BWR type nuclear reactors

International Nuclear Information System (INIS)

Yamamoto, Toru.

1987-01-01

Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

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)

BWR stability analysis at Brookhaven National Laboratory

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

1991-01-01

Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

GPE-BWR and the containment venting and filtering issue

International Nuclear Information System (INIS)

Palomo, J.; Santiago, J. de

1988-01-01

The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

Laguna Verde nuclear power plant: an experience to consider in advanced BWR design

International Nuclear Information System (INIS)

Fuentes Marquez, L.

2001-01-01

Laguna Verde is a BWR 5 containment Mark II. Designed by GE, two external re-circulation loops, each of them having two speed re-circulation pump and a flow control valve to define the drive flow and consequently the total core flow an power control by total core flow. Laguna Verde Design and operational experience has shown some insights to be considering in design for advanced BRW reactors in order to improve the potential of nuclear power plants. NSSS and Balance of plant design, codes used to perform nuclear core design, margins derived from engineering judgment, at the time Laguna Verde designed and constructed had conducted to have a plant with an operational license, generating with a very good performance and availability. Nevertheless, some design characteristics and operational experience have shown that potential improvements or areas of opportunity shall be focused in the advanced BWR design. Computer codes used to design the nuclear core have been evolved relatively fast. The computers are faster and powerful than those used during the design process, also instrumentation and control are becoming part of this amazing technical evolution in the industry. The Laguna Verde experience is the subject to share in this paper. (author)

Experience and development of on-line BWR surveillance system at Onagawa nuclear power station unit-1

International Nuclear Information System (INIS)

Kishi, A.; Chiba, K.; Kato, K.; Ebata, S.; Ando, Y.; Sakamoto, H.

1986-01-01

ONAGAWA nuclear power station Unit-1 (Tohoku Electric Power Co.) is a BWR-4 nuclear power station of 524 MW electric power which started commercial operation in June 1984. To attain high reliability and applicability for ONAGAWA-1, Tohoku Electric Power Co. and Toshiba started a Research and Development project on plant surveillance and diagnosis from April 1982. Main purposes of this project are to: (1) Develop an on-line surveillance system and acquire its operating experience at a commercial BWR, (2) Assist in plant operation and maintenance by data acquisition and analysis, (3) Develop a new technique for plant surveillance and diagnosis. An outline of the project, operating experience gained from the on-line surveillance system and an introduction to new diagnosis techniques are reported in this paper. (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.)

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.)

An analysis of instabilities of nuclear-coupled density-wave in BWR using modern frequency-domain control theory

International Nuclear Information System (INIS)

Zhao Yangping; Gao Huahun; Fu Longzhou

1991-01-01

A state-of-the-art multi-variable frequency-domain model has been developed for analysis of instabilities of nuclear-coupled density-wave in BWR core. The characteristic locus method is used for analysing the stability of BWR. A computer code-NUCTHIA has been derived. The model has been tested against the existing experimental data and compared with results of past single-variable analyses. By using the NUCTHIA code, the investigations of effects of main system parameters on BWW core stability have also been made. All the results are consistent with the experimental data

Seismic risk assessment of a BWR: status report

International Nuclear Information System (INIS)

Chuang, T.Y.; Bernreuter, D.L.; Wells, J.E.; Johnson, J.J.

1985-02-01

The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and analyzed. The SSMRP methodology can equally be applied to a boiling water reactor (BWR). To demonstrate its applicability, to identify fundamental differences in seismic risk between a PWR and a BWR, and to provide a basis of comparison of seismic risk between a PWR and a BWR when analyzed with comparable methodology and assumptions, a seismic risk analysis is being performed on the LaSalle County Station nuclear power plant

An A BWR demonstration simulator for training and developing technical staff

International Nuclear Information System (INIS)

Powers, J.; Yonezawa, H.; Aoyagi, Y.; Kataoka, K.

2015-09-01

The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

An A BWR demonstration simulator for training and developing technical staff

Energy Technology Data Exchange (ETDEWEB)

Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Yonezawa, H.; Aoyagi, Y.; Kataoka, K., E-mail: jim.powers@toshiba.com [Toshiba Corporation, Kawasaki, Kanagawa (Japan)

2015-09-15

The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

Maximum thermal loading test of BWR fuel assembly

International Nuclear Information System (INIS)

Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

1987-01-01

Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

Development of next BWR plant

International Nuclear Information System (INIS)

Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

1995-01-01

It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

Development of next BWR plant

Energy Technology Data Exchange (ETDEWEB)

Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

1995-04-01

It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

Strategies of operation cycles in BWR type reactors

International Nuclear Information System (INIS)

Molina, D.; Sendino, F.

1996-01-01

The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

Reducing BWR O and M costs through on-line performance monitoring

International Nuclear Information System (INIS)

Jonas, T.; Gross, R.; Logback, F.; Josyula, R.

1995-01-01

Competition in the electric power industry has placed significant emphasis on reducing operating and maintenance (O and M) costs at nuclear facilities. Therefore, on-line performance monitoring to locate power losses for boiling water reactor (BWR) plants is creating tremendous interest. In addition, the ability to automate activities such as data collection, analysis, and reporting increases the efficiency of plant engineers and gives them more time to concentrate on solving plant efficiency problems. This capability is now available with a unique software product called GEBOPS. GE Nuclear Energy, in conjunction with Joint Venture partner Black and Veatch, has undertaken development of the General Electric/Black and Veatch On-line Performance System (GEBOPS), an on-line performance monitoring system for BWR plants. The experience and expertise of GE Nuclear Energy with BWR plants, coupled with the proven on-line monitoring software development experience and capability of Black and Veatch, provide the foundation for a unique product which addresses the needs of today's BWR plants

Standard Technical Specifications, General Electric plants, BWR/4

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power specifications. This report contains three volumes. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS

Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island Design (Docket No. 50-447). Supplement No. 3

International Nuclear Information System (INIS)

1985-01-01

Supplement 3 to the Safety Evaluation Report (SER) for the application filed by General Electric Company for the final design approval for the GE BWR/6 nuclear island design has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report supplements the GESSAR II SER (NUREG-0979), issued in April 1983, summarizing the results of the staff's safety review of the GESSAR II BWR/6 nuclear island design. Subject to favorable resolution of the items discussed in this supplement, the staff concludes that the GESSAR II design satisfactorily addresses the severe-accident concerns described in draft NUREG-1070

An ecological interface design for BWR nuclear power plants

International Nuclear Information System (INIS)

Monta, K.; Itoh, J.

1992-01-01

An ecological interface design was applied to realize the support function for the operator's direct perception and analytical reasoning in the development of an intelligent man-machine system for BWR nuclear power plants. The abstraction-aggregation functional hierarchy representation of the work domain is a base of the ecological interface design. Another base is the concept of the level of cognitive control. The former was mapped into the interface to externalize the operator's normative mental model of the plants, which will reduce his/her cognitive work load and support knowledge-based problem solving. In addition, the same framework can be used for the analytical evaluation of man-machine interfaces. The information content and structure of a prototype interface were evaluated. This approach seems promising from these experiences. (author)

BWR AXIAL PROFILE

International Nuclear Information System (INIS)

Huffer, J.

2004-01-01

The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

Nuclear energy

International Nuclear Information System (INIS)

2007-01-01

This digest document was written by members of the union of associations of ex-members and retired people of the Areva group (UARGA). It gives a comprehensive overview of the nuclear industry world, starting from radioactivity and its applications, and going on with the fuel cycle (front-end, back-end, fuel reprocessing, transports), the nuclear reactors (PWR, BWR, Candu, HTR, generation 4 systems), the effluents from nuclear facilities, the nuclear wastes (processing, disposal), and the management and safety of nuclear activities. (J.S.)

Quantitative evaluation for training results of nuclear plant operator on BWR simulator

International Nuclear Information System (INIS)

Sato, Takao; Sato, Tatsuaki; Onishi, Hiroshi; Miyakita, Kohji; Mizuno, Toshiyuki

1985-01-01

Recently, the reliability of neclear power plants has largely risen, and the abnormal phenomena in the actual plants are rarely encountered. Therefore, the training using simulators becomes more and more important. In BWR Operator Training Center Corp., the training of the operators of BWR power plants has been continued for about ten years using a simulator having the nearly same function as the actual plants. The recent high capacity ratio of nuclear power plants has been mostly supported by excellent operators trained in this way. Taking the opportunity of the start of operation of No.2 simulator, effort has been exerted to quantitatively grasp the effect of training and to heighten the quality of training. The outline of seven training courses is shown. The technical ability required for operators, the items of quantifying the effect of training, that is, operational errors and the time required for operation, the method of quantifying, the method of collecting the data and the results of the application to the actual training are described. It was found that this method is suitable to quantify the effect of training. (Kako, I.)

Compact modular BWR (CM-BWR)

International Nuclear Information System (INIS)

Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

2003-01-01

A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

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

Review of international solutions to NEACRP benchmark BWR lattice cell problems

International Nuclear Information System (INIS)

Halsall, M.J.

1977-12-01

This paper summarises international solutions to a set of BWR benchmark problems. The problems, posed as an activity sponsored by the Nuclear Energy Agency Committee on Reactor Physics, were as follows: 9-pin supercell with central burnable poison pin, mini-BWR with 4 pin-cells and water gaps and control rod cruciform, full 7 x 7 pin BWR lattice cell with differential U 235 enrichment, and full 8 x 8 pin BWR lattice cell with water-hole, Pu-loading, burnable poison, and homogenised cruciform control rod. Solutions have been contributed by Denmark, Japan, Sweden, Switzerland and the UK. (author)

Synergistic failure of BWR internals

International Nuclear Information System (INIS)

Ware, A. G.; Chang, T.Y.

1999-01-01

Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

Utility experience with BWR-PSMS

International Nuclear Information System (INIS)

Bond, G.R.

1986-01-01

The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

Investigation of control rod worth and nuclear end of life of BWR control rods

International Nuclear Information System (INIS)

Magnusson, Per

2008-01-01

This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of 10 B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% 10 B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in 10 B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming

BWR plant analyzer development at BNL

International Nuclear Information System (INIS)

Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

1985-01-01

Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

Complete BWR--EM LOCA analysis using the WRAP--EM system

International Nuclear Information System (INIS)

Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

1979-01-01

The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR; BUTREN-RC un sistema hibrido para la optimizacion de recargas de combustible nuclear en un BWR

Energy Technology Data Exchange (ETDEWEB)

Ortiz S, J.J.; Castillo M, J.A. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico); Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

2004-07-01

The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

BWR power oscillation evaluation methodologies in core design

International Nuclear Information System (INIS)

Hotta, Akitoshi

1995-01-01

At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

BWR Refill-Reflood Program. Final report

International Nuclear Information System (INIS)

Myers, L.L.

1983-09-01

The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

BWR and PWR chemistry operating experience and perspectives

International Nuclear Information System (INIS)

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

2014-01-01

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

Nuclear energy; Le nucleaire

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

This digest document was written by members of the union of associations of ex-members and retired people of the Areva group (UARGA). It gives a comprehensive overview of the nuclear industry world, starting from radioactivity and its applications, and going on with the fuel cycle (front-end, back-end, fuel reprocessing, transports), the nuclear reactors (PWR, BWR, Candu, HTR, generation 4 systems), the effluents from nuclear facilities, the nuclear wastes (processing, disposal), and the management and safety of nuclear activities. (J.S.)

Nuclear fuel activity with minor actinides after their useful life in a BWR; Actividad del combustible nuclear con actinidos menores despues de su vida util en un reactor BWR

Energy Technology Data Exchange (ETDEWEB)

Martinez C, E.; Ramirez S, J. R.; Alonso V, G., E-mail: eduardo.martinez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

Nuclear fuel used in nuclear power reactors has a life cycle, in which it provides energy, at the end of this cycle is withdrawn from the reactor core. This used fuel is known as spent nuclear fuel, a strong problem with this fuel is that when the fuel was irradiated in a nuclear reactor it leaves with an activity of approximately 1.229 x 10{sup 15} Bq. The aim of the transmutation of actinides from spent nuclear fuel is to reduce the activity of high level waste that must be stored in geological repositories and the lifetime of high level waste; these two achievements would reduce the number of necessary repositories, as well as the duration of storage. The present work is aimed at evaluating the activity of a nuclear fuel in which radioactive actinides could be recycled to remove most of the radioactive material, first establishing a reference of actinides production in the standard nuclear fuel of uranium at end of its burning in a BWR, and a fuel rod design containing 6% of actinides in an uranium matrix from the enrichment tails is proposed, then 4 standard uranium fuel rods are replaced by 4 actinide bars to evaluate the production and transmutation of the same, finally the reduction of actinide activity in the fuel is evaluated. (Author)

Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

International Nuclear Information System (INIS)

Nagano, M.; Sakurai, S.; Yamaguchi, H.

1997-01-01

MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

Standard Technical Specifications, General Electric Plants, BWR/6

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

Standard Technical Specifications, General Electric plants, BWR/4

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

Analysis of the integrity of the pressure vessel of the BWR type nuclear reactor

International Nuclear Information System (INIS)

Silva Luna, O.

1982-01-01

The presssure vessel of a BWR type reactor was monitored for cracking during alternating events of its in-service life. The monitoring was to determine criticality of fractures catastrophic fractures and the velocity of fracture propagation. Detected cracks were evaluated as specified in ASME code section XI, of a minimum wall thickness of 2.5% crack growths were compared a) of 1/10 of the critical maximum size and b) at in-service inspection intervals according to ASME recommendations to be established at the Laguna Verde nuclear plant. Finally conclusions are made and discussed. (author)

An overview of the BWR ECCS strainer blockage issues

International Nuclear Information System (INIS)

Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R.

1996-01-01

This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, open-quotes Containment Emergency Sump Performance,close quotes and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts

Safety analysis of thorium-based fuels in the General Electric Standard BWR

International Nuclear Information System (INIS)

Colby, M.J.; Townsend, D.B.; Kunz, C.L.

1980-06-01

A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

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)

Power plant and utility performance: how world-record outages are being achieved in the USA

International Nuclear Information System (INIS)

Anon.

1997-01-01

Two record-breaking refuelling outages at power reactors in the USA are described. The first, at Browns Ferry 3 BWR, was accomplished in 19 days 39 minutes - a shorter time than for an General Electric BWR anywhere in the world hitherto. The management attribute this success to planning, personnel and performance. As well as refuelling, inspections and maintenance, major modifications were carried out. These included the completion of the installation of digital feedwater reactor level control and digital feedwater heater level control. The second outage, at South Texas Project 2 BWR, at 17 days 14 hours and 10 minutes was the fastest yet recorded for any US nuclear unit. This achievement is ascribed to excellent outage preparation and scheduling, the superior condition of the plant equipment and teamwork and safety consciousness on behalf of the plant personnel. Finally, brief consideration is given to the nuclear performance recovery programme of Commonwealth Edison and Ontario Hydro Nuclear. (UK)

BWR control blade replacement strategies

Energy Technology Data Exchange (ETDEWEB)

Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

2000-02-01

The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B{sub 4}C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

BWR control blade replacement strategies

International Nuclear Information System (INIS)

Kennard, M.W.; Harbottle, J.E.

2000-01-01

The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B 4 C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

Nuclear fuel activity with minor actinides after their useful life in a BWR

International Nuclear Information System (INIS)

Martinez C, E.; Ramirez S, J. R.; Alonso V, G.

2016-09-01

Nuclear fuel used in nuclear power reactors has a life cycle, in which it provides energy, at the end of this cycle is withdrawn from the reactor core. This used fuel is known as spent nuclear fuel, a strong problem with this fuel is that when the fuel was irradiated in a nuclear reactor it leaves with an activity of approximately 1.229 x 10 15 Bq. The aim of the transmutation of actinides from spent nuclear fuel is to reduce the activity of high level waste that must be stored in geological repositories and the lifetime of high level waste; these two achievements would reduce the number of necessary repositories, as well as the duration of storage. The present work is aimed at evaluating the activity of a nuclear fuel in which radioactive actinides could be recycled to remove most of the radioactive material, first establishing a reference of actinides production in the standard nuclear fuel of uranium at end of its burning in a BWR, and a fuel rod design containing 6% of actinides in an uranium matrix from the enrichment tails is proposed, then 4 standard uranium fuel rods are replaced by 4 actinide bars to evaluate the production and transmutation of the same, finally the reduction of actinide activity in the fuel is evaluated. (Author)

Experimental and analytical studies for a BWR nuclear reactor building. Evaluation of soil-structure interaction behaviour

International Nuclear Information System (INIS)

Mizuno, N.; Tsushima, Y.

1975-01-01

This paper evaluates the spatial characteristics of dynamic properties, especially soil-structure interaction behaviour, of the BWR nuclear building by experimental and analytical studies. It is well known that the damping effects in soil-structure interaction are remarkable on the building with short periods by the dissipation of vibrational energy to the soil. The authors have previously reported an analytical method for estimating the damping effects the properties of which are characterized as follows: 1) The complex damping is used, because the so-called structural damping may be more suitable for estimating the damping effects of an elastic structure. 2) H. Tajimi's theory is used for estimating the dynamical soil-foundation stiffness with the dissipation of vibrational energy on the elastic half-space soil. In this paper, an approximate explanation is presented in regard to the more developmental mathematical method for estimating the damping effects than the above-mentioned previous method, which is 'Modes Superposition Method for Multi-Degrees of Freedom System' with the constant complex stiffness showing the structural damping effects and the dynamical soil-foundation stiffness approximated by the linear or quadratic functions of the eigenvalues. An approximate explanation is presented in regard to the experimental results of the No. 1 reactor building (BWR) of Hamaoka Nuclear Power Station, The Chubu Electric Power Co., Ltd. (Auth.)

