. Scope of these lessons includes knowledge and human resources management, financial and industrial infrastructure development, nuclear safety, legislative and regulatory experiences, fuel cycle and waste management, international cooperation. Fourteen lessons learned either positive or not are derived from the Korean case and are suggested for incorporation in the IAEA's efforts in support of developing countries' development of nuclear infrastructure and planning.

ELECNUC. Nuclear power plants in the world - 2012 edition, Status on 2011-12-31

International Nuclear Information System (INIS)

2012-01-01

This small booklet summarizes in tables all data relative to the nuclear power plants worldwide. Data come from the IAEA's PRIS database and from specific I-tese studies. The following aspects are reviewed: 2011 highlights; Main characteristics of reactor types; Map of the French nuclear power plants on 2011/01/01; Worldwide status of nuclear power plants (12/31/2011); Units distributed by countries; Nuclear power plants connected to the Grid- by reactor type groups; Nuclear power plants under construction on 2011; Evolution of nuclear power plants capacities connected to the grid; First electric generations supplied by a nuclear unit in each country; Electrical generation from nuclear powe plants by country at the end 2011; Performance indicator of french PWR units; Evolution of the generation indicators worldwide by type; Nuclear operator ranking according to their installed capacity; Units connected to the grid by countries at 12/31/2011; Status of licence renewal applications in USA; Nuclear power plants under construction at 12/31/2011; Shutdown reactors; Exported nuclear capacity in net MWe; Exported and national nuclear capacity connected to the grid; Exported nuclear power plants under construction; Exported and national nuclear capacity under construction; Nuclear power plants ordered at 12/31/2011; Long term shutdown units at 12/31/2011; COL (Combined Licence) applications in the USA; Recycling of Plutonium in reactors and experiences; Mox licence plants projects; Appendix - historical development; Meaning of the used acronyms; Glossary

Approach and results of the PWR low power and shutdown accident frequencies program - Coarse screening analysis for Surry

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Luckas, W.; Wong, S.M.; Fitzpatrick, R.G.

1991-01-01

Traditionally, probabilistic risk analyses of severe accidents in nuclear power plants have limited themselves to consideration of the set of initiating events occurring during full power operation. However, some analyses of accident initiators during low power, shutdown, and other modes of plant operation other than full power have been performed. These studies as well as the Chernobyl accident and recent operating experience at US pressurized water reactors (PWRs) suggested that risks during low power and shutdown could be significant. As such, the analysis of the frequencies, consequences, and risks of these accidents was identified as one task in the Nuclear Regulatory Commission staff's study of the implications of the Chernobyl accident to US commercial nuclear power plants. This program is an ongoing high priority effort at Brookhaven National Laboratory (BNL). The scope includes a Level 1 probabilistic risk assessment (PRA) with internal fire and flood for Surry Unit 1 (PWR). This program is also closely coupled to a parallel project for the Grand Gulf plant (BWWR) being conducted by SNL. The program is being performed in two phases. Phase 1 represents a coarse screening analysis to identify dominant accident scenarios as well as risk dominant plant configurations and plant operating states. In Phase 2, a detailed PRA will be performed for the dominant accident scenarios/operating states identified in Phase 1. The objectives, results and insights of Phase 1 are discussed in the paper

Study of PWR reactor efficiency as a function of turbine steam extractions; Estudo da otimizacao da eficiencia de reator PWR em funcao das extracoes de vapor da turbina

Energy Technology Data Exchange (ETDEWEB)

Rocha, Janine Gandolpho da; Alvim, Antonio Carlos Marques; Martinez, Aquilino Senra [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

2002-07-01

The objective of this work is to optimize the extractions of the low-pressure turbine of a PWR nuclear reactor, in order to obtain the best thermodynamic cycle efficiency. We have analyzed typical data of a 1300 MW PWR reactor, operating at 25%, 50%, 75% and 100% capacities, respectively. The first stage of this study consists of generating a mathematical model capable of describing the reactor behavior and efficiency at any power level. The second stage of this study consists of to combine the generated mathematical model in an optimization computer program that optimize the extractions flow of the low-pressure turbine until it finds the optimal system efficiency. This work does not alter the nuclear facility project in any way. (author)

Optimization of control area ventilation systems for Japanese PWR plants

International Nuclear Information System (INIS)

Naitoh, T.; Nakahara, Y.

1987-01-01

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

Advanced I and C system of security level for nuclear power station

International Nuclear Information System (INIS)

Liu Yanyang

2001-01-01

Advanced I and C system of security level using for PWR developed by Framatome and Schneider collective, SPINLINE3, are introduced. The technology is used to outside reactor nuclear measurement system in Qinshan II period. It's succeed benefits by Framatome and Schneider's more years development experience in nuclear power station digitallization security level I and C system field, which improve security and reliability of PWR, and, easy operation and maintains. SPINLINE3 based on digitallization and modularization technical proposal, and covered entireness reactor protect system and correlative control system. The paper also introduce CLARISSE (computer aided design aid) and SCADE (embedded software aid) for developing SPINLINE3. SPINLINE3 fills correlative IS and rule, based on software and hardware unit which certificate and launch into operation. After brief review of Framatome and Schneider's experience, the paper are introducing design guideline, application technology and how to fill demand of security level I and C system

Analysis of the alternatives for the chemical treatment of the secondary circuit of PWR power plants

International Nuclear Information System (INIS)

Lopes, J.P.G.; Silva Neto, A.J. da; Braganca Junior, A.; Dominguez, D.