Development of RBWR (Resource-renewable BWR) for environmental burden reduction of radioactive wastes

International Nuclear Information System (INIS)

Hino, Tetsushi; Ohtsuka, Masaya; Moriya, Kumiaki; Matsuura, Masayoshi

2014-01-01

Accumulation of long-life transuranium elements produced as by-products with uranium fuel burning became an issue of nuclear power. Hitachi had been developing the reactor with transuranium elements burning as fuels based on BWR type reactors successfully used as commercial reactors: RBWR (Resource-renewable BWR). Efficient transmutation and fissioning of transuranium elements needed adjustment of in-core neutron energy spectra distribution better for nuclear reaction of transuranium elements. Taking advantage of characteristics of BWR type reactors with neutron spectra hardening more easily adjustable than other type of reactors, multiple recycling and fissioning transuranium elements as fuels could make environmental burden reduction of radioactive wastes and efficient use of resources compatible. This article described the concept and history of RBWR and showed its specifications and reactor core characteristics. (T. Tanaka)

BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR

International Nuclear Information System (INIS)

Ortiz S, J.J.; Castillo M, J.A.; Valle G, E. del

2004-01-01

The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

World nuclear energy paths

International Nuclear Information System (INIS)

Connolly, T.J.; Hansen, U.; Jaek, W.; Beckurts, K.H.

1979-01-01

In examing the world nuclear energy paths, the following assumptions were adopted: the world economy will grow somewhat more slowly than in the past, leading to reductions in electricity demand growth rates; national and international political impediments to the deployment of nuclear power will gradually disappear over the next few years; further development of nuclear power will proceed steadily, without serious interruption but with realistic lead times for the introduction of advanced technologies. Given these assumptions, this paper attempts a study of possible world nuclear energy developments, disaggregated on a regional and national basis. The scenario technique was used and a few alternative fuel-cycle scenarios were developed. Each is an internally consistent model of technically and economically feasible paths to the further development of nuclear power in an aggregate of individual countries and regions of the world. The main purpose of this modeling exercise was to gain some insight into the probable international locations of reactors and other nuclear facilities, the future requirements for uranium and for fuel-cycle services, and the problems of spent-fuel storage and waste management. The study also presents an assessment of the role that nuclear power might actually play in meeting future world energy demand

BWR radiation buildup control with ionic zinc

International Nuclear Information System (INIS)

Marble, W.J.; Wood, C.J.; Leighty, C.E.; Green, T.A.

1986-01-01

In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

Transmutation of minor actinide using BWR fueled mixed oxide

International Nuclear Information System (INIS)

Susilo, Jati

2000-01-01

Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

World nuclear performance report 2016. A new study by World Nuclear Association

Energy Technology Data Exchange (ETDEWEB)

Cobb, Jonathan [World Nuclear Association, London (United Kingdom)

2016-08-15

A larger number of nuclear power units are under construction than at any other time in the last 25 years, and with another ten new reactors coming online 2015 demonstrated improving new build performance all round. The existing global fleet, totally 439 by year-end, generated roughly 10 % of the world's electricity, making up around one-third of the world's low-carbon electricity supply. Nevertheless, there are challenges ahead for the global nuclear industry. The World Nuclear Association's vision for the future global electricity system consists of a diverse mix of low-carbon technologies - where renewables, nuclear and a fossil fuels work together in harmony to ensure a reliable, affordable and clean energy supply.

World nuclear performance report 2016. A new study by World Nuclear Association

International Nuclear Information System (INIS)

Cobb, Jonathan

2016-01-01

A larger number of nuclear power units are under construction than at any other time in the last 25 years, and with another ten new reactors coming online 2015 demonstrated improving new build performance all round. The existing global fleet, totally 439 by year-end, generated roughly 10 % of the world's electricity, making up around one-third of the world's low-carbon electricity supply. Nevertheless, there are challenges ahead for the global nuclear industry. The World Nuclear Association's vision for the future global electricity system consists of a diverse mix of low-carbon technologies - where renewables, nuclear and a fossil fuels work together in harmony to ensure a reliable, affordable and clean energy supply.

BWR Water Chemistry Guidelines: 1993 Revision, Normal and hydrogen water chemistry

International Nuclear Information System (INIS)

Karlberg, G.; Goddard, C.; Fitzpatrick, S.

1994-02-01

The goal of water chemistry control is to extend the operating life of the reactor and rector coolant system, balance-of-plant components, and turbines while simultaneously controlling costs to safeguard the continued economic viability of the nuclear power generation investment. To further this goal an industry committee of chemistry personnel prepared guidelines to identify the benefits, risks, and costs associated with water chemistry in BWRs and to provide a template for an optimized water chemistry program. This document replaces the BWR Normal Water Chemistry Guidelines - 1986 Revision and the BWR Hydrogen Water Chemistry Guidelines -- 1987 Revision. It expands on the previous guidelines documents by covering the economic implications of BWR water chemistry control

BWR mechanics and materials technology update

International Nuclear Information System (INIS)

Kiss, E.

1983-01-01

This paper discusses technical results obtained from a variety of important programs underway at General Electric's Nuclear Engineering Division. The principal objective of these programs is to qualify and improve BWR product related technologies that fall broadly under the disciplines of Applied Mechanics and Materials Engineering. The paper identifies and deals with current technical issues that are of general importance to the LWR industry albeit the specific focus is directed to the development and qualification of analytical predictive methods and criteria, and improved materials for use in the design of the BWR. In this paper, specific results and accomplishments are summarized to provide a braod perspective of technology advances. Results are presented in sections which discuss: dynamic analysis and modeling; fatigue and fracture evaluation; materials engineering advances; and flow induced vibration. (orig.)

Seismic PRA of a BWR plant

International Nuclear Information System (INIS)

Nishio, Masahide; Fujimoto, Haruo

2014-01-01

Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

BWR shutdown analyzer using artificial intelligence (AI) techniques

International Nuclear Information System (INIS)

Cain, D.G.

1986-01-01

A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

Tritium in liquid phase in a BWR-5 like Laguna Verde; Tritio en fase liquida en un BWR-5 como Laguna Verde

Energy Technology Data Exchange (ETDEWEB)

Tijerina S, F.; Vargas A, A.; Cardenas J, J., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Veracruz-Medellin Km 7.5, Veracruz (Mexico)

2011-11-15

In boiling water reactors (BWR), the tritium (H{sub 3}) takes place mainly as a result of ternary fissions in the nuclear reactors, of those which 75% are in gaseous form and 25% in liquid form. In the liquid phase, the tritium is transported to the pipes of the primary coolant toward condensed tanks or tanks of drainage excesses of radioactive equipment, located in external areas of a BWR, as well as to the processes of radioactive wastes to be able to be directed to the liquid effluents. For that reason, is necessary to know the possible routes of the transport and processes of the tritium in a BWR to control this radioisotope in the site of the event of leaks in equipment s and buried pipes, avoiding that emigrates toward underground flowing and an impact to the environment and to the people in general. (Author)

Experiences with monitoring and control of microbiological growth in the standby service water system of a BWR nuclear power plant

International Nuclear Information System (INIS)

Zisson, P.S.; Whitaker, J.M.; Neilson, H.L.; Mayne, L.L.

1995-01-01

In 1989, the Unites States Nuclear Regulatory Commission formally recognized the potential for nuclear accidents resulting from microbiological causes. Such causes range from loss of heat transfer due to microbiological fouling, to loss of system integrity caused by microbiologically influenced corrosion (MIC). As a result of these potential problems, monitoring, mitigation, and control procedures must be developed by all regulated plants. In developing a control and mitigation strategy for the standby service water system of a boiling water reactor (BWR) nuclear power plant, numerous monitoring techniques were employed to evaluate effectiveness. This paper describes the monitoring techniques that were evaluated, and those that ultimately proved to be effective

The BWR owners' group planning guide for life extension

International Nuclear Information System (INIS)

Smith, S.K.; Lehnert, D.F.; Locke, R.K.

1991-01-01

Extending the operating life of a commercial nuclear power plant has been shown to be economically beneficial to both the utility and the electric customer. As such, many utilities are planning and implementing plant life extension (PLEX) programs. A document has been developed which provides guidance to utilities in formulating a PLEX program plant for one or more boiling water reactor (BWR) plants. The guide has been developed by the BWR Owners' Group Plant Life Extension Committee. The principal bases for this guide were the BWR Pilot and Lead Plant Programs. These programs were used as models to develop the 'base plan' described in this guide. By formulating their program plant utilizing the base plan, utilities will be able to maximize the use of existing evaluations and results. The utility planner will build upon the base plan by adding any tasks or features that are unique to their programs. (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

Operational experience of human-friendly control and instrumentation systems for BWR nuclear power plants

International Nuclear Information System (INIS)

Makino, M.; Watanabe, T.; Suto, O.; Asahi, R.

1987-01-01

In recent BWR nuclear power plants in Japan, an advanced centralized monitoring and control system PODIA (Plant Operation by Displayed Information and Automation), which incorporates many operator aid functions, has been in operation since 1985. Main functions of the PODIA system as a computerized operator aid system are as follows. CRT displays for plant monitoring. Automatic controls and operation guides for plant operation. Stand-by status monitoring for engineered safety features during normal operation. Surveillance test procedure guides for engineered safety features. Integrated alarm display. The effectiveness of these functions have been proved through test and commercial operation. It has been obtained that operators have preferred PODIA much more than conventional monitoring and control systems

World status - nuclear power

International Nuclear Information System (INIS)

Holmes, A.

1984-01-01

The problems of nuclear power are not so much anti-nuclear public opinion, but more the decrease of electricity consumption growth rate and the high cost of building reactors. Because of these factors, forecasts of world nuclear capacity have had to be reduced considerably over the last three years. The performance of reactors is considered. The CANDU reactor remains the world's best performer and overall tends to out-perform larger reactors. The nuclear plant due to come on line in 1984 are listed by country; this shows that nuclear capacity will increase substantially over a short period. At a time of stagnant demand this will make nuclear energy an important factor in the world energy balance. Nuclear power stations in operation and under construction in 1983 are listed and major developments in commercial nuclear power in 1983 are taken country by country. In most, the report is the same; national reactor ordering cut back because the expected increase in energy demand has not happened. Also the cost-benefit of nuclear over other forms of energy is no longer as favourable. The export opportunities have also declined as many of the less developed countries are unable to afford reactors. (U.K.)

Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide; Analisis del impacto de modificar la composicion del combustible nuclear de un BWR con oxido de berilio

Energy Technology Data Exchange (ETDEWEB)

Gallardo V, J. M.; Morales S, J. B., E-mail: euqrop@hotmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

2013-10-15

The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO{sub 2}) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO{sub 2} mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

Development of a BWR core burn-up calculation code COREBN-BWR

International Nuclear Information System (INIS)

Morimoto, Yuichi; Okumura, Keisuke

1992-05-01

In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island design, Docket No. 50-447

International Nuclear Information System (INIS)

1983-04-01

The Safety Evaluation Report for the application filed by General Electric Company for the Final Design Approval for the General Electric Standard Safety Analysis Report (GESSAR II FSAR) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report summarizes the results of the staff's safety review of the GESSAR II BWR/6 Nuclear Island Design. Subject to favorable resolution of items discussed in the Safety Evaluation Report, the staff concludes that the facilities referencing GESSAR II, subject to approval of the balance-of-plant design, can conform with the provisions of the Act and the regulations of the Nuclear Regulatory Commission

BWR zinc addition Sourcebook

International Nuclear Information System (INIS)

Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

2014-01-01

Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

Scaling and uncertainty in BWR instability problems

International Nuclear Information System (INIS)

Di Auria, F.; Pellicoro, V.

1995-01-01

This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

Design and axial optimization of nuclear fuel for BWR reactors

International Nuclear Information System (INIS)

Garcia V, M.A.

2006-01-01

In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

Nuclear Power Plants in the World

International Nuclear Information System (INIS)

2003-01-01

The Japan Atomic Industrial Forum (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Nuclear power plants in the world'. In this report, some data at the end of 2002 was made up on bases of answers on questionnaires from 65 electric power companies and other nuclear organizations in 28 countries and regions around the world by JAIF. This report is comprised of 19 items, and contains generating capacity of the plants; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; status of MOX use in the world; location of the plants; the plants in the world; directory of the plants; nuclear fuel cycle facilities; and so forth. (J.P.N.)

The HAMBO BWR simulator of HAMMLAB

International Nuclear Information System (INIS)

Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

2001-02-01

Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

Nuclear power plants in the world

International Nuclear Information System (INIS)

2008-01-01

The Japan Atomic Industrial Forum, Inc. (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Nuclear power plants in the world'. In this report, some data at the end of 2007/2008 was made up on bases of answers on questionnaires from electric power companies and other nuclear organizations around the world by JAIF. This report is comprised of 18 items, and contains generating capacity of the plants; effect of the Niigata-ken chuetsu-oki earthquake; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; status of MOX use in the world; location of the plants; the plants in the world; directory of the plants; nuclear fuel cycle facilities, and so forth. (J.P.N.)

Nuclear Power Plants in the World

International Nuclear Information System (INIS)

2004-01-01

The Japan Atomic Industrial Forum, Inc. (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Nuclear power plants in the world'. In this report, some data at the end of 2003 was made up on bases of answers on questionnaires from 81 electric power companies and other nuclear organizations in 33 countries and regions around the world by JAIF. This report is comprised of 19 items, and contains generating capacity of the plants; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; status of MOX use in the world; location of the plants; the plants in the world; directory of the plants; nuclear fuel cycle facilities; and so forth. (J.P.N.)

Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

2015-01-01

The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

Experimental and analytical studies for a BWR nuclear reactor building evaluation of soil-structure interaction behavior

International Nuclear Information System (INIS)

Mizuno, N.; Tsushima, Y.

1975-01-01

The purpose of this paper is to evaluate the spatial characteristics of dynamic properties, especially soil-structure interaction behavior, or the BWR nuclear reactor building by experimental and analytical studies. An analytical method (SMIRT-1 Paper K 2/4) for estimating the damping effects is reported. The complex damping is used, because the so-called structural damping may be more suitable for estimating the damping effects of an elastic structure. H. Tajimi's theory is used for estimating the dynamical soil-foundation stiffness with the dissipation of vibrational energy on the elastic half-space soil. An approximate explanation is presented in regard to the more developmental mathematical method for estimating the damping effects than the above-mentioned previous method, which is 'Modes Superposition Method for Multi-Degrees of Freedom System' with the constant complex stiffness showing the structural damping effects and the dynamical soil-foundation stiffness approximated by the linear or quadratic functions of the eigenvalues. Next, an approximate explanation is presented in regard to the experimental results of the No.1 reactor building (BWR) of Hamaoka Nuclear Power Station, The Chubu Electric Power Co., Ltd. The regression analyses of the experimental resonance curves by one degree system show that the critical damping ratio is larger than the 0.10 used in the design for the fundamental natural period. It is attempted to simulate the experimental results by the above-mentioned method. The simulated model is a fourty-eight degrees of freedom spring mass system because of the eight masses for the eight floors including the base foundation and the six degrees of freedom for a mass

Stability analysis of BWR nuclear-coupled thermal-hyraulics using a simple model

Energy Technology Data Exchange (ETDEWEB)

Karve, A.A.; Rizwan-uddin; Dorning, J.J. [Univ. of Virginia, Charlottesville, VA (United States)

1995-09-01

A simple mathematical model is developed to describe the dynamics of the nuclear-coupled thermal-hydraulics in a boiling water reactor (BWR) core. The model, which incorporates the essential features of neutron kinetics, and single-phase and two-phase thermal-hydraulics, leads to simple dynamical system comprised of a set of nonlinear ordinary differential equations (ODEs). The stability boundary is determined and plotted in the inlet-subcooling-number (enthalpy)/external-reactivity operating parameter plane. The eigenvalues of the Jacobian matrix of the dynamical system also are calculated at various steady-states (fixed points); the results are consistent with those of the direct stability analysis and indicate that a Hopf bifurcation occurs as the stability boundary in the operating parameter plane is crossed. Numerical simulations of the time-dependent, nonlinear ODEs are carried out for selected points in the operating parameter plane to obtain the actual damped and growing oscillations in the neutron number density, the channel inlet flow velocity, and the other phase variables. These indicate that the Hopf bifurcation is subcritical, hence, density wave oscillations with growing amplitude could result from a finite perturbation of the system even where the steady-state is stable. The power-flow map, frequently used by reactor operators during start-up and shut-down operation of a BWR, is mapped to the inlet-subcooling-number/neutron-density (operating-parameter/phase-variable) plane, and then related to the stability boundaries for different fixed inlet velocities corresponding to selected points on the flow-control line. The stability boundaries for different fixed inlet subcooling numbers corresponding to those selected points, are plotted in the neutron-density/inlet-velocity phase variable plane and then the points on the flow-control line are related to their respective stability boundaries in this plane.

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

Energy Technology Data Exchange (ETDEWEB)

Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

2015-01-01

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k_eff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

Nuclear danger in the modern world

International Nuclear Information System (INIS)

Sulejmenov, O.O.

2000-01-01

It is noted, that nowadays a nuclear danger proceeds from nuclear depositions of countries having own nuclear weapons. Since Kazakhstan is one of the first country in the world which fulfilled regulations of Lisbon Protocol and liquidated own nuclear potential, author regards that Kazakhstan have moral right for initiating process of attachment to Comprehensive Nuclear Test Ban Treaty by countries having nuclear weapon. Now for Kazakhstan there are urgent problems: financing of post-conversion processes; re-cultivation of territory contaminated by residuals from nuclear weapons test; rehabilitation of population health, damaged from test of mass destruction weapon. Scientists of Kazakhstan estimated damage from nuclear test on Kazakstan territory in 10 billion dollars. It is necessary international efforts of all public organizations of the world for all world sites. One of the financing source could be means from reduction of nuclear arms production

Report on the BWR owners group radiation protection/ALARA Committee

International Nuclear Information System (INIS)

Aldrich, L.R.

1995-01-01

Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming open-quotes World Classclose quotes performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance

Report on the BWR owners group radiation protection/ALARA Committee

Energy Technology Data Exchange (ETDEWEB)

Aldrich, L.R. [Commonwealth Edison Co., Downers Grove, IL (United States)

1995-03-01

Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming {open_quotes}World Class{close_quotes} performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance.

Developing and modeling of the 'Laguna Verde' BWR CRDA benchmark

International Nuclear Information System (INIS)

Solis-Rodarte, J.; Fu, H.; Ivanov, K.N.; Matsui, Y.; Hotta, A.

2002-01-01

Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant - unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The 'Laguna Verde' (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTREE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions

Advanced chemistry management system to optimize BWR chemistry control

International Nuclear Information System (INIS)

Maeda, K.; Nagasawa, K.

2002-01-01

BWR plant chemistry control has close relationships among nuclear safety, component reliability, radiation field management and fuel integrity. Advanced technology is required to improve chemistry control [1,3,6,7,10,11]. Toshiba has developed TACMAN (Toshiba Advanced Chemistry Management system) to support BWR chemistry control. The TACMAN has been developed as response to utilities' years of requirements to keep plant operation safety, reliability and cost benefit. The advanced technology built into the TACMAN allows utilities to make efficient chemistry control and to keep cost benefit. TACMAN is currently being used in response to the needs for tools those plant chemists and engineers could use to optimize and identify plant chemistry conditions continuously. If an incipient condition or anomaly is detected at early stage, root causes evaluation and immediate countermeasures can be provided. Especially, the expert system brings numerous and competitive advantages not only to improve plant chemistry reliability but also to standardize and systematize know-how, empirical knowledge and technologies in BWR chemistry This paper shows detail functions of TACMAN and practical results to evaluate actual plant. (authors)

World supply of nuclear energy

International Nuclear Information System (INIS)

Pecqueur, Michel.

1981-01-01

At the end of 1980 nuclear energy accounted for 9% of the world production of electricity stemming from 262 power stations, utilising mainly the process of water reactors and representing an installed capacity of 142 GWe. This production, apparently limited, already represents the equivalent of 150 million TOE. The 600 nuclear power stations in service, under construction or ordered represent a total of 450 GWe. In 1985, their production ought to cover 15% of the world requirements of electricity, which corresponds to a doubling of the share of nuclear energy within 6 years. During these recent years, the development of nuclear energy has undergone a significant slowing down and the number of orders for new nuclear power stations has dropped considerably in particular in the United States. Considering the time required and the available industrial capacity, the accumulated capacity which could be installed worlwide by 1990 could attain 530 GWe, equivalent to 650 MTOE covering 24% of the world production of electricity and 7% of the world consumption of primary energy. A determined effort for the end of this century could end up by the installation of 1200 GWe of capacity, generating 1.5 GTOE. The share of nuclear energy would then represent 35% of the production of electricity [fr

Power oscillations in BWR reactors

International Nuclear Information System (INIS)

Espinosa P, G.

2002-01-01

One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

The unstoppable world nuclear development

International Nuclear Information System (INIS)

Dominguez, M. T.

2009-01-01

To meet energy needs and curb climate change, the number of reactors will continue to increase because more and more countries are going the need nuclear power. At present, there are 436 nuclear reactors in the world that produce 16% of the electricity, and another 48 units are under construction in all, 31 countries in the world use nuclear power to produce electricity, and some countries that do not have reactors, e.g. Poland and Italy, are seriously planning to include nuclear power in their energy mix. Global nuclear development is a reality; energy and environmental challenges have led to new support for nuclear power, which is a safe, stable emission-free source. (Author)

Tritium in liquid phase in a BWR-5 like Laguna Verde

International Nuclear Information System (INIS)

Tijerina S, F.; Vargas A, A.; Cardenas J, J.

2011-11-01

In boiling water reactors (BWR), the tritium (H 3 ) takes place mainly as a result of ternary fissions in the nuclear reactors, of those which 75% are in gaseous form and 25% in liquid form. In the liquid phase, the tritium is transported to the pipes of the primary coolant toward condensed tanks or tanks of drainage excesses of radioactive equipment, located in external areas of a BWR, as well as to the processes of radioactive wastes to be able to be directed to the liquid effluents. For that reason, is necessary to know the possible routes of the transport and processes of the tritium in a BWR to control this radioisotope in the site of the event of leaks in equipment s and buried pipes, avoiding that emigrates toward underground flowing and an impact to the environment and to the people in general. (Author)

Seismic risk assessment of a BWR

International Nuclear Information System (INIS)

Wells, J.E.; Bernreuter, D.L.; Chen, J.C.; Lappa, D.A.; Chuang, T.Y.; Murray, R.C.; Johnson, J.J.

1987-01-01

The simplified seismic risk methodology developed in the USNRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant (PWR). The simplified seismic risk methodology was developed to reduce the costs associated with a seismic risk analysis while providing adequate results. A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models, was developed and used in assessing the seismic risk of the Zion nuclear power plant (FSAR). The simplified seismic risk methodology was applied to the LaSalle County Station nuclear power plant, a BWR; to further demonstrate its applicability, and if possible, to provide a basis for comparing the seismic risk from PWRs and BWRs. (orig./HP)

Nuclear Power Plants in the World

International Nuclear Information System (INIS)

2000-01-01

The Japan Atomic Industrial Forum (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Developmental trends on nuclear power plants in the world'. In this report, some data at the end of 1999 was made up on bases of answers on questionnaires from 72 electric companies in 31 nations and regions in the world by JAIF. This report is comprised of 19 items, and contains generating capacity of the plants; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; location of the plants; the plants in the world; and so forth. And, it also has some survey results on the 'Liberalization of electric power markets and nuclear power generation' such as some 70% of respondents in nuclear power for future option, gas-thermal power seen as power source with most to gain from liberalization, merits on nuclear power generation (environmental considerations and supply stability), most commonly voiced concern about new plant orders in poor economy, and so forth. (G.K.)

Recent operating experience during startup testing at latest 1100 MWe BWR-5 nuclear power plants

International Nuclear Information System (INIS)

Tanabe, Akira; Tateishi, Mizuo; Kajikawa, Makoto; Hayase, Yuichi.

1986-01-01

In June and September 1985, the latest two 1100 Mwe BWR-5 nuclear power plants started commercial operation about ten days earlier than initially expected without any unscheduled shutdown. These latest plants, 2F-3 and K-1, are characterized by an improved core with new 8 x 8 fuel assemblies, highly reliable control systems, advanced control room system and turbine steam full bypass system for full load rejection (2F3). This paper describes the following operating experiences gained during their startup testing. 1) Continuous operation at full load rejection. 2) Stable operation at natural circulating flow condition. 3) 31 and 23 hour short time start up operation. 4) 100-75-100 %, 1-8-1-14 hours daily load following operation. (author)

Nuclear-coupled thermal-hydraulic nonlinear stability analysis using a novel BWR reduced order model. Pt. 1. The effects of using drift flux versus homogeneous equilibrium models

International Nuclear Information System (INIS)

Dokhane, A.; Henning, D.; Chawla, R.; Rizwan-Uddin

2003-01-01

BWR stability analysis at PSI, as at other research centres, is usually carried out employing complex system codes. However, these do not allow a detailed investigation of the complete manifold of all possible solutions of the associated nonlinear differential equation set. A novel analytical, reduced order model for BWR stability has been developed at PSI, in several successive steps. In the first step, the thermal-hydraulic model was used for studying the thermal-hydraulic instabilities. A study was then conducted of the one-channel nuclear-coupled thermal-hydraulic dynamics in a BWR by adding a simple point kinetic model for neutron kinetics and a model for the fuel heat conduction dynamics. In this paper, a two-channel nuclear-coupled thermal-hydraulic model is introduced to simulate the out-of phase oscillations in a BWR. This model comprises three parts: spatial mode neutron kinetics with the fundamental and fist azimuthal modes; fuel heat conduction dynamics; and thermal-hydraulics model. This present model is an extension of the Karve et al. model i.e., a drift flux model is used instead of the homogeneous equilibrium model for two-phase flow, and lambda modes are used instead of the omega modes for the neutron kinetics. This two-channel model is employed in stability and bifurcation analyses, carried out using the bifurcation code BIFDD. The stability boundary (SB) and the nature of the Poincare-Andronov-Hopf bifurcation (PAF-B) are determined and visualized in a suitable two-dimensional parameter/state space. A comparative study of the homogeneous equilibrium model (HEM) and the drift flux model (DFM) is carried out to investigate the effects of the DFM parameters the void distribution parameter C 0 and the drift velocity V gi -on the SB, the nature of PAH bifurcation, and on the type of oscillation mode (in-phase or out-of-phase). (author)

A BWR 24-month cycle analysis using multicycle techniques

International Nuclear Information System (INIS)

Hartley, K.D.

1993-01-01

Boiling water reactor (BWR) fuel cycle design analyses have become increasingly challenging in the past several years. As utilities continue to seek improved capacity factors, reduced power generation costs, and reduced outage costs, longer cycle lengths and fuel design optimization become important considerations. Accurate multicycle analysis techniques are necessary to determine the viability of fuel designs and cycle operating strategies to meet reactor operating requirements, e.g., meet thermal and reactivity margin constraints, while minimizing overall fuel cycle costs. Siemens Power Corporation (SPC), Nuclear Division, has successfully employed multi-cycle analysis techniques with realistic rodded cycle depletions to demonstrate equilibrium fuel cycle performance in 24-month cycles. Analyses have been performed by a BWR/5 reactor, at both rated and uprated power conditions

The World Nuclear University: New partnership in nuclear education

International Nuclear Information System (INIS)

2007-07-01

The important role which the IAEA plays in assisting Member States in the preservation and enhancement of nuclear knowledge and in facilitating international collaboration in this area has been recognized by the General Conference of the International Atomic Energy Agency in resolutions GC(46)/RES/11B, GC(47)/RES/10B, GC(48)/RES/13 and GC(50)/RES/13. A continued focus of IAEA activities in managing nuclear knowledge is to support Member States to secure and sustain human resources for the nuclear sector, comprising both the replacement of retiring staff and building of new capacity. The IAEA assists Member States, particularly developing ones, in their efforts to sustain nuclear education and training in all areas of nuclear technology for peaceful purposes, which is a necessary prerequisite for succession planning, in particular through the networking of nuclear education and training, including activities of the World Nuclear University (WNU) and the Asian Network for Education in Nuclear Technology (ANENT). The report on the attached CD-ROM, The World Nuclear University: New Partnership in Nuclear Education, gives an overview of the history of the development of the World Nuclear University and related IAEA activities and contains an analysis and recommendations from the first WNU Summer Institute, held in 2005 in the USA

BWR condensate filtration studies

International Nuclear Information System (INIS)

Wilson, J.A.; Pasricha, A.; Rekart, T.E.

1993-09-01

Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

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

Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide

International Nuclear Information System (INIS)

Gallardo V, J. M.; Morales S, J. B.

2013-10-01

The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO 2 ) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO 2 mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

BWR recirculation pump diagnostic expert system

International Nuclear Information System (INIS)

Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

2004-01-01

At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

Sophistication of operator training using BWR plant simulator

International Nuclear Information System (INIS)

Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

1986-01-01

In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)

Managing the aging of BWR control rod drive systems

International Nuclear Information System (INIS)

Greene, R.H.; Farmer, W.S.

1992-01-01

This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

Results of the Simulator smart against synthetic signals using a model of reduced order of BWR with additive and multiplicative noise; Resultados del simulador smart frente a senales sinteticas utilizando un modelo de orden reducido de BWR con ruido aditivo y multiplicativo

Energy Technology Data Exchange (ETDEWEB)

Munoz-Cobo, J. L.; Montesino, M. E.; Pena, J.; Escriva, A.; Melara, J.

2011-07-01

Results of SMART-simulator front of synthetic signals with models of reduced order of BWR with additive and multiplicative noise Under the SMART project, which aims to monitor the signals Cofrentes nuclear plant, we have developed a signal generator of synthetics BWR that will allow together real signals of plant the validation of the monitor.

BWR 9 X 9 Fuel Assembly Thermal-Hydraulic Tests (2): Hydraulic Vibration Test

International Nuclear Information System (INIS)

Yoshiaki Tsukuda; Katsuichiro Kamimura; Toshiitsu Hattori; Akira Tanabe; Noboru Saito; Masahiko Warashina; Yuji Nishino

2002-01-01

Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8 x 8 and 9 x 9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9 x 9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude. (authors)

The BWR Stability Issue

International Nuclear Information System (INIS)

D'Auria, F.

2008-01-01

The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

Parameter identification of a BWR nuclear power plant model for use in optimal control

International Nuclear Information System (INIS)

Volf, K.

1976-02-01

The problem being considered is the modeling of a nuclear power plant for the development of an optimal control system of the plant. Current system identification concepts, combining input/output information with a-priori structural information are employed. Two of the known parameter identification methods i.e., a least squares method and a maximum likelihood technique, are studied as ways of parameter identification from measurement data. A low order state variable stochastic model of a BWR nuclear power plant is presented as an application of this approach. The model consists of a deterministic and a noise part. The deterministic part is formed by simplified modeling of the major plant dynamic phenomena. The moise part models the effects of input random disturbances to the deterministic part and additive measurement noise. Most of the model parameters are assumed to be initially unknown. They are identified using measurement data records. A detailed high order digital computer simulation is used to simulate plant dynamic behaviour since it is not conceivable for experimentation of this kind to be performed on the real nuclear power plant. The identification task consists in adapting the performance of the simple model to the data acquired from this plant simulation ensuring the applicability of the techniques to measurement data acquired directly from the plant. (orig.) [de

Development and management of world nuclear power in 2012

International Nuclear Information System (INIS)

2012-01-01

It deals with development and management of nuclear power of foreign countries by the 1st of January 2012 with tables and figures, which includes outline of investigation, operation experience of nuclear power plant of the world, the cardinal number according to the type of operating power plant of the world, using Mox of the world and site of nuclear power plant of the world. There are list of world nuclear power plant, explanation of abbreviations, address book of nuclear power plant of the world and table and figure of major nuclear fuel cycle.

Seismic proving test of BWR primary loop recirculation system

International Nuclear Information System (INIS)

Sato, H.; Shigeta, M.; Karasawa, Y.

1987-01-01

The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

Metallurgical factors that contribute to cracking in BWR piping

International Nuclear Information System (INIS)

Weeks, J.R.

1975-01-01

During the fall of 1974 and early winter of 1975, cracks have been discovered in the 4 in. bypass lines of several Boiling Water Reactors (BWR's) in the United States. Further, similar cracks were discovered at two BWR's in Japan during the same period. More recently, cracks have been discovered in the core spray piping and in a furnace-sensitized ''safe end'' and adjacent ''dutchman'' at the Dresden Nuclear Power Station, Unit No. 2. Although inspections at all other U.S. BWR's have not disclosed further instances of cracking in core spray piping, leaking cracks have been found in the core spray piping of two BWR's overseas. Metallurgical examinations of these cracks are not yet complete. The following observations have been made to date. All cracks (except those in the furnace-sensitized safe end and dutchman) occurred in seamless type 304 stainless steel piping or in elbows fabricated from such piping, in the outer heat affected zone of either field or shop welds, in lines isolated from the main primary coolant flow during full power operation, except for the not yet examined cracks in the Monticello bypass lines. The cracks are exclusively intergranular, and occur in metal that has been lightly sensitized by the welding process, with only intermittent grain boundary carbides. They developed in the areas of peak axial residual stresses from welding rather than in the most heavily sensitized areas. No fatigue striations have been found on the fracture surfaces. The evidence received to date strongly indicates that these cracks were caused by intergranular stress corrosion of weld-sensitized stainless steel by BWR water containing greater than 0.2 ppM oxygen. The possible role of fatigue or alternating stresses in this corrosion is not clear. Further, not all the cracks detected to date necessarily have occurred by the same mechanism

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

1992-01-01

This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

European BWR R and D cluster for innovative passive safety systems

International Nuclear Information System (INIS)

Hicken, E.F.; Lensa, W. von

1996-01-01

The main technological innovation trends for future nuclear power plants tend towards a broader use of passive safety systems for the prevention, mitigation and managing of severe accident scenarios. Several approaches have been undertaken in a number of European countries to study and demonstrate the feasibility and charateristics of innovative passive safety systems. The European BWR R and D Cluster combines those experimental and analytical efforts that are mainly directed to the introduction of passive safety systems into boiling water reactor technology. The Cluster is grouped around thermohydraulic test facilities in Europe for the qualification of innovative BWR safety systems, also taking into account especially the operating experience of the nuclear power plant Dodewaard and other BWRs, which already incorporated some passive safety features. The background, the objectives, the structure of the project and the work programme are presented in this paper as well as an outline of the significance of the expected results. (orig.) [de

A World Nuclear University

International Nuclear Information System (INIS)

Yanev, Y.