1990-01-01

The operational experiences within PWR power plants shows that the major problems which affect the plant availability occurs in the secondary side, mainly in the steam generators and condenser. The aim of this report is to perform an evaluation of the main chemical treatment processes, which are applied to the secondary side of PWR power plants in order to reduce the corrosion problems to which are exposed the system equipment, minimizing in this way the shut down and maintenance cost for repairs and replacement of damaged components. (author)

Radiation risk analysis of tritium in PWR plants

International Nuclear Information System (INIS)

Yang Maochun; Wang Shimin

1999-03-01

Tritium is a common radionuclide in PWR nuclear power plant. In the normal operation conditions, its radiation risk to plant workers is the internal radiation exposure when tritium existing in air as HTO (hydrogen tritium oxide) is breathed in. As the HTO has the same physical and chemical characteristics as water, the main way that HTO entering the air is by evaporation. There are few opening systems in Nuclear Power Plant, the radiation risk of tritium mainly exists near the area of spent fuel pit and reactor pit. The highest possible radiation risk it may cause--the maximum concentration in air is the level when equilibrium is established between water and air phases for tritium. The author analyzed the relationship among the concentration of HTO in water, in air and the water temperature when equilibrium is established, the equilibrated HTO concentration in air increases with HTO concentration in water and water temperature. The analysis revealed that at 30 degree C, the equilibrated HTO concentration in air might reach 1 DAC (derived air concentration) when the HTO concentration in water is 28 GBq/m 3 . Owing to the operation of plant ventilation systems and the existence of moisture in the input air of the ventilation, the practical tritium concentration in air is much lower than its equilibrated levels, the radiation risk of tritium in PWR plant is quite limited. In 1997, Daya Bay Nuclear Power Plant's practical monitoring result of the HTO concentration in the air of the nuclear island and the urine of workers supported this conclusion. Based on this analysis, some suggestions to the reduction of tritium radiation risk were made

PIE of test assembly of Qinshan nuclear power plant

International Nuclear Information System (INIS)

Ran, M.; Yan, J.; Wang, S.

2000-01-01

The small dimensional test fuel assembly (3x3-2) for the Qinshan Nuclear Power Plant was irradiated up to 25.7 Gwd/tU in the in-pile loop (15.5 Mpa,320 C) in Heavy Water Research Reactor (HWRR), CIAE, at simulative condition to Qinshan PWR normal and short time overpower operation for verifying the design, technology, and material properties of the fuel assembly. Comprehensive post-irradiation examination (PIE) including dimension measurement, gamma scanning, eddy current test, X ray, radiography, measurement of fission gas release, and quantitative metallography etc. were performed. PIE results show that the diameter of the fuel rods changed, ridges appeared on the cladding, pellets swelled, and the rate of fission gas release was higher than what we expected. The results would be an important basis for further improvement of design, technology and material properties for Qinshan PWR assembly. (author)

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

Directory of Open Access Journals (Sweden)

Tagor Malem Sembiring

2017-01-01

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

Experimental research on passive residual heat remove system for advanced PWR

International Nuclear Information System (INIS)

Huang Yanping; Zhuo Wenbin; Yang Zumao; Xiao Zejun; Chen Bingde

2003-01-01

The experimental and qualified results of MISAP in the research of passive residual heat remove system of advanced PWR performed in the Bubble physics and natural circulation laboratory in Nuclear Power Institute of China in the past ten years is overviewed. Further researches for engineering research and design are also suggested

Survey of the power ramp performance testing of KWU'S PWR UO 2, fuel

Science.gov (United States)

Ga¨rtner, M.; Fischer, G.

1987-06-01

To determine the power ramp performance of KWU's PWR UO 2 fuel, 134 fuel rodlets with burnups of up to 46 GWd/ t (U) and several fuel assemblies with 19 to 30 GWd/t (U) burnup were ramped in power in the research reactors HFR Petten/The Netherlands and R2 Studsvik/Sweden and in the power plants KWO and KWB-A/Germany, respectively. The power ramp tests demonstrate decreasing resistance of the PWR fuel rods to PCI (pellet-to-clad interaction) up to fuel burnups of 35 GWd/t (U) and a reversal effect at higher burnups. The fuel rods can be operated free of defects at fast power transients to linear heat generation rates of up to 400 W/cm, at least.Power levels of up to 490 W/cm can be reached without defects by reducing the ramp rate. After reshuffling according to an out-in scheme, 1-cycle fuel assemblies may return to rod powers of up to 480 W/cm with a power increase rate of up to 10 W/(cm min) without fuel rod damage. Set points basing on these test results and incorporated into the power distribution control and power density limitation system of KWU's advanced power plants guarantee safe plant operation under normal and load follow operating conditions.

Operating experience with nuclear power stations in Member States in 2005

International Nuclear Information System (INIS)

2006-10-01

difficulty in reporting unavailability in energy due to external causes with relation to energy losses due to load following operation and grid limitation. It should be noted that, for load, operation and unavailability factors, there might be differences between the data of this report and those published elsewhere. To avoid confusion, reference should be made to the definitions given in Section 3. In Section 4 this report presents figures illustrating worldwide performance indicators up to 2005. According to the information available to the Agency at the end of 2005, there were 443 nuclear power reactors operating in the world, with a total net capacity of 369.6 GW (e) . Four new reactor units were connected to the grid in 2005 (two in Japan and one each in India and the South Korea), and one laid-up plant was reconnected in Canada. There were two NPP retirements in 2005 - one 340 MW (e) reactor in Germany and the one 600 MW (e) reactor in Sweden. This compares to five retirements in 2004. At the end of 2005 here were 27 nuclear power plants under construction in the world with a total net capacity 21.8 GW (e) . Construction began on three NPPs in 2005, Finland's 1600 MW (e) PWR (EPR) reactor, China's 1000 MW (e) PWR and Pakistan's 300 MW (e) PWR. In addition, active construction has resumed on two NPPs in the Bulgaria. The scope of publication has been enhanced by information related to non-electrical application of nuclear power reactors. The Chapter 7 consists of production data related to district heating, industrial process heat delivery and to water desalination process. In 2005 the nuclear energy was utilized for nonelectrical application in 9 Member States involving energy from 66 nuclear reactors. This publication includes information received by the Agency up to 1 September 2005. Up to this date data from all operating units had been reported. Any data modification received after that date, although not included in this publication, is available in the PRIS

Corrosion products in the coolant circuits of PWR nuclear power plants

International Nuclear Information System (INIS)

Darras, R.

1984-01-01

The characteristics of corrosion products formed in the primary and secondary circuits of pressurized light water nuclear power plants are first briefly recalled. The problem set by the pollution of coolants and metallic surfaces is then examined. Finally, the measures of precaution to take and the possible solutions to minimize the disturbing effects of this pollution by corrosion products are presented [fr

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)

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)

Study of PWR reactor efficiency as a function of turbine steam extractions

International Nuclear Information System (INIS)

Rocha, Janine Gandolpho da; Alvim, Antonio Carlos Marques; Martinez, Aquilino Senra

2002-01-01

The objective of this work is to optimize the extractions of the low-pressure turbine of a PWR nuclear reactor, in order to obtain the best thermodynamic cycle efficiency. We have analyzed typical data of a 1300 MW PWR reactor, operating at 25%, 50%, 75% and 100% capacities, respectively. The first stage of this study consists of generating a mathematical model capable of describing the reactor behavior and efficiency at any power level. The second stage of this study consists of to combine the generated mathematical model in an optimization computer program that optimize the extractions flow of the low-pressure turbine until it finds the optimal system efficiency. This work does not alter the nuclear facility project in any way. (author)

The empirical intensity of PWR primary coolant pumps failure and repair

International Nuclear Information System (INIS)

Milivojevicj, S.; Riznicj, J.

1988-01-01

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

Coolant flow monitoring in a PWR core using noise analysis

International Nuclear Information System (INIS)

Kostic, Lj.

1992-01-01

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

Dynamic analysis of Boushehr Nuclear Power Plant in connected to grid system

International Nuclear Information System (INIS)

Karimi Fard, A.

1999-01-01

Models of generating with the pressurized water reactor (PWR) have been developed for simulating. the plant dynamics under system disturbances. These models include power plant, energy sources, turbine, transmission system and control system such as Avr and govern and other local control devices. Simulink toolbox of Matlab software is used for simulations. The study is mainly based on the Bushehr Nuclear Power Plants (BNPP) parameters. Assuming that BNPP is connected to infinite bus with double tie line. Four cases are studied to examine the internal dynamic behavior of BNPP. First and second cases are used to load following studies in nuclear power plant. Another cases are used to study the dynamic behavior after short circuit fault and line outages in transmission systems. The results discussed in the thesis

Design and Development of Virtual Reactivity System for PWR

International Nuclear Information System (INIS)

Anwar, M. I.

2012-01-01

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

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)

Evaluating and improving nuclear power plant operating performance

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-07-01

This report aims to provide the basis for improvements in the understanding of nuclear power plants operation and ideas for improving future productivity. The purpose of the project was to identify good practices of operating performance at a few of the world`s most productive plants. This report was prepared through a series of consultants meetings, a specialists meeting and an Advisory Group meeting with participation of experts from 23 Member States. The report is based on self-assessment of half a dozen plants that have been chosen as representatives of different reactor types in as many different countries, and the views and assessment of the participants on good practices influencing plant performance. Three main areas that influence nuclear power plant availability and reliability were identified in the discussions: (1) management practices, (2) personnel characteristics, and (3) working practices. These areas cover causes influencing plant performance under plant management control. In each area the report describes factors or good practices that positively influence plant availability. The case studies, presented in annexes, contain the plant self-assessment of areas that influence their availability and reliability. Six plants are represented in the case studies: (1) Dukovany (WWER, 1760 MW) in the Czech Republic; (2) Blayais (PWR, 3640 MW) in France; (3) Paks (WWER, 1840 MW) in Hungary; (4) Wolsong 1 (PHWR, 600 MW) in the Republic of Korea; (5) Trillo 1 (PWR, 1066 MW) in Spain; and (6) Limerick (BWR, 2220 MW) in the United States of America Figs, tabs