2004-01-01

The paper discusses the mission and tasks of the World Nuclear University (WNU) established to build worldwide knowledge and support the effective use of nuclear techniques for solving the global human and environmental problems of 21 century and thereby support the global sustainable development. In this respect the WNU would build Human resources, Technical knowledge and Public Support. A Network of educational and research institutions with strong programmes in nuclear science and engineering will be created. The WNU Head quarters and Regional Centers will: 1) Facilitate agreement on curriculum and WNU certification curriculum 2) Develop and administer scholarships; 3) Foster educational exchanges within WNU family institutions; 4) Build core faculty for summer 1/2 year Masters degrees; 5) Co-ordinate research, grants and knowledge management research; 6) Operate think tank and public information service; 7) Emphasise key areas such as safeguards systems and the nuclear-renewable-hydrogen economy; 8) Oversee world-wide human resources pool; 9) Orchestrate alumni support for nuclear technology. The possible participants and possible location of the Regional Centres are given

The World Nuclear Industry Status Report 2013

International Nuclear Information System (INIS)

Schneider, Mycle; Froggatt, Antony; Hosokawa, Komei; Thomas, Steve; Yamaguchi, Yukio; Hazemann, Julie; Bradford, Peter A.

2013-07-01

Two years after the Fukushima disaster started unfolding on 11 March 2011, its impact on the global nuclear industry has become increasingly visible. Global electricity generation from nuclear plants dropped by a historic 7 percent in 2012, adding to the record drop of 4 percent in 2011. This World Nuclear Industry Status Report 2013 (WNISR) provides a global overview of the history, the current status and the trends of nuclear power programs worldwide. It looks at nuclear reactor units in operation and under construction. Annex 1 provides 40 pages of detailed country-by-country information. A specific chapter assesses the situation in potential newcomer countries. For the second time, the report looks at the credit-rating performance of some of the major nuclear companies and utilities. A more detailed chapter on the development patterns of renewable energies versus nuclear power is also included. Annex 6 provides an overview table with key data on the world nuclear industry by country. The 2013 edition of the World Nuclear Industry Status Report also includes an update on nuclear economics as well as an overview of the status, on-site and off-site, of the challenges triggered by the Fukushima disaster. However, this report's emphasis on recent post-Fukushima developments should not obscure an important fact: as previous editions (see www.WorldNuclearReport.org) detail, the world nuclear industry already faced daunting challenges long before Fukushima, just as the U.S. nuclear power industry had largely collapsed before the 1979 Three Mile Island accident. The nuclear promoters' invention that a global nuclear renaissance was flourishing until 3/11 is equally false: Fukushima only added to already grave problems, starting with poor economics. The performance of the nuclear industry over the year from July 2012 to July 2013 is summed up in this report

Is a nuclear weapon-free world desirable?

International Nuclear Information System (INIS)

Tertrais, Bruno

2009-01-01

In this article, the author shows that a nuclear weapon-free world would probably be more dangerous than today's world because benefits of the existence of nuclear weapons are probably more important that the risks related to their existence. He outlines that nuclear deterrence has been very efficient for these last 65 years. He states that the disappearance of nuclear weapons could be envisaged only after a large transformation of safety conditions, but that such transformations are actually not at all under way. It would indeed require peaceful and democratic world governance

SCORPIO-BWR: status and future plans

International Nuclear Information System (INIS)

Porsmyr, Jan; Bodal, Terje; Beere, William H.

2004-01-01

Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

SCORPIO-BWR: status and future plans

Energy Technology Data Exchange (ETDEWEB)

Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

2004-07-01

Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR

After the world court opinion: Towards a world without nuclear weapons

International Nuclear Information System (INIS)

Roche, D.

1998-01-01

With the World Court advisory opinion on the treat or use of nuclear weapons, a new moment has arrived in the 51-year history of atomic bomb. The highest legal body in the world has said that governments must not only pursue but conclude negotiations leading to total nuclear disarmament. Both a sharp focus and a comprehensive action are required in developing public support and political action towards a safer, more peaceful world. Three-pronged interrelated course of action is proposed: a Nuclear Weapon Convention; reduction in conventional arms and control of arms trade; and new spending priorities or sustainable development

Estimation of dose rate around the spent control rods of a BWR; Estimacion de la rapidez de dosis alrededor de las barras de control gastadas de un BWR

Energy Technology Data Exchange (ETDEWEB)

Cancino P, G.

2016-10-01

The energy can come from fossil renewable sources (solar (natural gas, oil), wind, hydro, tidal, geothermal, biomass, bio energy and nuclear. Nuclear power can be obtained by fission reactions and fusion (still under investigation) atomic nuclei. Fission, is a partition of a very heavy nucleus (Uranium 235, for example) into two lighter nuclei. Much of the world's electric power is generated from the energy released by fission processes. In a nuclear power reactor, light water as the BWR, there are many important elements that allow safe driving operation, one of them are the elements or control systems, the burnable poison or neutron absorber inherently allow control power reactor. The control rods, which consist mostly of stainless steel and absorbing elements (such as boron carbide, hafnium, cadmium, among others) of thermal neutrons is able to initiate, regulate or stop the reactor power. These, due to the use of depleted burned or absorbing material and therefore reach their lifespan, which can be 15 years or have other values depending on the manufacturer. Control rods worn should be removed, stored or confined in expressly places. Precisely at this stage arises the importance of knowing their radiological condition to manipulate safely and without incident to the people health responsible for conducting these proceedings state arises. This thesis consists in the estimation of the dose rate in spent control rod made of boron carbide, from a typical BWR reactor. It will be estimated by direct radiation measurements with measurement equipment for radiotherapy ionization chamber, in six spent control rods, which were taken at different reactor operating cycles and are in a spent fuel pool. Using bracket electromechanical and electronic equipment for positioning and lifting equipment for radiation measurement around the control rod in the axial and radial arrangement for proper scanning. Finally will be presented a graphic corresponding to the dose

High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-12-01

High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

Appraisal of BWR plutonium burners for energy centers

International Nuclear Information System (INIS)

Williamson, H.E.

1976-01-01

The design of BWR cores with plutonium loadings beyond the self-generation recycle (SGR) level is investigated with regard to their possible role as plutonium burners in a nuclear energy center. Alternative plutonium burner approaches are also examined including the substitution of thorium for uranium as fertile material in the BWR and the use of a high-temperature gas reactor (HTGR) as a plutonium burner. Effects on core design, fuel cycle facility requirements, economics, and actinide residues are considered. Differences in net fissile material consumption among the various plutonium-burning systems examined were small in comparison to uncertainties in HTGR, thorium cycle, and high plutonium-loaded LWR technology. Variation in the actinide content of high-level wastes is not likely to be a significant factor in determining the feasibility of alternate systems of plutonium utilization. It was found that after 10,000 years the toxicity of actinide high-level wastes from the plutonium-burning fuel cycles was less than would have existed if the processed natural ores had not been used for nuclear fuel. The implications of plutonium burning and possible future fuel cycle options on uranium resource conservation are examined in the framework of current ERDA estimates of minable uranium resources

Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

2004-07-01

The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

High Fidelity BWR Fuel Simulations

Energy Technology Data Exchange (ETDEWEB)

Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-08-01

This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

BWR core melt progression phenomena: Experimental analyses

International Nuclear Information System (INIS)

Ott, L.J.

1992-01-01

In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

Estimation of dose rate around the spent control rods of a BWR

International Nuclear Information System (INIS)

Cancino P, G.

2016-01-01

The energy can come from fossil renewable sources (solar (natural gas, oil), wind, hydro, tidal, geothermal, biomass, bio energy and nuclear. Nuclear power can be obtained by fission reactions and fusion (still under investigation) atomic nuclei. Fission, is a partition of a very heavy nucleus (Uranium 235, for example) into two lighter nuclei. Much of the world's electric power is generated from the energy released by fission processes. In a nuclear power reactor, light water as the BWR, there are many important elements that allow safe driving operation, one of them are the elements or control systems, the burnable poison or neutron absorber inherently allow control power reactor. The control rods, which consist mostly of stainless steel and absorbing elements (such as boron carbide, hafnium, cadmium, among others) of thermal neutrons is able to initiate, regulate or stop the reactor power. These, due to the use of depleted burned or absorbing material and therefore reach their lifespan, which can be 15 years or have other values depending on the manufacturer. Control rods worn should be removed, stored or confined in expressly places. Precisely at this stage arises the importance of knowing their radiological condition to manipulate safely and without incident to the people health responsible for conducting these proceedings state arises. This thesis consists in the estimation of the dose rate in spent control rod made of boron carbide, from a typical BWR reactor. It will be estimated by direct radiation measurements with measurement equipment for radiotherapy ionization chamber, in six spent control rods, which were taken at different reactor operating cycles and are in a spent fuel pool. Using bracket electromechanical and electronic equipment for positioning and lifting equipment for radiation measurement around the control rod in the axial and radial arrangement for proper scanning. Finally will be presented a graphic corresponding to the dose

BWR water chemistry impurity studies

International Nuclear Information System (INIS)

Ljungberg, L.G.; Korhonen, S.; Renstroem, K.; Hofling, C.G.; Rebensdorff, B.

1990-03-01

Laboratory studies were made on the effect of water impurities on environmental cracking in simulated BWR water of stainless steel, low alloy steel and nickel-base alloys. Constant elongation rate tensile (CERT) tests were run in simulated normal water chemistry (NWC), hydrogen water chemistry (HWC), or start-up environment. Sulfate, chloride and copper with chloride added to the water at levels of a fraction of a ppM were found to be extremely deleterious to all kinds of materials except Type 316 NG. Other detrimental impurities were fluoride, silica and some organic acids, although acetic acid was beneficial. Nitrate and carbon dioxide were fairly inoccuous. Corrosion fatigue and constant load tests on compact tension specimens were run in simulated normal BWR water chemistry (NWC) or hydrogen water chemistry (HWC), without impurities or with added sulfate or carbon dioxide. For sensitized Type 304 SS in NWC, 0.1 ppM sulfate increased crack propagation rates in constant load tests by up to a factor of 100, and in fatigue tests up to a factor of 10. Also, cracking in Type 316 nuclear grade SS and Alloy 600 was enhanced, but to a smaller degree. Carbon dioxide was less detrimental than sulfate. 3 figs., 4 tabs

International nuclear technology transfer

International Nuclear Information System (INIS)

Cartwright, P.; Rocchio, J.P.

1978-01-01

Light water reactors (LWRs), originally developed in the United States, became the nuclear workhorses for utilities in Europe and Japan largely because the U.S. industry was willing and able to transfer its nuclear know-how abroad. In this international effort, the industry had the encouragement and support of the U.S. governement. In the case of the boiling water reactor (BWR) the program for technology transfer was developed in response to overseas customer demands for support in building local designs and manufacturing capabilities. The principal vehicles have been technology exchange agreements through which complete engineering and manufacturing information is furnished covering BWR systems and fuel. Agreements are held with companies in Germany, Japan, Italy, and Sweden. In recent years, a comprehensive program of joint technology development with overseas manufacturers has begun. The rapidly escalating cost of nuclear research and development make it desirable to minimize duplication of effort. These joint programs provide a mechanism for two or more parties jointly to plan a development program, assign work tasks among themselves, and exchange test results. Despite a slower-than-hoped-for start, nuclear power today is playing a significant role in the economic growth of some developing countries, and can continue to do so. Roughly half of the 23 free world nations that have adopted LWRs are developing countries

World nuclear performance report 2017

International Nuclear Information System (INIS)

Cobb, Jonathan

2017-01-01

World Nuclear Association recently published the 2017 edition of the World Nuclear Performance Report. The report presents key metrics that illustrate current performance, both of reactors currently operating and those under construction. The article highlights some of the most important findings of the report. The pace of new build will need to accelerate if nuclear energy is going to make a growing contribution to the global electricity generation mix, a requirement of many projections of future scenarios that aim to meet the objective of limiting the rise average temperatures to below two degrees Celsius, while at the same time meeting the growing worldwide demand for electricity.

World nuclear performance report 2017

Energy Technology Data Exchange (ETDEWEB)

Cobb, Jonathan [World Nuclear Association, London (United Kingdom)

2017-08-15

World Nuclear Association recently published the 2017 edition of the World Nuclear Performance Report. The report presents key metrics that illustrate current performance, both of reactors currently operating and those under construction. The article highlights some of the most important findings of the report. The pace of new build will need to accelerate if nuclear energy is going to make a growing contribution to the global electricity generation mix, a requirement of many projections of future scenarios that aim to meet the objective of limiting the rise average temperatures to below two degrees Celsius, while at the same time meeting the growing worldwide demand for electricity.

BWR type reactor core

International Nuclear Information System (INIS)

Tatemichi, Shin-ichiro.

1981-01-01

Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

The World Nuclear Industry Status Report 2004

International Nuclear Information System (INIS)

Schneider, Mycle; Froggatt, Antony

2004-12-01

Fifty years ago, in September 1954, the head of the US Atomic Energy Commission stated that nuclear energy would become 'too cheap to meter': The cost to produce energy by nuclear power plants would be so low that the investment into electricity meters would not be justified. By coincidence the US prophecy came within three months of the announcement of the world's first nuclear power plant being connected to the grid in.. the then Soviet Union. In June 2004, the international nuclear industry celebrated the anniversary of the grid connection at the site of the world's first power reactor in Obninsk, Russia, under the original slogan '50 Years of Nuclear Power - The Next 50 Years'. This report aims to provide a solid basis for analysis into the prospects for the nuclear power industry. Twelve years ago, the Worldwatch Institute in Washington, WISE-Paris and Greenpeace International published the World Nuclear Industry Status Report 1992. In the current international atmosphere of revival of the nuclear revival debate - it has been a periodically recurring phenomenon for the past twenty years - two of the authors of the 1992 report, now independent consultants, have carried out an updated review of the status of the world nuclear industry. The performance of the nuclear industry over the past year is summed up in this report

Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

2014-10-15

In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

Characteristics of fluctuating pressure generated in BWR main steam lines

International Nuclear Information System (INIS)

Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

2009-01-01

The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

Phenomenology of BWR fuel assembly degradation

Science.gov (United States)

Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

2018-03-01

Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

Energy Technology Data Exchange (ETDEWEB)

Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

2006-07-01

In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

Flex concept for US-A BWR extended loss of AC power events

International Nuclear Information System (INIS)

Powers, J.; Aoyagi, Y.; Kataoka, K.; Thomas, S.; Mookhoek, B.

2015-09-01

The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

Flex concept for US-A BWR extended loss of AC power events

Energy Technology Data Exchange (ETDEWEB)

Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Aoyagi, Y.; Kataoka, K. [Toshiba Corporation, Kawasaki, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

2015-09-15

The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

Nuclear energy in the world

International Nuclear Information System (INIS)

Grippi, Sidney

2006-01-01

The chapter reports the nuclear energy beginning in the world including a chronology of the atomic bomb birth, the annual growth rate of electronuclear energy in the world, a comparison of energy production in thermoelectric bases

ECP measurements in the BWR-1 water loop relative to water composition changes

Energy Technology Data Exchange (ETDEWEB)

Kus, P.; Vsolak, R.; Kysela, J., E-mail: ksp@ujv.cz [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic); Hanawa, S.; Nakamura, T.; Uchida, S., E-mail: hanawa.satoshi@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

2010-07-01

The goal of this study is to investigate the usage of ECP sensors in nuclear power plants. ECP sensors were tested using the LVR-15 reactor at the Nuclear Research Institute Rez plc (NRI) in the Czech Republic. The experiment took place on the BWR-1 loop, which was designed for investigating the behaviour of structural materials and radioactivity transport under BWR conditions. The BWR-1 loop facilitates irradiation experiments within a wide range of operating parameters (max. pressure of 10 MPa, max. temperature of 573 K and a neutron flux of 1.0* 10{sup 18} n/m{sup 2}s). This study involves the measurement of electrochemical potential (ECP). Corrosion potential is the main parameter for monitoring of water composition changes in nuclear power plants (NPP). The electrochemical potentials of stainless steel were measured under high temperatures in a test loop (BWR-1) under different water composition conditions. Total neutron flux was ∼10{sup -3} to ∼10{sup 12} n/cm{sup 2}s (>0.1 MeV) at a temperature of 560K, neutral pH, and water resistivity of 18.2 MOhm. ECP sensor response related to changes in water composition was monitored. Switching from NWC (normal water conditions) to HWC (hydrogen water conditions) was controlled using oxygen dosage. Water chemistry was monitored approx. 50 meters from the active channel. The active channel temperature was maintained within a range of 543 - 561 K from the start of irradiation for the entire duration of the experiment. A total of 24 reference electrodes composed of platinum (Pt), silver/silver chloride (Ag/AgCl) and a zircon membrane containing silver oxide (Ag{sub 2}O) powder were installed inside the active channel of the LVR-15 test reactor. The active channel (Field tube) was divided into four zones, with each zone containing six sensors. A mathematical radiolysis code model was created in cooperation with the Japan Atomic Energy Agency. (author)

Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1998-07-01

The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

Nuclear news from the world

International Nuclear Information System (INIS)

Anon.

2010-01-01

The Monju breeder reactor (Japan) has diverged again after a shutdown period of 14 years. This prototype reactor was stopped in 1995 after a sodium leak from its cooling system had initiated a fire. The construction works of the third power reactor on Brazilian soil has begun on the Angra site. The Chinese advanced research reactor (CARR - 60 MWth) diverged for the first time on May 2010 the 13., its design nears that of the Australian Opale reactor. The first reactor of the Ling Ao Phase 2 has been commissioned, this reactor is the first Chinese model (CPR-1000) to be built, it is the fruit of the technological transfer agreements that were signed in 1992 and 1995 with Framatome. The Finn government has given its agreement for the extension of the storing capacity in deep geological layers of the Posiva Oy center, this capacity is now of 9000 tonnes of uranium. Russia and Namibia have signed an agreement for the exploration and exploitation of uranium mines. The 2 first BWR-type reactors (1350 MW) on American soil will be constructed in the state of Texas. Russia and Turkey have signed an agreement for the construction of the first nuclear power plant in Turkey, this plant will be composed of 4 reactors. A study shows that there is no supplementary health hazards in the surroundings of nuclear power plants in Spain. (A.C.)

The World Nuclear University: Addressing global needs. London, 4 September 2003. Inauguration ceremony, World Nuclear University

International Nuclear Information System (INIS)

ElBaradei, M.

2003-01-01

For some time, there has been a growing awareness of the need for succession planning in the nuclear industry, to ensure that we cultivate a new generation of young people with the proper education and skills to replace the aging nuclear workforce as its members retire. Today's inauguration of the 'World Nuclear University' (WNU) is the most substantive action taken to date to address this need. This is a challenge, because the widespread perception clearly exists that nuclear energy is a dying field. The IAEA, with its constituency of 135 Member States, is hopeful that this will truly become a World Nuclear University. Almost 2 billion people, nearly one third of the population of the planet, remain without access to modern energy supplies - a shortfall that could be addressed, at least in part, by nuclear energy. But any major expansion in the future use of nuclear power will only be feasible if the nuclear industry is successful in developing innovative reactor and fuel cycle technology - as well as operational and regulatory approaches - that effectively address concerns related to cost competitiveness, safety and security, proliferation resistance and waste disposal. And global development needs go well beyond the electricity sector. The IAEA's recognition of these situations underlies our assistance to Member States, through which we try to address areas of high national priority wherever nuclear technology provides the best option for success. A significant part of that effort lies in the development of human capacity - through training and education in how to apply nuclear technology safely and effectively. 'Atoms for Peace' is a vision nearly five decades old, focused on using nuclear science for the advancement of humankind. It is my hope that this 'World Nuclear University' can be an effective instrument towards the achievement of that vision

Emerging nuclear suppliers in the Third World

International Nuclear Information System (INIS)

Stahl, K.

1990-01-01

The emergence of new supplier states of nuclear technology within the Third World has raised concern, if those nuclear supplier states will promote an unrestricted and uncontrolled transfer of nuclear technology to developing countries and augment the risk of nuclear weapons proliferation. The article analyses the nuclear export capacities, nuclear exports and the export policy of Argentina, Brazil and India. Argentina is considered as the most important emerging nuclear supplier state in the Third World. Nuclear exports have to be authorisized by the government in all three states and will be covered by IAEA-safeguards in the recipient country. The three states will exercise restraint in the transfer of sensitive nuclear technology. Nuclear exports of Argentina, Brazil and India so far will not augment the danger of nuclear weapons proliferation. (orig./HSCH) [de

Nuclear security in a transformed world

International Nuclear Information System (INIS)

Gottfried, K.; Dean, J.

1991-01-01

In the wake of the failed coup attempt in the Soviet Union, the world stands hopeful that a new era of international peace and cooperation lies ahead. President Bush's unilateral reductions in tactical nuclear weapons and in the alert levels of US forces, coupled with President Gorbachev's largely reciprocal actions, are important steps toward realizing that hope. While bold in the context of recent arms control history, however, these actions are modest in the face of the current enormous opportunity and the shifting threats the world now faces. Even with these welcome unilateral actions, the likely continued presence of thousands of nuclear weapons throughout many of the Soviet republics, the temporary uncertainty over central government command during the coup, and fuller knowledge of Iraq's aggressive efforts to build a nuclear bomb serve as stark reminders that the danger of nuclear catastrophe has not disappeared. Although a deliberate attack by the Soviet Union against the US or Europe is now almost inconceivable, nuclear weapons continue to pose significant threats to US security and world peace. These threats fall into three broad categories: a persistent risk of regional nuclear war involving countries other than the Soviet Union that are already in possession of nuclear weapons or capable of building them; the spread of nuclear weapons to other countries; accidental or unauthorized use. To meet this new challenge, three key steps must be taken: reduce dramatically Soviet and US nuclear arsenals; negotiate restrictions on the arsenals of other nuclear powers; strengthen the nuclear nonproliferation regime

BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-04-30

The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 10⁷ cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m^-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

Nuclear energy in Europe and the world

International Nuclear Information System (INIS)

Koenig, H.H.; Brown, Boveri und Cie A.G., Mannheim

1982-01-01

The author provides an account of opinions expressed at the 1982 Euratom Congress on the world's economical situation, public views on nuclear energy, the energy problem of the third world an on the development status of nuclear technology. (orig.) [de

Current status of light water reactor and Hitachi's technical improvements for BWR

International Nuclear Information System (INIS)

Miki, Minoru; Ohki, Arahiko.

1984-01-01

Gradual technical improvements in Japan over the years has improved the reliability of light water reactors, and has achieved the highest capacity factor level in the world. Commercial operation of Fukushima 2-2 (1,100 MW) of the Tokyo Electric Power Co. was started in February, 1984, as the first standardized BWR base plant, ushering in a new age of domestic light water reactor technology. The ABWR (1,300 MW class) has been developed as Japan's next generation light water reactor, with construction aimed at the latter half of the 1980's. Hitachi's extensive efforts range from key nuclear equipment to various related robots, directed at improving safety, reliability, and the capacity factor, while reducing radiation exposure. This paper presents an outline of Hitachi's participation in the light water reactor's improvement and standardization, and the current status of our role in the international cooperation plan for the ABWR. (author)

Nuclear energy in a nuclear weapon free world

Energy Technology Data Exchange (ETDEWEB)

Pilat, Joseph [Los Alamos National Laboratory

2009-01-01

The prospect of a nuclear renaissance has revived a decades old debate over the proliferation and terrorism risks of the use of nuclear power. This debate in the last few years has taken on an added dimension with renewed attention to disarmament. Increasingly, concerns that proliferation risks may reduce the prospects for realizing the vision of a nuclear-weapon-free world are being voiced.

A BWR licensing experience in the USA

International Nuclear Information System (INIS)

Powers, J.; Ogura, C.; Arai, K.; Thomas, S.; Mookhoek, B.

2015-09-01

The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

A BWR licensing experience in the USA

Energy Technology Data Exchange (ETDEWEB)

Powers, J.; Ogura, C. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Arai, K. [Toshiba Corporation, Yokohama, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

2015-09-15

The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

World Integrated Nuclear Evaluation System: Model documentation

International Nuclear Information System (INIS)

1991-12-01

The World Integrated Nuclear Evaluation System (WINES) is an aggregate demand-based partial equilibrium model used by the Energy Information Administration (EIA) to project long-term domestic and international nuclear energy requirements. WINES follows a top-down approach in which economic growth rates, delivered energy demand growth rates, and electricity demand are projected successively to ultimately forecast total nuclear generation and nuclear capacity. WINES could be potentially used to produce forecasts for any country or region in the world. Presently, WINES is being used to generate long-term forecasts for the United States, and for all countries with commercial nuclear programs in the world, excluding countries located in centrally planned economic areas. Projections for the United States are developed for the period from 2010 through 2030, and for other countries for the period starting in 2000 or 2005 (depending on the country) through 2010. EIA uses a pipeline approach to project nuclear capacity for the period between 1990 and the starting year for which the WINES model is used. This approach involves a detailed accounting of existing nuclear generating units and units under construction, their capacities, their actual or estimated time of completion, and the estimated date of retirements. Further detail on this approach can be found in Appendix B of Commercial Nuclear Power 1991: Prospects for the United States and the World

BWR containments license renewal industry report; revision 1. Final report

International Nuclear Information System (INIS)

Smith, S.; Gregor, F.

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 (IR's). License renewal applicants may choose to reference these IR's in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (IOCFR54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) containments. The scope of the report includes containments constructed of reinforced or prestressed concrete with steel liners and freestanding stell containments. Those domestic BWR containments designated as Mark I, Mark II or Mark III are covered, but no containments are addressed before these designs. The report includes those items within the jurisdictional boundaries for metal and concrete containments defined by Section III of the ASME Boiler and Pressure Vessel Code, Division 1, Subsection NE (Class MC) and Division 2 (Class CC) and their supports, but excluding snubbers

The World Nuclear Industry Status Report: 1992

International Nuclear Information System (INIS)

Flavin, Christopher; Lenssen, Nicholas; Froggatt, Antony; Willis, John; Kondakji, Assad; Schneider, Mycle

1992-05-01

The World Nuclear Industry Status Report provides a comprehensive overview of nuclear power plant data, including information on operation, production and construction. The WNISR assesses the status of new-build programs in current nuclear countries as well as in potential newcomer countries. This first WNISR Report was issued in 1992 in a joint publication with WISE-Paris, Greenpeace International and the World Watch Institute, Washington

World nuclear developments after Chernobyl

International Nuclear Information System (INIS)

Rippon, S.

1987-01-01

1986 will inevitably go down in history as the year of Chernobyl, the consequences of which must be delays in and even withdrawals from the development of nuclear power. On the credit side, the Soviet Union has done a rapid and remarkable job in sealing the damaged reactor and rehabilitating the station and the area while improving the safety of its total program. Equally effective has been the response of the IAEA. In terms of nuclear power's claim as a major source of energy, nothing has changed as a result of Chernobyl. 15% of the world's electricity is now produced from nearly 400 power reactors. In comparison with any other energy form nuclear energy must rank high in terms of economy, safety and environmental effects. What has changed is the public perception of nuclear power, and the effort world-wide which will need to be made to restore public confidence

Recent SCDAP/RELAP5 improvements for BWR severe accident simulations

International Nuclear Information System (INIS)

Griffin, F.P.

1995-01-01

A new model for the SCDAP/RELAP5 severe accident analysis code that represents the control blade and channel box structures in a boiling water reactor (BWR) has been under development since 1991. This model accounts for oxidation, melting, and relocation of these structures, including the effects of material interactions between B 4 C, stainless steel, and Zircaloy. This paper describes improvements that have been made to the BWR control blade/channel box model during 1994 and 1995. These improvements include new capabilities that represent the relocation of molten material in a more realistic manner and modifications that improve the usability of the code by reducing the frequency of code failures. This paper also describes a SCDAP/RELAP5 assessment calculation for the Browns Ferry Nuclear Plant design based upon a short-term station blackout accident sequence

Water chemistry control and decontamination experience with TEPCO BWR`s and the measures planned for the future

Energy Technology Data Exchange (ETDEWEB)

Suzuki, N.; Miyamaru, K. [Tokyo Electric Power Co. (Japan)

1995-03-01

The new TEPCO BWR`s are capable of having the occupational radiation exposure controlled successfully at a low level by selecting low cobalt steel, using corrosion-resistant steel, employing dual condensate polishing systems, and controlling Ni/Fe ratio during operation. The occupational radiation exposure of the old BWR`s, on the other hand, remains high though reduced substantially through the use of low cobalt replacement steel and the partial addition of a filter in the condensate polishing system. Currently under review is the overall decontamination procedure for the old BWR`s to find out to measures needed to reduce the amount of crud that is and has been carried over into the nuclear reactor. The current status of decontamination is reported below.

System control model of a turbine for a BWR; Modelo del sistema de control de una turbina para un BWR

Energy Technology Data Exchange (ETDEWEB)

Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rodolfo.amador@inin.gob.mx

2009-10-15

In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

BWR Radiation Assessment and Control Program: assessment and control of BWR radiation fields. Volume 1. Executive summary

International Nuclear Information System (INIS)

Anstine, L.D.

1983-05-01

This report covers work on the BWR Radiation Assessment and Control (BRAC) Program from 1978 to 1982. The major activities during this report period were assessment of the radiation-level trends in BWRs, evaluation of the effects of forward-pumped heater drains on BWR water quality, installation and operation of a corrosion-product deposition loop in an operating BWR, and analyzation of fuel-deposit samples from two BWRs. Radiation fields were found to be controlled by cobalt-60 and to vary from as low as 50 mr/hr to as high as 800 mr/hr on the recirculation-system piping. Detailed information on BWR corrosion films and system deposits is presented in the report. Additionally, the results of an oxygen-injection experiment and recontamination monitoring studies are provided

Development of a computerized operator support system for BWR power plant

International Nuclear Information System (INIS)

Monta, K.; Sekimizu, K.; Sato, N.; Araki, T.; Mori, N.

1985-01-01

A computerized operator support system for BWR power plant has been developed since 1980 supported by the Japanese government. The main functions of the systems are post trip operational guidance, disturbance analysis, standby system management, operational margin monitoring and control rod operational guidance. The former two functions aim at protection against incidents during operation of nuclear power plants and the latter three functions aim at their prevention. As the final stage of the development, these functions are combined with the plant supervision function and are organized as an advanced man-machine interface for BWR power plant. During the above process, operator task analyses are performed to enable synthesis of these support functions for right fit to operator tasks and to realize a hierarchical structure for CRT displays for right fit to operators cognitive needs. (author)

Directory of nuclear power plants in the world, 1985

International Nuclear Information System (INIS)

Fujii, Haruo

1985-01-01

This book presents technical information and estimates trends of load factors and construction costs of nuclear power plants. Particularly road maps indicating plants are drawn in, which would be practical in visiting them. The data used here are directly confirmed by operators in every part of the world. Therefore, they reflect up-to-date nuclear power developments and its future. This allows wide and exact understanding of world's nuclear power. Chapter 1 presents nuclear power growth around the world and estimates forecasts based on information from electric power companies: nuclear power growths and the growths in the number of reactors around the world, in WOCA (World outside the Centrally Planned Economies Area), in CPEA (Centrally Planned Economies Area) are analyzed in detail. Chapter 2 presents nuclear power plants on maps by country. The maps show exact locations of nuclear power plants with local cities around them, rivers and lakes. For convenience, symbols are given to aid in identifying the types of reactors. Chapter 3 presents general information of nuclear power plants. Also the addresses of operators, all segments of nuclear power supply industries and nuclear organizations are included. For convenience, the index of nuclear power plants is added. Chapter 4 presents technical information, road maps in large scales and photographs of nuclear power plants in the world. The road maps show exact locations of plants. Chapter 5 presents operating experiences, load factors, refuelling and maintenance outages. The trends of data are analyzed both regionally (WOCA, CPEA) and world-widely. Chapter 6 presents trends of construction costs, component costs as percent of total construction costs and direct costs, and construction durations. (J.P.N.)

Behavior of small-sized BWR fuel under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

1992-01-01

The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

The World Nuclear Industry Status Report 2012

International Nuclear Information System (INIS)

Schneider, Mycle; Froggatt, Antony; Hazemann, Julie

2012-07-01

Twenty years after its first edition, World Nuclear Industry Status Report 2012 portrays an industry suffering from the cumulative impacts of the world economic crisis, the Fukushima disaster, ferocious competitors and its own planning and management difficulties. The report provides a global overview of the history, the current status and trends of nuclear power programs in the world. It looks at units in operation and under construction. Annex 1 also provides detailed country-by-country information. A specific chapter assesses the situation in potential newcomer countries. For the first time, the report looks at the credit-rating performance of some of the major nuclear companies and utilities. A more detailed chapter on the development patterns of renewable energies versus nuclear power is also included. The performance of the nuclear industry over the 18 months since the beginning of 2011 is summed up in this report

Facility of BWR type nuclear power plant

International Nuclear Information System (INIS)

Kubo, Mitsuji

1998-01-01

A condensate filtering device for cleaning condensate flown from a low pressure turbine and a condensate desalting device are connected by way of a condensate pipeline. Control rod drives (CRD) are disposed to the lower portion of BWR. A CRD pump and one end of a CRD feedwater pipeline are connected in series to the upstream of CRD. The other end of the CRD feedwater pipeline is connected to a CRD water taking pipeline branched from the condensate pipeline. Water is taken to the CRD from downstream of the condensate filtering device and upstream of a connecting portion between a low pressure heater drain pipeline and the condensate pipeline. Flow of impurities leached out of the condensate desalting device to the reactor can be suppressed, and rising of temperature of CRD water by the low pressure heater drain water is prevented. In addition, flowing of dissolved oxygen to the CRD system can be suppressed. (I.N.)

Exxon Nuclear WREM-based NJP-BWR ECCS evaluation model and example application to the Oyster Creek Plant

International Nuclear Information System (INIS)

Krysinski, T.L.; Bjornard, T.A.; Steves, L.H.