NCS--a software for visual modeling and simulation of PWR nuclear power plant control system

International Nuclear Information System (INIS)

Cui Zhenhua

1998-12-01

The modeling and simulation of nuclear power plant control system has been investigated. Some mathematical models for rapid and accurate simulation are derived, including core models, pressurizer model, steam generator model, etc. Several numerical methods such as Runge-Kutta Method and Treanor Method are adopted to solve the above system models. In order to model the control system conveniently, a block diagram-oriented visual modeling platform is designed. And the Discrete Similarity Method is used to calculate the control system models. A corresponding simulating software, NCS, is developed for researching on the control systems of commercial nuclear power plant. And some satisfactory results are obtained. The research works will be of referential and applying value to design and analysis of nuclear power plant control system

Study on in-core fuel management for CNP1500 nuclear power plant

International Nuclear Information System (INIS)

Li Dongsheng

2005-10-01

CNP1500 is a four-loop PWR nuclear power plant with light water as moderator and coolant. The reactor core is composed of 205 AFA-3GXL fuel assemblies. The active core height at cold is 426.4 cm and equivalent diameter is 347.0 cm. The reactor thermal output is 4250 MW, and average linear power density is 179.5 W/cm. The cycle length of equilibrium cycle core is 470 equivalent full power days. For all cycles, the moderator temperature coefficients at all conditions are negative values, the nuclear enthalpy rise factors F ΔH at hot full power, all control rods out and equilibrium xenon are less than the limit value, the maximum discharge assembly burnup is less 55000 MW·d/tU, and the shutdown margin values at the end of life meet design criteria. The low-leakage core loading reduces radiation damage on pressure vessel and is beneficial to prolong use lifetime of it. The in-core fuel management design scheme and main calculation results for CNP1500 nuclear power plant are presented. (author)

An intelligent pedagogic tool for teaching the operators of PWR type reactors

International Nuclear Information System (INIS)

Cordier, B.; Guillermard, M.

1990-01-01

A tool was developed for assisting the instruction of the operators of a PWR type nuclear power plant. For achieving the objectives, an expert system and a simulator were combined. The main objective of the system is to improve the work of the operators in performing remedial actions in case of accident. The simulator applies two IBM PC AT3 and a MC 680 20 microprocessor. The use and the validation of the expert system are presented. The perspectives for the system, implanted on the Tricastin nuclear power plant, are analyzed [fr

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

Institute of Scientific and Technical Information of China (English)

2002-01-01

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

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

Simulation of a PWR power plant for process control and diagnosis

International Nuclear Information System (INIS)

Ravnsbjerg Nielsen, F.

1991-12-01

A computer model of a simplified pressurized nuclear power plant is developed with aim at studies concerning process control, diagnosis and decision making. The model includes the traditional PWR plant components, primary circuit with reactor, pressurizer and steam generator, steam circuit with steam line, turbine and condenser, interconnected with pumps, valves and controllers. The model can be used for calculation of transients for both normal operation and incidents such as turbine trip, loss of feedwater, run down of pumps or various valve failures. The computer model is not directed to any specific existing plant. For convenience and alleviation in implementation the physical description of many components are simplified to an extent where the qualitative behavior of the system is not violated. For computer memory economy a variety of thermodynamical functions for water and steam have been approximated with analytical expressions based on table values. The model is implemented in the C language and has been run on both the IBM PC and the SUN workstation. (au) 8 tabs., 25 ills., 10 refs

Nuclear and radioactivity

International Nuclear Information System (INIS)

2000-01-01

Among the industrial risks of nuclear facilities, the nuclear risk is often associated to the Chernobyl accident. This paper presents the nuclear major risk in a french PWR type power plant, with consequences on the personnel, the surrounding population and the environment. (A.L.B.)

A comparison of nuclear power systems for Brazil using plutonium and binary cycles

International Nuclear Information System (INIS)

Ishiguro, Y.; Fernandes, J.E.

1985-01-01

Nuclear power systems based on plutonium cycle and binary cycle are compared taking into account natural uranium demand and reactor combination. The systems start with PWR type reactors (U5/U8) and change to systems composed exclusively of FBR type reactors or PWR-FBR symbiotic systems. Four loading modes are considered for the PWR and two for the FBR. The FBR is either a LMFBR loaded with PU/U or a LMFBR loaded the binary way. A linear and a non-linear capacity growth and two different criteria for the FBR introduction are considered. The results show that a 100 GWe permanent system can be established in 50 years in all cases, based on 300000 t of natural uranium and in case of delay in the FBR introduction and if a thermal-fast symbiotic system is chosen, a binary cycle could be more advantageous than a plutonium cycle. (F.E.) [pt

The performance trends of nuclear power plants worldwide

Energy Technology Data Exchange (ETDEWEB)

Glorian, D. [Electricite de France (EDF), 93 - Saint-Denis (France)

2001-07-01

Looking back to the worldwide operating experience feedback, which performance trends and conclusions could be drawn up? What is the specific situation of the French nuclear units, in comparison with the average worldwide performance? The performance of a unit or group of facilities is measured not only in technical terms (safety, availability, load control capability), but also from an economic and financial standpoint (operating and maintenance costs, fuel costs, etc). Performance in terms of radiological protection and on-the-job safety, as well as environmental protection, is also monitored in order to give the broadest possible overview of nuclear power plant performance. The main technical results are presented on the basis of selected performance indicators. The results obtained by French units are benchmarked against those of other PWR facilities in operation around the world, in accordance with comparisons made by the World Association of Nuclear Operators (WANO). (author)

The performance trends of nuclear power plants worldwide

International Nuclear Information System (INIS)

Glorian, D.

2001-01-01

Looking back to the worldwide operating experience feedback, which performance trends and conclusions could be drawn up? What is the specific situation of the French nuclear units, in comparison with the average worldwide performance? The performance of a unit or group of facilities is measured not only in technical terms (safety, availability, load control capability), but also from an economic and financial standpoint (operating and maintenance costs, fuel costs, etc). Performance in terms of radiological protection and on-the-job safety, as well as environmental protection, is also monitored in order to give the broadest possible overview of nuclear power plant performance. The main technical results are presented on the basis of selected performance indicators. The results obtained by French units are benchmarked against those of other PWR facilities in operation around the world, in accordance with comparisons made by the World Association of Nuclear Operators (WANO). (author)

Application of the minicomputer at Genkai Nuclear Power Station

Energy Technology Data Exchange (ETDEWEB)

Kitamura, H [Kyushu Electric Power Co., Inc., Fukuoka (Japan)

1977-03-01

Genkai Nuclear Power Station introduced a minicomputer system for the data control purpose in addition to a process control computer introduced in the same manner as other PWR nuclear power stations. This system employs two computers; the one for on-line data aquisition, and another for data processing. The control system introduced includes four systems amoung various data control businesses in the nuclear power station. The language used is mainly an assembler language. The first is the meteorological control system which collects, edits and transmits the weather data sent from the observation instruments around the power station. The second is the personal radiation exposure control system which is designed to realize the labor-saving in book-keeping, the speed-up and the improvement of accuracy in the preparation of the reports to the authorities and the head office and the data for exposure control, and the unification of data processing. The third is the waste control system composed of three subsystems of gas, liquid and solid waste control. The fourth is the maintenance and repair control system which gives inputs to the computer according to the classification written in the slips for maintenance and repair, and prepares a number of statistical tables for maintenance control.

Study of aging effects in PWR power plants components - 15043

International Nuclear Information System (INIS)

Silva Borges, D. da; Lava, D.D.; Guimaraes, A.C.F.; Moreira, M. de L.

2015-01-01

In this paper we present a simulation about the aging process of the containment spray injection system (CSIS) of a pressurized water reactor (PWR) using the fault tree method (FT). The FT has the capacity to present the logic of events that leads to system unavailability, to capture frequency estimation of events, to model and calculate hazardous events frequency (before they happen) and help developing protective layers. The Monte Carlo method and Fussell-Vesely importance measure are used in this paper to determine the system unavailability probability and the most sensitive events to the aging process. The injection system fault tree consists of a main tree and 10 sub-trees. The main tree is composed of 35 basic events, 5 gates and 1 top event. The paper details the methodology. It can be seen that the increase of the failure rate of components due to the aging process, generates the increase in the general unavailability of the system that contains these components. The extension of the operating life of nuclear power plant must be accompanied by a special attention to the aging process of its components

The Spanish Nuclear Safety Council and nuclear power stations in operation in Spain

International Nuclear Information System (INIS)

Perello, M.

1984-01-01

On 20 April 1980 the Spanish Congress of Deputies passed an Act setting up the Nuclear Safety Council (CSN) as the sole organization responsible for nuclear safety and radiation protection. In this paper it is stated that that date marked the beginning of a new nuclear safety policy in Spain. As one of its objectives, this policy is aimed at the monitoring and testing of operating nuclear installations. A detailed description is given of the Operating Nuclear Installation Service (SINE), including its basic structure, its functions and the technical and manpower resources available to it. The maintenance of close relations with other organs of the CSN is considered of paramount importance in order for the tasks allotted to SINE to be fulfilled. International co-operation and outside contracting greatly assist importing countries which have limited manpower resources. A description is then given of the present state of the nuclear power stations in operation in Spain together with an account of the most important initiatives which have been taken so far. The year 1968 saw the beginning of commercial operation of the Jose Cabrera nuclear power station, which has the only single-loop PWR reactor in the world. At present, it is being subjected to the Systematic Evaluation Programme (SEP). The Santa Maria de Garona nuclear power station has been operating for over twelve years and is also being subjected to the SEP although design modifications derived from operating experience have already been introduced. The Vandellos I station was the last of the first generation and has also benefited from the operating experience of similar French plants. Unit 1 of the Almaraz power station opens the door to the second generation and the generic problem which has occurred with the steam generators is in process of being solved. Lastly, some general conclusions are presented about the organization of and experience acquired with operating nuclear power stations. (author)

The nuclear industry in France

International Nuclear Information System (INIS)

Degot, D.