1975-01-01

A proposed integrated ECCS model for non-jet pump boiling water reactors is presented, using the RELAP4-EM/BLOWDOWN and RELAP4-EM/SMALL BREAK portions of the Exxon Nuclear WREM-based Generic PWR Evaluation Model coupled with the ENC NJP-BWR Fuel Heatup Model. The results of the application of the proposed model to Oyster Creek are summarized. The results of the break size sensitivity study using the proposed model for the Oyster Creek Plant are presented. The application of the above results yielded the MAPLHGR curves. Included are a description of the proposed non-jet pump boiling water reaction evaluation model, justification of its conformance with TOCFR50, Appendix K, the adopted Oyster Creek plant model, and results of the analysis and sensitivity studies. (auth)

State-of-the-art of world nuclear power

International Nuclear Information System (INIS)

Margulova, T.Kh.

1987-01-01

World-wide development of nuclear power is reviewed in short. It noted, that by the 1970 the overall capacity of world nuclear power plants have been reached 24 GW and the cost of nuclear power became equal the cost of power generated at coal-fired stations. By the end of 1985 the LWR-type reactors generated 87 per sent of overale nuclear capacity. Especially considerable developmet of nuclear power have been achieved in France, where 50 per sent of power consumption is provided with nuclear power

Development status of compact containment BWR

International Nuclear Information System (INIS)

Heki, H.; Nakamaru, M.; Mori, H.; Sekiguchi, K.; Kuroki, M.; Arai, K.; Hida, T.

2005-01-01

In Japan, increase of nuclear plant unit capacity has been promoted to take advantage of economies of scale while further enhancing safety and reliability. As a result, more than 50 units of nuclear power plants are playing important role in electric power generation. However, the factors, such as stagnant growth in the recent electricity demand, limitation in electricity grid capacity and limited in initial investment avoiding risk, will not be in favor of large plant outputs. The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response

BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

1985-10-01

Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor

International Nuclear Information System (INIS)

Ramirez S, J.R.; Alonso V, G.; Palacios H, J.

2004-01-01

The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

Best-estimate analysis development for BWR systems

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

1986-01-01

The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

Connected analysis nuclear-thermo-hydraulic of parallel channels of a BWR reactor using distributed computation; Analisis acoplado nuclear-termohidraulico de canales paralelos de un reactor BWR empleando computacion distribuida

Energy Technology Data Exchange (ETDEWEB)

Campos Gonzalez, Rina Margarita

2007-07-15

This work consists of the integration of three models previously developed which are described widely in Literature: model of the thermo-hydraulic channel, model of the modal neutronic and the model of the recirculation bows. The tool used for this connection of models is the PVM system, Parallel Virtual Machine that allowed paralleling the model by means of the concept of distributed computation. The purpose of making this connection of models is the one of obtaining a more complete tool than better represents the real configuration and the phenomenology of the nucleus of a BWR reactor, thus obtaining better results. In addition to maintaining the flexibility to improve the resulting model at any time, since the very complex or sophisticated models are difficult to improve being impossible to modify the equations they use and can include variables that are not of primary importance in the tackled problem or that mask relations among variables due to the excess of results. Also maintaining the flexibility for adding component of models or systems of the BWR reactor, all of this following the modeling needs. The Swedish Ringhals power plant was chosen to characterize the resulting connected model for counting on a Stability Benchmark that offers the opportunity to count on real plant data. Besides that in case 9 of cycle 14 of this Benchamark oscillations outside phase appeared, which are from great interest because the detection systems that register the average of the power of the nucleus do not detect them. Additionally in this work the model of the recirculation bows as an independent module is obtained in an individual way, since this model belongs to another work and works connected to the reactor vessel. The model of the recirculation bows is able to model several transients of interest, as it is shown in the Appendix A of this work, among which are found the tripping of recirculation pumps or the transference at low or high velocity of them. The scope of the

A nuclear-weapon-free world and true disarmament

International Nuclear Information System (INIS)

Salvini, G.

1999-01-01

This preliminary note about is important to consider when discussing hopes of achieving a nuclear-weapon-free world. Without a serious effort to establish intelligent, powerful bodies to control and judge the behavior of the nations on Earth, whatever they future weapons may be, the objective to reach a nuclear-weapon-free world may even succeed, but it is not enough to stop wars and death. Even more than that: if taken alone, as the 'Great Way', it could prove negative, for it could slow down the general effort to achieve peace on out planet. A nuclear-weapon-free world is of course a very good idea but two points must be discussed: how to achieve the nuclear-weapon-free world; and what will happen afterwards. Some considerations on the second point are made

The role of nuclear power in the world

International Nuclear Information System (INIS)

Goodman, E.I.

1977-01-01

The role of nuclear energy in the world is discussed from the near term and long term. For the period through the mid 1980s sufficient nuclear capacity is considered critical to forestall serious shortages of oil and possible high prices leading to economic stagnation. Over the next 30-35 years it is estimated that world nuclear power will reach a capacity of approx. 3 million megawatts electrical when world electrical capacity will be about 8 million megawatts. With this nuclear capacity and if a annual growth rate of 5% is achieved for coal, oil and gas would remain at their present rate of consumption and would be increasingly reserved for specialized uses where substitution is not feasible. Caution is stressed, however, especially in using long term forecasts except for overall guidance and even in short term projections frequent up-dating and revision is recommended. The factors which have inhibited nuclear power growth are discussed including: 1) rapidly rising capital costs and financing problems, 2) rising and uncertain fuel cycle costs, 3) uncertainties in licensing and public acceptance. Despite the foregoing, nuclear power still retains an economic edge over fossil-fired units in substantial portions of the world. Assuming satisfactory solution of its major problems it is estimated that about 27-40% of the electrical capacity of developing countries will be nuclear by the year 2000. This nuclear capacity will comprise approx. 20% of the world's total nuclear power capacity around the turn of the century. (orig.) [de

Paths to a nuclear world with reliable safeguards

International Nuclear Information System (INIS)

Zebroski, E.L.

1978-01-01

The effectiveness of safeguards in the nuclear industry in reducing proliferation is surveyed. Several basic topics relative to proliferation which are discussed are: (1) ''the recognition that 'proliferation' encompasses at least four different issues which may require distinct approaches and policies;'' (2) ''in the context of the regulatory process by which the perceived risks to society are managed;'' (3) ''in the context of the realistic options and objectives for an attainable world nuclear structure;'' and (4) ''in the perception of the public and of decision-makers of the attainable reduction in risks - and at what costs - and the recognition of the extent to which some costs have already been accrued.'' Options open to the world are: (1) a structured nuclear world, (2) an unstructured nuclear world, or (3) a benign energy world. Current US policy of denial of nuclear energy by indefinite delay is seen as indirectly pointing the US toward option 1 or 3, as the basic cause of American economic decline, and as a root cause of many international tensions resulting from the US decline. Certain alternate approaches to a breeder-type reactor program or to fuel reprocessing which should contribute to a more proliferation-resistant nuclear program are briefly discussed

Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors

International Nuclear Information System (INIS)

Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M.; Reyes F, M. del C.; Del Valle G, E.

2014-10-01

In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

Application of EASY5 and MMS modules to BWR controller design

International Nuclear Information System (INIS)

Carmichael, L.A.; Rayes, L.; Yasutake, T.

1987-01-01

The application of EPRI's MMS Library and BCS' EASY5 simulation language to the design of a digital feedwater control system for the Monticello Boiling Water Nuclear Power Plant is discussed. In order to first design and then verify the digital feedwater controller algorithms, a digital simulation model of the Monticello plant was constructed using a combination of custom designed modules, existing MMS two-phase library modules, and standard modules available in the EASY5 library. Details of the process models, namely the BWR nuclear steam supply system, the steamline piping, and the feedwater piping are described in a companion paper. Details of the models for the existing BWR turbine pressure inlet pressure control and recirculation flow control system are described. These models are required to be operational during the transient analysis portion of the feedwater controller design verification, since they interact strongly with the reactor steam flow and water level. The design of the digital feedwater flow control loop is described. Its design is of particular interest because it requires consideration of control loop interaction and is, therefore, a simple example of multivariable non-interacting control design

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

Performances of nuclear installations in the world

International Nuclear Information System (INIS)

Pate, Z.T.

1999-01-01

During the last years the operators of nuclear power plants in the world, have realized numerous improvements. This success is imputable to several factors, especially an important data exchange. The Chernobyl accident, in 1986, provoked the creation of the World Association of Nuclear Operators (W.A.N.O.). It allowed to exchange information and to develop cooperation in order to go beyond cultural barriers, linguistics and policies. Then, operators in the world have brought important improvements in matter of safety, reliability. (N.C.)

BWR - Spent Fuel Transport and Storage with the TNTM9/4 and TNTM24BH Casks

International Nuclear Information System (INIS)

Wattez, L.; Marguerat, Y.; Hoesli, C.

2006-01-01

The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., the Muehleberg Nuclear Power Plant owner, is involved in this process and has elected to store its BWR spent fuel in a new high capacity dual-purpose cask, the TNeTeM24BH from the COGEMA Logistics/TRANSNUCLEAR TN TM 24 family. The Muehleberg BWR spent fuels are transported by road in a medium size shuttle transport cask and then transferred to a heavy transport/storage cask (dry transfer) in the hot cell of ZWILAG site. For that purpose, COGEMA Logistics designed and supplied: - Two shuttle casks, TN TM 9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TN TM 9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TN TM 24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TN TM 9/4 round trips are performed, and one TN TM 24BH is loaded. 5 additional TN TM 24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TN TM 24BH high capacity dual purpose cask and the TN TM 9/4 transport cask characteristics and capabilities will then be detailed. (authors)

World nuclear outlook 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-29

As part of the EIA program to provide energy information, this analysis report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the uranium market. Long-term projections of US nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed for the Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM). In turn, the OCRWM provides partial funding for preparation of this report. The projections of uranium requirements are provided to the Organization for Economic Cooperation and Development (OECD) for preparation of the Nuclear Energy Agency/OECD report, Summary of Nuclear Power and Fuel Cycle Data in OECD Member Countries.

World nuclear outlook 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-01

As part of the EIA program to provide energy information, this analysis report presents the current status and projections through 2010 of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the uranium market. Long-term projections of US nuclear capacity, generation, and spent fuel discharges for three different scenarios through 2040 are developed for the Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM). In turn, the OCRWM provides partial funding for preparation of this report. The projections of uranium requirements are provided to the Organization for Economic Cooperation and Development (OECD) for preparation of the Nuclear Energy Agency/OECD report, Summary of Nuclear Power and Fuel Cycle Data in OECD Member Countries.

World nuclear outlook 1994

International Nuclear Information System (INIS)

1994-12-01

As part of the EIA program to provide energy information, this analysis report presents the current status and projections through 2010 of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the uranium market. Long-term projections of US nuclear capacity, generation, and spent fuel discharges for three different scenarios through 2040 are developed for the Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM). In turn, the OCRWM provides partial funding for preparation of this report. The projections of uranium requirements are provided to the Organization for Economic Cooperation and Development (OECD) for preparation of the Nuclear Energy Agency/OECD report, Summary of Nuclear Power and Fuel Cycle Data in OECD Member Countries

World nuclear outlook 1995

International Nuclear Information System (INIS)

1995-01-01

As part of the EIA program to provide energy information, this analysis report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the uranium market. Long-term projections of US nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed for the Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM). In turn, the OCRWM provides partial funding for preparation of this report. The projections of uranium requirements are provided to the Organization for Economic Cooperation and Development (OECD) for preparation of the Nuclear Energy Agency/OECD report, Summary of Nuclear Power and Fuel Cycle Data in OECD Member Countries

BWR Servicing and Refueling Improvement Program: Phase I summary report

Energy Technology Data Exchange (ETDEWEB)

Perry, D.R.

1978-09-01

Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.

BWR Servicing and Refueling Improvement Program: Phase I summary report

International Nuclear Information System (INIS)

Perry, D.R.

1978-09-01

Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development

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

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

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

Nuclear power in the developing world

International Nuclear Information System (INIS)

Poneman, D.

1982-01-01

This book explores the increasingly urgent issue of nuclear power policies in developing countries. It examines the motives which drive nuclear policies in the developing world and explores how security and economic objectives, domestic politics, and foreign influence shape nuclear policies, enriching the analysis with examples from South American, African and Asian experiences. (author)

World interest in nuclear desalination

International Nuclear Information System (INIS)

1969-01-01

Nuclear power will be used in a desalination plant for the first time in a USSR plant now nearing completion. Studies are in progress to expand the concept of linking the power to chemical industries. These and other developing ideas were subjects of keen discussion by world experts at an Agency conference on nuclear desalination in Madrid. (author)

Standard Technical Specifications, General Electric plants, BWR/4

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

Standard Technical Specifications, General Electric plants, BWR/6

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. This document Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

Standard Technical Specifications, General Electric Plants, BWR/6

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1, contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

OECD/NRC BWR Turbine Trip Transient Benchmark as a Basis for Comprehensive Qualification and Studying Best-Estimate Coupled Codes

International Nuclear Information System (INIS)

Ivanov, Kostadin; Olson, Andy; Sartori, Enrico

2004-01-01

An Organisation for Economic Co-operation and Development (OECD)/U.S. Nuclear Regulatory Commission (NRC)-sponsored coupled-code benchmark has been initiated for a boiling water reactor (BWR) turbine trip (TT) transient. Turbine trip transients in a BWR are pressurization events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. In addition, the available real plant experimental data make this benchmark problem very valuable. Over the course of defining and coordinating the BWR TT benchmark, a systematic approach has been established to validate best-estimate coupled codes. This approach employs a multilevel methodology that not only allows for a consistent and comprehensive validation process but also contributes to the study of different numerical and computational aspects of coupled best-estimate simulations. This paper provides an overview of the OECD/NRC BWR TT benchmark activities with emphasis on the discussion of the numerical and computational aspects of the benchmark

Validations of BWR nuclear design code using ABWR MOX numerical benchmark problems

International Nuclear Information System (INIS)

Takano, Shou; Sasagawa, Masaru; Yamana, Teppei; Ikehara, Tadashi; Yanagisawa, Naoki

2017-01-01

BWR core design code package (the HINES assembly code and the PANACH core simulator), being used for full MOX-ABWR core design, has been benchmarked against the high-fidelity numerical solutions as references, for the purpose of validating its capability of predicting the BWR core design parameters systematically from UO 2 to 100% MOX cores. The reference solutions were created by whole core critical calculations using MCNPs with the precisely modeled ABWR cores both in hot and cold conditions at BOC and EOC of the equilibrium cycle. A Doppler-Broadening Rejection Correction (DCRB) implemented MCNP5-1.4 with ENDF/B-VII.0 was mainly used to evaluate the core design parameters, except for effective delayed neutron fraction (β eff ) and prompt neutron lifetime (l) with MCNP6.1. The discrepancies in the results between the design codes HINES-PANACH and MCNPs for the core design parameters such as the bundle powers, hot pin powers, control rod worth, boron worth, void reactivity, Doppler reactivity, β eff and l, are almost within target accuracy, leading to the conclusion that HINES-PANACH has sufficient fidelity for application to full MOX-ABWR core design. (author)

Applied methods for mitigation of damage by stress corrosion in BWR type reactors

International Nuclear Information System (INIS)

Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C.

1998-01-01

The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

Identification and assessment of BWR in-vessel severe accident mitigation strategies

Energy Technology Data Exchange (ETDEWEB)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M.

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

Impact of advanced BWR core physics method on BWR core monitoring

Energy Technology Data Exchange (ETDEWEB)

Moon, H; Wells, A [Siemens Power Corporation, Richland (United States)

2000-07-01

Siemens Power Corporation recently initiated development of POWERPLEX{sup TM}-III for delivery to the Grand Gulf Nuclear Power Station. The main change introduced in POWERPLEX{sup TM}-III as compared to its predecessor POWERPLEX{sup TM}-II is the incorporation of the advances BWR core simulator MICROBURN-B2. A number of issues were identified and evaluated relating to the implementation of MICROBURN-B2 and its impact on core monitoring. MICROBURN-B2 demands about three to five times more memory and two to three times more computing time than its predecessor MICROBURN-B in POWERPLEX {sup TM}-II. POWERPLEX{sup TM}-III will improve thermal margin prediction accuracy and provide more accurate plant operating conditions to operators than POWERPLEX{sup TM}-II due to its improved accuracy in predicted TIP values and critical k-effective. The most significant advantage of POWERPLEX{sup TM}-III is its capability to monitor a relaxed rod sequence exchange operation. (authors)

Logical model for the control of a BWR turbine;Modelo logico para el control de una turbina de un BWR

Energy Technology Data Exchange (ETDEWEB)

Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R., E-mail: yonaeton@hotmail.co [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2009-07-01

In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

World nuclear fuel market. Eighteenth annual meeting

International Nuclear Information System (INIS)

Anon.

1991-01-01

The papers presented at the eighteenth World Nuclear Fuels Market meeting are cataloged separately. This volume includes information on the following areas of interest: world uranium enrichment capacity and enriched uranium inventories; the impact of new enrichment technologies; predictions of future market trends; non-proliferation aspects of nuclear trade; and a debate as to whether uranium can be successfully traded on a commodities exchange

The Nuclear Non-Proliferation Treaty: Regulating Nuclear Weapons around the World

Science.gov (United States)

Middleton, Tiffany Willey

2010-01-01

In May 2010, scientists, national security experts, and state delegates from nations around the world will convene in New York for the 2010 Nuclear Non-Proliferation Treaty Review Conference. They will review current guidelines for nuclear testing and possession of nuclear weapons in accordance with the Nuclear Non-Proliferation Treaty of 1968,…

A practical methodology of radiological protection for the reduction of hot particles in BWR type reactors; Una metodologia practica de proteccion radiologica para la reduccion de particulas calientes en reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Alvarez G, G [Comision Federal de Electricidad, Gerencia del Proyecto Nucleoelectrico Laguna Verde, Disciplina de Fisica Aplicada (Mexico)

1991-07-01

The purpose of this work, in general form, is to describe a practical method for reduction of hot particles generated as consequence of the operational activities of BWR nuclear reactors. This methodology provides a description of the localizations and/or probable activities of finding particles highly radioactive denominated hot particles. For this purpose it was developed a strategy based on the decontamination lineaments, as well as the manipulation, gathering, registration, contention, documentation, control and final disposition of the hot particles. In addition, some recommendations are reiterated and alternative, in order to gathering the hot particles in a dynamic way given to the activities of the personal occupationally exposed in highly radioactive areas. The structure of the methodology of hot particles is supported in the radiological controls based on the Code of Federal Regulation 10 CFR 20 as well as the applicable regulatory documents. It provides an idea based on administrative controls of radiological protection, in order to suggesting the responsibilities and necessary directing for the control of the hot particles required in nuclear plants of the BWR type. (author)

Nuclear energy and the developing world

International Nuclear Information System (INIS)

Mustafa, A.

1982-01-01

The importance of cooperation between the developed and developing countries with regard to nuclear power is discussed. Moves towards global interdependence were strengthened when OAPEC was set up with proposals for cooperation and depletion of world reserves of gas and oil will encourage this. Developing countries will increasingly look to nuclear power to meet their energy needs, particularly in the light of depleting oil and gas reserves, their increasing cost and the possible 'greenhouse effect' produced by fossil fuels. International cooperation concerning uranium reserves, reprocessing and technology transfer may need World Bank funding. (U.K.)

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.)

System control model of a turbine for a BWR

International Nuclear Information System (INIS)

Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A.

2009-10-01

In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

Analysis of void reactivity measurements in full MOX BWR physics experiments

International Nuclear Information System (INIS)

Ando, Yoshihira; Yamamoto, Toru; Umano, Takuya

2008-01-01

In the full MOX BWR physics experiments, FUBILA, four 9x9 test assemblies simulating BWR full MOX assemblies were located in the center of the core. Changing the in-channel moderator condition of the four assemblies from 0% void to 40% and 70% void mock-up, void reactivity was measured using Amplified Source Method (ASM) technique in the subcritical cores, in which three fission chambers were located. ASM correction factors necessary to express the consistency of the detector efficiency between measured core configurations were calculated using collision probability cell calculation and 3D-transport core calculation with the nuclear data library, JENDL-3.3. Measured reactivity worth with ASM correction factor was compared with the calculated results obtained through a diffusion, transport and continuous energy Monte Carlo calculation respectively. It was confirmed that the measured void reactivity worth was reproduced well by calculations. (author)

An application of risk-informed evaluation on MOVs and AOVs for Taiwan BWR-type nuclear power plants

International Nuclear Information System (INIS)

Ting, K.; Chen, K.T.; Li, Y.C.; Hwang, S.H.; Chien, F.T.; Kang, J.C.

2008-01-01

Implementing a risk-informed inservice testing (RI-IST) program provides a good aspect to the nuclear power plant licensee as an alternating program in the current ASME Section XI and 10 CFR 50.55a relevant testing programs. RI-IST concentrates testing resources on highly significant components, reduces excess testing burden, increases plant's availability, decreases dose rate on the plant's staff and also reduces cost on plant's operation and maintenance under nuclear safety expectations. Furthermore, RI-IST also gives a feature on prospective licensing change basis to a nuclear power plant's licensee. This study will focus on safety-related and PRA-molded motor-operated valves (MOVs) and air-operated valves (AOVs) under the inservice testing program in boiling water reactor (BWR)-type nuclear power plant. As MOVs and AOVs have crucial safety functions throughout the nuclear power plant's safety systems, the steady operation and performance of MOVs and AOVs will definitely ensure that the nuclear power plant operates under safety expectations; therefore, this is the key reason to implement risk-informed evaluation for MOVs and AOVs in this study and being able to provide the safety significance classification for MOVs and AOVs under the current IST program to the plant's management. As a pilot study of RI-IST, the methodology of qualitative assessment will incorporate with probabilistic risk assessment (PRA) analyzing MOVs' and AOVs' safety significance within the current PRA model. The evaluating result will then classify its safety significance into a high-safety significant component (HSSC) and a low-safety significant component (LSSC) for MOVs and AOVs based on relevant regulatory criteria. With this initiating achievement, it can provide a cornerstone for further studies on the other types of valves and pumps in RI-IST program and also provide a valuable reference as proposing license change to the licensee

High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

International Nuclear Information System (INIS)

Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, A.

2013-01-01

The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

Final results of the XR2-1 BWR metallic melt relocation experiment

International Nuclear Information System (INIS)

Gauntt, R.O.; Humphries, L.L.

1997-08-01

This report documents the final results of the XR2-1 boiling water reactor (BWR) metallic melt relocation experiment, conducted at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. The objective of this experiment was to investigate the material relocation processes and relocation pathways in a dry BWR core following a severe nuclear reactor accident such as an unrecovered station blackout accident. The imposed test conditions (initial thermal state and the melt generation rates) simulated the conditions for the postulated accident scenario and the prototypic design of the lower core test section (in composition and in geometry) ensured that thermal masses and physical flow barriers were modeled adequately. The experiment has shown that, under dry core conditions, the metallic core materials that melt and drain from the upper core regions can drain from the core region entirely without formation of robust coherent blockages in the lower core. Temporary blockages that suspended pools of molten metal later melted, allowing the metals to continue draining downward. The test facility and instrumentation are described in detail. The test progression and results are presented and compared to MERIS code analyses. 6 refs., 55 figs., 4 tabs

BWR Assembly Optimization for Minor Actinide Recycling

International Nuclear Information System (INIS)

Maldonado, G. Ivan; Christenson, John M.; Renier, J.P.; Marcille, T.F.; Casal, J.

2010-01-01

The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

Energy Technology Data Exchange (ETDEWEB)

Tavera D, L; Camacho L, M E

1991-01-15

The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

Nuclear weapons and the World Court ruling

International Nuclear Information System (INIS)

Singh, J.

1998-01-01

based on the initiatives by non-governmental organizations, the World Health Organisation (WHO) Assembly asked the International Court of Justice for an advisory opinion in 1993 whether, considering the environmental and health consequences, the use of nuclear weapons by a state in war or other armed conflict would be a breach of its obligations under international law. The World Court decided that it was not able to give an advisory opinion as requested, because of the fact that questions of use of force and such like were beyond the scope of specialized agencies like the WHO. The Court has ruled that the international community, especially the five nuclear weapon states have not only an obligation to negotiate a treaty for total nuclear disarmament, but also have an obligation to conclude such treaty. We may expect that the nuclear weapon states will cynically disregard the ruling of the World Court as they have been doing to the basic obligation itself in pursuit of nuclear hegemony. But the remaining 150 countries or so also bear a responsibility to keep nudging the recalcitrant states into implementing their commitments to disarm

World Council of Nuclear Workers

International Nuclear Information System (INIS)

Maisseu, Andre

2007-01-01

WONUC is an association of Trade Unions, Scientific Societies and Social Organizations of the employees, workers and professionals of the nuclear energy related industries and technologies; integrated by 35 Countries and 1.8 millions members. This paper expose the products and services that WONUC provide for the promotion of peaceful uses of nuclear energy and the result of their work around all the world

List of the world's nuclear power plants

International Nuclear Information System (INIS)

Kempken, M.

1984-01-01

This list published once a year presents, subdivided into countries, data on all nuclear power plants in operation, under construction, or for which a contract has been placed, referring to the following aspects: Year the contract has been placed, name and/or size, owner or operator, design type, manufacturers, net output, first year of commercial operation, and total electricity output up to the data June 30, 1984. Two additional tables present a survey on the world's nuclear power plants, also grouped by countries, and the largest commercially used nuclear power plants of the world. (UA) [de

BWR stability analysis

International Nuclear Information System (INIS)

Valtonen, K.

1990-01-01

The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

The World Nuclear Industry Status Report 2015

International Nuclear Information System (INIS)

Schneider, Mycle; Froggatt, Antony; Hazemann, Julie; Katsuta, Tadahiro; Ramana, M.V.; Thomas, Steve; Porritt, Jonathon

2015-07-01

The World Nuclear Industry Status Report 2015 provides a comprehensive overview of nuclear power plant data, including information on operation, production and construction. The WNISR assesses the status of new-build programs in current nuclear countries as well as in potential newcomer countries. Japan without nuclear power for a full calendar year for the first time since the first commercial nuclear power plant started up in the country 50 years ago. Nuclear plant construction starts plunge from fifteen in 2010 to three in 2014. 62 reactors under construction - five fewer than a year ago - of which at least three-quarters delayed. In 10 of the 14 building countries all projects are delayed, often by years. Five units have been listed as 'under construction' for over 30 years. Share of nuclear power in global electricity mix stable at less than 11% for a third year in a row. AREVA, technically bankrupt, downgraded to 'junk' by Standard and Poor's, sees its share value plunge to a new historic low on 9 July 2015-a value loss of 90 percent since 2007 China, Germany, Japan-three of the world's four largest economies-plus Brazil, India, Mexico, the Netherlands, and Spain, now all generate more electricity from non-hydro renewables than from nuclear power. These eight countries represent more than three billion people or 45 percent of the world's population. In the UK, electricity output from renewable sources, including hydropower, overtook the output from nuclear. Compared to 1997, when the Kyoto Protocol on climate change was signed, in 2014 there was an additional 694 TWh per year of wind power and 185 TWh of solar photovoltaics- each exceeding nuclear's additional 147 TWh

Fundamentals of boiling water reactor (BWR)

International Nuclear Information System (INIS)

Bozzola, S.

1982-01-01

These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

Study of transient turbine shot without bypass in a BWR; Estudio del transitorio disparo de turbina sin bypass en un BWR

Energy Technology Data Exchange (ETDEWEB)

Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

2015-09-15

The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

Nuclear physics in colourful worlds. Quantumchromodynamics and nuclear binding

International Nuclear Information System (INIS)

Muether, H.; Engelbrecht, C.A.; Brown, G.E.

1987-01-01

When quantumchromodynamics (QCD) is generalized from SU(3) to an SU(N c ) gauge theory, where N c is the number of colours, it depends on only two parameters: N c and the bare quark mass m q . A more general understanding of nuclear physics can be achieved by considering what it would be like in worlds with the number of colours different from 3, and bare quark masses different from the 'empirical' ones. Such an investigation can be carried out within a framework of meson-exchange interactions. The empirical binding energy of nuclear matter results from a very near cancellation between attractive and repulsive terms which are two orders of magnitude larger and may be expected to depend sensitively on the parameters of QCD. It is indeed found that our world is wedged into a small corner of the two-dimensional manifold of m q versus N c . If the number of colours were decreased by one, or the bare quark masses raised by more than 20%, nuclear matter would become unbound. By tracing the origin of this state of affairs, one obtains a clearer picture of the relative importance of various effects on the behaviour of the bulk nuclear matter. In particular, correlations like those embodied in the Coester band of saturation points appear to have a broader degree of validity than is implied by fits to the actual physical world only. (orig.)

Power generator in BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1984-01-01

Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

Thermohydraulic stability coupled to the neutronic in a BWR

International Nuclear Information System (INIS)

Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.

2006-01-01

In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the Laguna Verde

BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

International Nuclear Information System (INIS)

Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

1983-09-01

TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

International Nuclear Information System (INIS)

Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

2005-01-01

Full text of publication follows: The goal of this project is to develop an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. This code will include more fundamental physical models than the current generation of sub-channel codes and advanced numerical algorithms for improved computational accuracy, robustness, and speed. It is highly desirable to understand the detailed two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for the analysis of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is still too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Recent progress in Computational Fluid Dynamics (CFD), coupled with the rapidly increasing computational power of massively parallel computers, shows promising potential for the fine-mesh, detailed simulation of fuel assembly two-phase flow phenomena. However, the phenomenological models available in the commercial CFD programs are not as advanced as those currently being used in the sub-channel codes used in the nuclear industry. In particular, there are no models currently available which are able to reliably predict the nature of the flow regimes, and use the appropriate sub-models for those flow regimes. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD Code STAR-CD which provides general two-phase flow modeling capabilities. The paper describes the model development strategy which has been adopted by the development team for the

Investigations on the thermal-hydraulics of a natural circulation cooled BWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Kok, H.V.; Hagen, T.H.J.J. van der; Mudde, R.F. [Delft Univ. of Technology (Netherlands)

1995-09-01

A scaled natural circulation loop facility has been built after the Dodewaard Boiling Water Reactor, which is the only operating natural circulation cooled BWR in the world. The loop comprises one fuel assembly, a riser with a downcomer and a condenser with a cooling system. Freon-12 is used as a scaling liquid. This paper reports on the first measurements done with this facility. Quantities like the circulation flow, carry-under and the void-fraction have been measured as a function of power, pressure, liquid level, riser length, condensate temperature and friction factors. The behavior of the circulation flow can be understood by considering the driving force. Special attention has been paid to the carry-under, which has been shown to have a very important impact on the dynamics of a natural circulation cooled BWR.

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Latest experiences in inspecting the inside of BWR vessel shields

Energy Technology Data Exchange (ETDEWEB)

Alberdi, R.; Gonzalez, E.

2001-07-01

In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

World nuclear generating capacity 1993/94

International Nuclear Information System (INIS)

Anon.

1994-01-01

This article is the annual summary of world nuclear generating capacity for 1994. A global summary is first provided, reviewing total installed capacity and growth in installed capacity over the next five years. A more detailed discussion of the nuclear efforts in 34 countries follows, with a tabular listing of nuclear projects in each of these countries. The listing includes reactor supplier, reactor type, size, current status, and date of commercial operation

BWR type reactors

International Nuclear Information System (INIS)

Watanabe, Shoichi

1986-01-01

Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

EPRI BWR Water Chemistry Guidelines Revision

International Nuclear Information System (INIS)

Garcia, Susan E.; Giannelli, Joseph F.

2014-01-01

BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

The world nuclear industry status report 2007

International Nuclear Information System (INIS)

Schneider, M.; Froggatt, A.

2007-11-01

The status and perspectives of the nuclear industry in the world have been subject to a large number of publications and considerable media attention over the last few years. The present report attempts to provide solid elements of key information for intelligent analysis and informed decision-making. As of 1 November 2007 there are 439 nuclear reactors operating in the world. That is five less than five years ago. There are 32 units listed by the International Atomic Energy Agency (IAEA) as 'under construction'. That is about 20 less than in the late 1990's. In 1989 a total of 177 nuclear reactors had been operated in what are now the 27 EU Member States. That number shrank to 146 units as of 1 November 2007. In 1992 the Worldwatch Institute in Washington, WISE-Paris and Greenpeace International published the first World Nuclear Industry Status Report. As a first updated review in 2004 showed the 1992 analyses proved correct. In reality, the combined installed nuclear capacity of the 436 units operating in the world in the year 2000 was less than 352,000 megawatts - to be compared with the forecast of the International Atomic Energy Agency (IAEA) from the 1970's of up to 4,450,000 megawatts. Today the 439 worldwide operating reactors total 371,000 megawatts. Nuclear power plants provide 16% of the electricity, 6% of the commercial primary energy and 2-3% of the final energy in the world - the tendency is downwards - less than hydropower alone. Twenty-one of the 31 countries operating nuclear power plants decreased their share of nuclear power within the electricity mix if compared with 2003. The average age of the operating power plants is 23 years. Some nuclear utilities envisage reactor lifetimes of 40 years or more. Considering the fact that the average age of all 117 units that have already been closed is equally about 22 years, the doubling of the operational lifetime seems already rather optimistic. However, we have assumed an average lifetime of 40 years

The world nuclear industry status report 2007

Energy Technology Data Exchange (ETDEWEB)

Schneider, M.; Froggatt, A

2007-11-15

The status and perspectives of the nuclear industry in the world have been subject to a large number of publications and considerable media attention over the last few years. The present report attempts to provide solid elements of key information for intelligent analysis and informed decision-making. As of 1 November 2007 there are 439 nuclear reactors operating in the world. That is five less than five years ago. There are 32 units listed by the International Atomic Energy Agency (IAEA) as 'under construction'. That is about 20 less than in the late 1990's. In 1989 a total of 177 nuclear reactors had been operated in what are now the 27 EU Member States. That number shrank to 146 units as of 1 November 2007. In 1992 the Worldwatch Institute in Washington, WISE-Paris and Greenpeace International published the first World Nuclear Industry Status Report. As a first updated review in 2004 showed the 1992 analyses proved correct. In reality, the combined installed nuclear capacity of the 436 units operating in the world in the year 2000 was less than 352,000 megawatts - to be compared with the forecast of the International Atomic Energy Agency (IAEA) from the 1970's of up to 4,450,000 megawatts. Today the 439 worldwide operating reactors total 371,000 megawatts. Nuclear power plants provide 16% of the electricity, 6% of the commercial primary energy and 2-3% of the final energy in the world - the tendency is downwards - less than hydropower alone. Twenty-one of the 31 countries operating nuclear power plants decreased their share of nuclear power within the electricity mix if compared with 2003. The average age of the operating power plants is 23 years. Some nuclear utilities envisage reactor lifetimes of 40 years or more. Considering the fact that the average age of all 117 units that have already been closed is equally about 22 years, the doubling of the operational lifetime seems already rather optimistic. However, we have assumed an average

Trend of field data on pipe wall thinning for BWR power plants

International Nuclear Information System (INIS)

Hakii, Junichi; Hiranuma, Naoki; Hidaka, Akitaka

2009-01-01

Strongly motivated by every stakeholder not to repeat Mihama Nuclear Power Station pipe rupture accident in August 2004, JSME Main Committee on Codes and Standards on Power Generation Facilities immediately launched a special task force to develop Rules on Pipe Wall Thinning Management for BWR, PWR and fossil Power Plants respectively. The authors describes the process of the development of Rules for BWR Power Plans from the view point of collections and analysis of fields data of pipe wall thinning. Through its activities, the authors confirmed the existing findings, like the effect of Oxygen injection, turbulence and dependence on coolant temperature, derived from series of laboratory-scaled experiments in FAC and coolant velocities effects in LDI. Further based upon the said proven findings with field data, they explain the adequacy of major concept of the rule such as separate treatment of FAC (Flow Accelerated Corrosion) and LDI (Liquid Droplet Impingement). (author)

Prospects of uranium supply-demand situation in world nuclear power development

International Nuclear Information System (INIS)

Chen Zuyi; Wang Xingwu

2010-01-01

Based on the newest materials and data published by authoritative organizations, this paper introduces the near-term and medium to long-term development situation of world nuclear power, summarizes the main characteristics of recent world uranium production, preliminarily analyses the relationship between uranium supply and demand to 2030. It is suggested that from the view-point of whole world, uranium resources are fully sufficient for the near-term and medium to long-term world uranium production and uranium demand of nuclear power. World uranium production can meet the near-term uranium demand for nuclear power. However, a big supply-demand gap may exist after 2015 as world nuclear power will be developed with high speed. In case if all const ruction plans of new uranium mines and production- expansion plans of existing uranium mines will be completed on time, it is quite possible for the world uranium production to meet the long-term uranium demand of nuclear power development. (authors)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

World electricity: will nuclear doubts affect growth

International Nuclear Information System (INIS)

Baum, Vladimir.