1981-02-01

The French nuclear industry is organized around the following main participants: - The E.D.F., owners, industrial architects and operators of the power stations, - The C.E.A. for research and development, with its subsidiary the COGEMA, who deal with all problems involving the fuel cycle, - The Industry with FRAMATOME in charge of the manufacture of nuclear boilers, and ALSTHOM-ATLANTIQUE in charge of turbo-generator units. This paper deals with the activities covered by FRAMATOME and its industrial environment. The standardization of PWR power stations built by French industry and the possibilities of exporting PWR power stations are given a brief mention [fr

Electricity Generation Through the Koeberg Nuclear Power Station of Eskom in South Africa

International Nuclear Information System (INIS)

Dladla, G.; Joubert, J.

2015-01-01

The poster provides information on the process of nuclear energy generation in a nuclear power plant in order to produce electricity. Nuclear energy currently provides approximately 11% of the world’s electricity needs, with Koeberg Nuclear Power Station situated in the Western Cape providing 4.4% of South Africa’s electricity needs. As Africa’s first nuclear power station, Koeberg has an installed capacity of 1910 MW of power. Koeberg’ s total net output is 1860 MW. While there are significant differences, there are many similarities between nuclear power plants and other electrical generating facilities. Uranium is used for fuel in nuclear power plants to make electricity. With the exception of solar, wind, and hydroelectric plants, all others including nuclear plants convert water to steam that spins the propeller-like blades of a turbine that spins the shaft of a generator. Inside the generator coils of wire and magnetic fields interact to create electricity. The energy needed to boil water into steam is produced in one of two ways: by burning coal, oil, or gas (fossil fuels) in a furnace or by splitting certain atoms of uranium in a nuclear energy plant. The uranium fuel generates heat through a controlled fission process fission, which is described in this poster presentation. The Koeberg Nuclear Power Station is a Pressurised water reactor (PWR). The operating method and the components of the Koeberg Power Station are also described. The nuclear waste generated at a nuclear power station is described under three headings— low-level waste, intermediate-level waste and used or spent fuel, which can be solid, liquid or gaseous. (author)

Water Chemistry Control Technology to Improve the Performance of Nuclear Power Plants for Extended Fuel Cycles

International Nuclear Information System (INIS)

Maeng, W. Y.; Na, J. W.; Lee, E. H.

2010-07-01

Ο To Develop the technology to manage the problems of AOA and radiation, corrosion as long term PWR operation. Ο To Establish the advanced water chemical operating systems. - Development of the proper water chemistry guidelines for long term PWR operation. AOA(Axial Offest Anomaly) has been reported in many PWR plants in the world, including Korea, especially in the plants of higher burn-up and longer cycle operation or power up-rate. A test loop has been designed and made by KAERI, in order to investigate and mitigate AOA problems in Korea. This project included the study of hydrodynamic simulation and the modeling about AOA. The analysis of radioactive crud was performed to investigate of NPPs primary water chemical effect on AOA and to reduce the radioactive dose rate. The high temperature measurement system was developed to on-line monitor of water chemistry in nuclear power plants. The effects of various environmental factors such as temperature, pressure, and flow rate on YSZ-based pH electrode were evaluated for ensuring the accuracy of high-temperature pH measurement. The inhibition technology for fouling and SCC of SG tube was evaluated to establish the water chemistry technology of corrosion control of nuclear system. The high temperature and high pressure crevice chemistry analysis test loop was manufactured to develop the water chemistry technology of crevice chemistry control

Thermo hydrodynamical analyses of steam generator of nuclear power plant

International Nuclear Information System (INIS)

Petelin, S.; Gregoric, M.

1984-01-01

SMUP computer code for stationary model of a U-tube steam generator of a PWR nuclear power plant was developed. feed water flow can enter through main and auxiliary path. The computer code is based on the one dimensional mathematical model. Among the results that give an insight into physical processes along the tubes of steam generator are distribution of temperatures, water qualities, heat transfer rates. Parametric analysis permits conclusion on advantage of each design solution regarding heat transfer effects and safety of steam generator. (author)

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)

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.

Structural experiences at the Kewaunee Nuclear Power Plant

International Nuclear Information System (INIS)

Setlur, A.V.

1983-01-01

This paper discusses the original structural and geotechnical design and subsequent structural experience at the Kewaunee Nuclear Power Plant. The original design of the 535 MWe Westinghouse two loop PWR nuclear plant operated by Wisconsin Public Service Corporation, was started in 1967 and was completed in 1974 when the unit was put into commercial operation. Since 1974 a number of changes in the regulations and additional requirements have been imposed on operating reactors. The paper traces the influence of the original plant criteria on the backfit evaluations and the minimal physical changes required in the plant's structures and components to comply with the new requirements. In addition, the unique design features and construction challenges of the original design are discussed. Kewaunee Nuclear Power Plant has had one of the best operating performance records in the world. Also, the exposure to radiation for plant personnel and radioactive waste generation has been significantly lower than the average. This has been achieved by a conscientious team effort of all parties involved. Some of the more significant structural design features contributing to the excellent performance is detailed in this paper. (orig.)

Mathematical modelling of plant transients in the PWR for simulator purposes

International Nuclear Information System (INIS)

Hartel, K.

1984-01-01

This chapter presents the results of the testing of anticipated and abnormal plant transients in pressurized water reactors (PWRs) of the type WWER 440 by means of the numerical simulation of 32 different, stationary and nonstationary, operational regimes. Topics considered include the formation of the PWR mathematical model, the physical approximation of the reactor core, the structure of the reactor core model, a mathematical approximation of the reactor model, the selection of numerical methods, and a computerized simulation system. The necessity of a PWR simulator in Czechoslovakia is justified by the present status and the outlook for the further development of the Czechoslovak nuclear power complex

Construction of new nuclear power plant in Bulgaria - incentives and problems

International Nuclear Information System (INIS)

Popov, M.; Bataklieva, L.; Hinovski, I.

2002-01-01

Advantages and disadvantages of constriction of new nuclear power plant in Bulgaria are discussed. Several options for further development are considered: Completion of the Belene NPP; Evolutionary design installations PWR technology, 600 MW (AP-600 Westinghouse design; B-407 Russian design); Standard design of Framatome, 900 MW; Standard design - heavy water technology CANDU-6 (700 MW) and other.The results of the preliminary economic analyses show that the first two options are approximately equal, but the all options are acceptable from investment point of view, considering an average macro economic environment. The need of assessment of all aspects (technical, economical, financial, social, environmental, national infrastructure) of nuclear sector development is outlined

SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

Energy Technology Data Exchange (ETDEWEB)

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

2010-03-01

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

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)

RCC-M - Design and Conception Rules for Mechanical Components of PWR Nuclear Islands

International Nuclear Information System (INIS)

2007-01-01

The design and construction rules applicable to mechanical components of PWR Nuclear Islands (RCC-M) are a part of the collection of design and construction rules for nuclear power plants. It covers the rules applicable to the design and manufacture of pressure boundaries of mechanical equipment of pressurized water reactors (PWR). The pressure components subject to the RCC-M are specified in A 4000. They include the reactor fluid systems (primary, secondary and auxiliary systems) and other components which are not subject to pressure: vessel internals, supports for pressure components subject to the RCC-M, nuclear island storage tanks. When a pressure equipment is subject to the RCC-M, all its elements subject to pressure are also, in accordance with the provisions of A 4000, and these elements are the same class as the component. In this case all the provisions of the RCC-M are applicable: design, procurement, manufacture, inspection and pressure testing. Elements which are not subject to pressure and which are subject to the RCC-M may be covered within the Code by limited specific provisions (procurement of materials for example). The other rules applicable to this equipment must be in contractual form. The assemblies comprising pressure equipment assembled by a manufacturer to constitute an integrated and functional whole, shall be subject to the rules indicated in this Code. Main objectives of Code Requirements are to ensure the integrity and mechanical stability over the equipment design life. Function ability and operability of equipment are not directly addressed in the Code. The RCC-M contributes to ensuring compliance with regulatory requirements. These requirements depend on the applicable regulatory context. The RCC-M is representative of the state of the art as concerns the design and manufacture of PWR components, ensuring an overall safety level tested through experience. The RCC-M consists of five sections, which provide rules for the design and

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)

A study for the establishment of regulatory requirement and evaluation guide for station blackout in nuclear power plants

International Nuclear Information System (INIS)

Lim, J. H.; Koo, C. S.; Joo, W. P.; Oh, S. H.; Shin, W. K.

1999-01-01

The consequence of SBO event could be a severe accident unless AC power was restored within a proper time, because many safety systems depend upon AC power. Based on the severity, the SBO has been extensively studied since it was identified as Unresolved Safety Issue at USNRC. The resolution of those studies is a rule-making such as 10 CFR 50.63 and Regulatory Guide 1.155. But there is no regulatory requirements of SBO for an operating domestic nuclear power plant up to the present time. This tudy has established SBO rule(regulatory requirements and evaluation guides) for an operating PWR type of the operating nuclear power plants in Korea

MELCOR/VISOR PWR desktop simulator

International Nuclear Information System (INIS)

With, Anka de; Wakker, Pieter

2010-01-01

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

A reliability centered maintenance model applied to the auxiliary feedwater system of a nuclear power plant; Um modelo de manutencao centrada em confiabilidade aplicada ao sistema de agua de alimentacaco auxiliar de uma usina nuclear

Energy Technology Data Exchange (ETDEWEB)

Araujo, Jefferson Borges

1998-01-15

The main objective of maintenance in a nuclear power plant is to assure that structures, systems and components will perform their design functions with reliability and availability in order to obtain a safety and economic electric power generation. Reliability Centered Maintenance (RCM) is a method of systematic review to develop or optimize Preventive Maintenance Programs. This study presents the objectives, concepts, organization and methods used in the development of RCM application to nuclear power plants. Some examples of this application are included, considering the Auxiliary Feedwater System of a generic two loops PWR nuclear power plant of Westinghouse design. (author)

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

Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

1983-08-01

This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

Recent developments of and future prospects for nuclear power plant instrumentation and control in Belgium

International Nuclear Information System (INIS)

Naisse, J.C.

1990-01-01

The production of electricity in Belgium is covered for almost 70% by seven nuclear power plants all of pressurized water reactor type. Four major topics are representative of the evolution of nuclear power production in Belgium: The suspentions of the studies for the new Belgian PWR unit (DOEL 5), the commissioning of the simulators in DOEL and TIHANGE, the revamping of the oldest nuclear units, the studies of new storage installations for radioactive waste. The delay imposed on the construction of DOEL 5 will allow to go further into the examination of the implications arising from the new technologies in I and C. The use modern I and C systems for modifications in existing plants will allow a better estimation of their advantages and disadvantages. (author)

Investigation on the use digital controls instead of PID analog controls in the level control of steam generators of nuclear power PWR; Investigacao sobre o uso de controladores digitais em substituicao aos controladores analogicos PID para o controle de nivel de geradores de vapor de centrais nucleares PWR

Energy Technology Data Exchange (ETDEWEB)

Alvarenga, Marco Antonio Bayout

2012-07-01

The aim of this study is to identify current alternatives for the implementation of digital controllers in the level control of steam generators of nuclear power PWR (Pressurized Water Reaetor). It is intended to identify the types of digital controls that are available from the theoretical and conceptual viewpoints for this purpose. We investigate the advantages and disadvantages of each controller model. From this assessment are pointed the most suitable models in hierarchical scale. This evaluation also serves to suggest possible types of control installation as a whole, where the level control of the steam generators becomes just one of many controls that are part of the plant. In this case, the use of digital controls allows the non-linear and multivariable treatment which is characteristic of complex systems, such as the nuclear power generation. The treatment of nonlinearities and multivariable aspects allows a more detailed study of the stability of these plants when they are subject to transients or several accidents, such as the case of losing external power of transients. In the specific case of steam generators, the instabilities result from the emergence of the shrink and swell phenomenas, depending on the load variations of thermonuclear plant. The application of several types and digital controllers, considering these inherent characteristics of the level control of steam generators, allows to infer which types of controllers are more appropriate to treat instabilities of this type and to make conjectures in its use for the cases of more complex instabilities, considering the integration of all nucleus-plant controls.

Establishment of the technical basis for applying viscous dampers to nuclear power plants

International Nuclear Information System (INIS)

Narahara, Yukiko; Higuchi, Tomokazu; Katayama, Hiroshi; Ito, Ryo; Hattori, Kiyoshi; Nakajima, Jun

2017-01-01

For the purpose of introducing viscous dampers to nuclear power plants, the damping characteristic of the viscous damper under specific conditions in nuclear power plants was examined. In particular the seismic response analysis method, the design evaluation method, and the maintenance and management guideline were studied. In the viscous dampers characteristic test, the damping characteristics under earthquake motion and the environmental condition of nuclear power plants have been examined. From the test results, if the parameters such as vibration amplitude, vibration frequency, repeated load, temperature, and radiation are considered, there is a possibility of viscous damper application to components in nuclear power plants. In order to evaluate the applicability of the complex modal analysis method using response spectrum, comparison with the time history response analysis result was performed using a PWR steam generator analysis model. The evaluation result from the complex modal analysis method was in good agreement with the time history response analysis result, and the availability of this method was confirmed. From the test results, considerations in design and maintenance in the case of applying viscous dampers to nuclear power plants were selected. The bases of the design evaluation method and the maintenance and management guideline were developed. (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

Modeling and control of a nuclear power plant using AI techniques

International Nuclear Information System (INIS)

Oliveira, Mauro Vitor de; Almeida, Jose Carlos Soares de

2013-01-01

In pressurized water reactor (PWR) nuclear power plants (NPPs) pressure control in the primary loops is fundamental for keeping the reactor in a safety condition and improve the generation process efficiency. The main component responsible for this task is the pressurizer. The pressurizer pressure control system (PPCS) utilizes heaters and spray valves to maintain the pressure within an operating band during steady state conditions, and limits the pressure changes during transient conditions. Relief and safety valves provide overpressure protection for the reactor coolant system (RCS) to ensure system integrity. Various protective reactor trips are generated if the system parameters exceed safe bounds. Historically, a proportional-integral-derivative (PID) controller is used in PWRs to keep the pressure in the set point, during those operation conditions. The purpose of this study is to develop fuzzy controllers for the PWR pressurizer modeled by an artificial neural network (ANN) and compare their performance with conventional ones. Data from a 2785 MWth Westinghouse 3-loop PWR simulator was used to test both the conventional and the fuzzy controllers. The simulation results show that the fuzzy controllers have better performance compared with conventional ones. (author)

Modeling and control of a nuclear power plant using AI techniques