1986-01-01

The world has shown a healthy appetite for electricity even during the years of high energy prices. Between 1970 and 1985 worldwide electricity production increased by 92%, from 4,906.7 terawatt hours (TWh) to 9,421.7 TWh (1TWh = 10 9 KWh). In the same period total world energy consumption rose by 44.8% from 220.2 exajoules to 318.8 EJ (1 EJ = 23.88 million tonnes of oil equivalent). The major part of this growth occurred in the 1970s. Over the last five years, from 1980 to 1985, world energy consumption inched forward only by 7.2%, while notwithstanding widespread economic recession, electricity production advanced by 16.1%, with nuclear power responsible for an increasing share. These figures are tabulated and analysed on a worldwide regional basis. The amount of electricity produced by nuclear power plants is given, and the situation in particular countries noted. The projected future electricity demand and future nuclear electricity generating capacity are given. The effect of the Chernobyl incident is assessed. It may prove to be the beginning of the end of nuclear energy or just an unfortunate hiccough in its progress. (U.K.)

Status report: Intergranular stress corrosion cracking of BWR core shrouds and other internal components

International Nuclear Information System (INIS)

1996-03-01

On July 25, 1994, the US Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 94-03 to obtain information needed to assess compliance with regulatory requirements regarding the structural integrity of core shrouds in domestic boiling water reactors (BWRs). This report begins with a brief description of the safety significance of intergranular stress corrosion cracking (IGSCC) as it relates to the design and function of BWR core shrouds and other internal components. It then presents a brief history of shroud cracking events both in the US and abroad, followed by an indepth summary of the industry actions to address the issue of IGSCC in BWR core shrouds and other internal components. This report summarizes the staff's basis for issuing GL 94-03, as well as the staff's assessment of plant-specific responses to GL 94-03. The staff is continually evaluating the licensee inspection programs and the results from examinations of BWR core shrouds and other internal components. This report is representative of submittals to and evaluations by the staff as of September 30, 1995. An update of this report will be issued at a later date

Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

International Nuclear Information System (INIS)

Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

2001-01-01

Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

Prospects for the world nuclear energy market

Energy Technology Data Exchange (ETDEWEB)

1976-04-01

Over the last few years projections of nuclear power generating capacity growth for the next two decades have progressively decreased. Dwindling load growth, increasing load lead time, costs of delays and high cost inflation, industrial recession, and fuel cycle delays are discussed as the main causes of the setback. The state of the fuel cycle business in the world market is examined and data are presented and discussed for predicted world supply and demand. Nuclear plans and fuel policies and requirements are then examined for individual countries.

Level controlling system in BWR type reactors

International Nuclear Information System (INIS)

Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

1981-01-01

Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

BWR Services maintenance training program

International Nuclear Information System (INIS)

Cox, J.H.; Chittenden, W.F.

1979-01-01

BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

Safe nuclear power for the Third World

International Nuclear Information System (INIS)

Johnson, W.R.; Lyon, C.F.; Redick, J.R.

1989-01-01

It is clear that using nuclear power for the generation of electricity is one way of reducing the emissions of CO 2 and other gases that contribute to the greenhouse effect. Equally clear is the fact that the reduction can be magnified by converting domestic, commercial, and industrial power-consuming activities from the direct use of fossil fuel sources to electrical energy. A major area for future progress in limiting CO 2 emissions is in the Third World, where population growth and expectations for a higher social and economic standard of living portend vast increases in future energy use. A number of problems come to mind as one contemplates the widespread expansion of nuclear energy use into the Third World. The authors propose a method involving the marriage of two currently evolving concepts by which nuclear electrical generation can be expanded throughout the world in a manner that will address these problems. The idea is to form multinational independent electric generating companies, or nuclear electric companies (NECs), that would design, build, operate, and service a standardized fleet of nuclear power plants. The plants would be of the Integral Fast Reactor (IFR) design, now under development at Argonne National Laboratory, and, in particular, a commercial conceptualization of the IFR sponsored by General Electric Company, the Power Reactor Inherently Safe Module (PRISM)

Evaluation of thermal margin during BWR neutron flux oscillation

International Nuclear Information System (INIS)

Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

1992-01-01

Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

Logical model for the control of a BWR turbine

International Nuclear Information System (INIS)

Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.

2009-01-01

In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

International Nuclear Information System (INIS)

Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.; Gauld, Ian C.; Ilas, Germina; Marshall, William BJ J.

2015-01-01

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10x10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

Aging assessment of BWR control rod drive systems

International Nuclear Information System (INIS)

Greene, R.H.

1992-01-01

This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

Aging assessment of BWR control rod drive systems

International Nuclear Information System (INIS)

Greene, R.H.

1991-01-01

This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the CRD system, and (4) personal information exchange with industry experts. As part of this study, nearly 3500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation was conducted to summarize the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented that identify specific actions utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain maintenance practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities. 5 refs., 8 figs., 2 tabs

Aging assessment of BWR control rod drive systems

International Nuclear Information System (INIS)

Greene, R.H.

1991-01-01

This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

2011-11-15

In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

Moderator temperature coefficient in BWR core

International Nuclear Information System (INIS)

Naito, Yoshitaka

1977-01-01

Temperature dependences of infinite multiplication factor k sub(infinity) and neutron leakage from the core must be examined for estimation of moderator temperature coefficient. Temperature dependence on k sub(infinity) has been investigated by many researchers, however, the dependence on neutron leakage of a BWR with cruciformed control rods has hardly been done. Because there are difficulties and necessity on calculations of three space dimensional and multi-energy groups neutron distribution in a BWR core. In this study, moderator temperature coefficients of JPDR-II (BWR) core were obtained by calculation with DIFFUSION-ACE, which is newly developed three-dimensional multi-group computer code. The results were compared with experimental data measured from 20 to 275 0 C of the moderator temperature and the good agreement was obtained between calculation and measurement. In order to evaluate neutron leakage from the core, the other two calculations were carried out, adjusting criticality by uniform absorption rate and by material buckling. The former underestimated neutron leakage and the latter overestimated it. Discussion on the results shows that in order to estimate the temperature coefficient of BWR, neutron leakage must be evaluated precisely, therefore the calculation at actual pattern of control rods is necessary. (auth.)

Nuclear energetics all over the world

International Nuclear Information System (INIS)

Wojcik, T.

2000-01-01

The actual state and tendencies of nuclear power further development for different world regions have been presented and discussed. The problem of safety of energetic nuclear reactors, radioactive waste management and related problems have been discussed in respect of regulations in different countries. The economical aspects of nuclear energetics in comparison with different fossil fuel power plants exploitation costs has been presented as well. The official state of international organizations (IAEA, WANO, HASA etc.) have been also shown in respect to subject presented

World nuclear capacity and fuel cycle requirements 1992

International Nuclear Information System (INIS)

1992-12-01

This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy's activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration's annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment for the Lower and Upper Reference case scenarios were obtained from the Office of Integrated Analysis and Forecasting, Energy Information Administration. Most of these projections were developed using the World Integrated Nuclear Evaluation System (WINES) model

Approach to securing of stable nuclear fuel supplies

International Nuclear Information System (INIS)

Koike, Kunihisa; Imamura, Isao; Noda, Tetsuya

2010-01-01

With the dual objectives of not only ensuring stable electric power supplies but also preventing global warming, the construction of new nuclear power plants is being planned in many countries throughout the world. Toshiba and Westinghouse Electric Company (WEC), a member of the Toshiba Group, are capable of supplying both boiling water reactor (BWR) and pressurized water reactor (PWR) plants to satisfy a broad range of customer requirements. Furthermore, to meet the growing demand for the securing of nuclear fuel supplies, Toshiba and WEC have been promoting the strengthening and further expansion of supply chains in the fields of uranium production, uranium hexafluoride (UF 6 ) conversion, uranium enrichment, and fuel fabrication. (author)

Usage of industrial robots in nuclear power industry

International Nuclear Information System (INIS)

Matsuo, Yoshio; Hamada, Kenjiro

1982-01-01

Japan is now at the top level in the world in robot technology.Its application to nuclear power field is one of the most expected. However, their usage spreads over various types of nuclear power plants, their manufacture and operation, and other areas such as fuel reprocessing plants and reactor plant decommissioning. The robots as used for the operation of BWR nuclear power plants, already developed and under development, are described: features in the nuclear-power usage of robots, the robots used currently for automatic fuel exchange, the replacement of control rod drives and in-service inspection; the robots under development for travelling inspection device and the inspection of main steam-relief safety valves, future development of robots. By robot usage, necessary personnel, work period and radiation exposure can be greatly reduced, and safety and reliability are also raised. (Mori, K.)

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

The first Summer Institute of the World Nuclear University - a personal record

International Nuclear Information System (INIS)

Denk, W.; Fischer, C.; Seidl, M.

2005-01-01

The first World Nuclear University Summer Institute was held at Idaho Falls, USA, between July 9 and August 20, 2005. The event was hosted by the Institute of Nuclear Science and Engineering of Idaho State University (ISU) and by the Idaho National Laboratory (INL), which has been planned to be the central nuclear technology research institution in the United States. The World Nuclear University (WNU) was founded in 2003 by the International Atomic Energy Agency (IAEA), the OECD Nuclear Energy Agency (OECD-NEA), the World Association of Nuclear Operators (WANO), and the World Nuclear Association (WNA) as a global association fo scientific and educational institutions in the nuclear field. The first WNU Summer Institute was designed at IAEA in Vienna in the course of the following year and planned by the WNU Coordinating Centre in London. The six weeks of lectures and presentations arranged by the World nuclear University in Idaho Falls are described in detail from the participants' perspective. (orig.)

The world nuclear power engineering. 1998 year

International Nuclear Information System (INIS)

Preobrazhenskaya, L.B.

2000-01-01

The purpose of this article consists in the analysis of the state and prospects of the world nuclear power engineering development. The data on the ratio and value of electrical energy obtained at the NPPs in the world in 1998, the specific capital expenditures on the NPPs construction by 2005, the forecast for the capacity of all NPPs by 2020 are presented. The progress in developing nuclear power engineering conditioned by improvement of the NPPs operation, optimization of their life-cycle and developing of new NPPs projects is noted [ru

Completion of high-efficiency BWR turbine plant 'Hamaoka unit No. 4'

International Nuclear Information System (INIS)

Tsuji, Kunio; Hamaura, Norikazu; Shibashita, Naoaki; Kazama, Seiichi

1995-01-01

Accompanying the increase of capacity of nuclear power plants in Japan, the plants having heightened economical efficiency, which are supported by the improvement of thermal efficiency and the reduction of dose, are demanded. Hitachi Ltd. has completed No. 4 turbine unit of 1137 MW output in Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., which is the largest capacity machine in Japanese BWR plants. In this unit, the moisture separator heater, the steam turbine with high efficiency, and the hollow thread film condensate filter which treats the total flow rate of condensate are used as the reheating type BWR plant for the first time in Japan, and the plan of heightened economy and operation was adopted. It was confirmed by the trial for about 10 months that the planned performance was sufficiently satisfied, and the commercial operation was started in September, 1993. The features of the 1137 MW turbine unit are explained. The turbine is of tandem six-flow exhaust condensation type. Diffuser type low pressure turbine exhaust chambers, butterfly type combination intermediate valve are adopted. The stages with the blades having moisture-separating grooves were corrected. The reliability of the shaft system was improved. The adoption of the moisture separator heater and the application of the hollow thread film type condensate filter are explained. (K.I.)

Investigation on Current Status of World Nuclear Education and Training

International Nuclear Information System (INIS)

Shin, J. Y.; Min, M. J.; Noh, B. C.

2010-04-01

All over the world, the interest of nuclear energy is increasing and the expectations of it are getting more as one of the most practical alternative energy resources. However, since 1990s, as a lot of nuclear specialists are being retired, now the problem of manpower shortage is taken into consideration for all of us and will be continued until 2011. In this point of view, the good quality of the professional nuclear training and education systems and the nuclear education centers are requested in order to breed and supply the next generation nuclear scientists and engineers. Thus, the objective of this study is to explore the current status of world nuclear education for both of nuclear power countries and potential nuclear power utilization countries in the near future. This report introduces the importance of nuclear energy, the current status of world nuclear power plants operation and the contribution of nuclear energy. Besides, it also includes the nuclear energy development plan of potential nuclear developing countries in the near future. In addition, this study also explores the nuclear training and education systems of the nuclear development countries and the current status of nuclear education in various fields such as government, industries, nuclear power plants ect. Especially, as considering the status of nuclear education classified such as Asia, the Americas, East and West Europe, the Middle East and Africa, it shows the different characteristics of nuclear education systems in each regions aimed to identify the good practices on the nuclear education systems. Finally, through observation of international cooperation and networks of the various nuclear organizations, this will be contributed to the development of nuclear education for member states and be suggested the various of the direction of development for nuclear education in Korea. The report presents in the basis of the recent status data of the world nuclear education systems collected

The World Nuclear University and its Summer Institute

Energy Technology Data Exchange (ETDEWEB)

Borysova, Irina [World Nuclear Association - WNA, Summer Institute of the World Nuclear University - WNU, 22a Saint James' s Sq., SW1Y 4JH London (United Kingdom)

2008-07-01

The World Nuclear University is a global partnership committed to enhancing international education and leadership in the peaceful applications of nuclear science and technology. The central elements of the WNU partnership are: - The global organizations of the nuclear industry: WNA and WANO; - The inter-governmental nuclear agencies: IAEA and OECD-NEA; - Leading institutions of nuclear learning in some thirty countries. The WNU was inaugurated in 2003 as a non-profit corporation. Operationally, the WNU is a public-private partnership. On the public side, the WNUCC's multinational secretariat is composed mainly of nuclear professionals supplied by governments; the IAEA further assists with financial support for certain WNU activities. On the private side, the nuclear industry provides administrative, logistical and financial support via the WNA. WNU activities fall into six programmatic categories: 1. Facilitate Multinational Academic Cooperation. 2. Build Nuclear Leadership. 3. Foster Policy Consensus on Institutional and Technological Innovation. 4. Enhance Public Understanding. 5. Shape Scientific and Regulatory Consensus on Issues Affecting Nuclear Operations. 6. Strengthen International Workforce Professionalism. This presentation will describe the WNU programmes addressed to young professionals. Among such programmes, the flagship of the WNU is the WNU Summer Institute. This unique six-week course occurs in a different country each year, offering an inspiring career opportunity for some 100 outstanding young nuclear professionals and academics from around the world. The WNU-SI programme combines an extensive series of 'big picture' presentations from world-class experts with daily team-building exercises. In the process, WNU Fellows become part of a global network of future nuclear leaders. Other WNU programmes for younger generation in the nuclear industry will also be briefly covered in this presentation. (author)

The design and use of proficiency based BWR reactor maintenance and refuelling training mockups

International Nuclear Information System (INIS)

Ford, G.J.

1996-01-01

The purpose of this paper is to describe the ABB experience with the design and use of boiling water reactor training facilities. The training programs were developed and implemented in cooperation with the nuclear utilities. ABB operates two facilities, the ABB ATOM Light Water Reactor Service Center located in Vasteras, Sweden, and the ABB Combustion Engineering Nuclear Operations BWR Training Center located in Chattanooga, Tennessee, USA. The focus of the training centers are reactor maintenance and refueling activities plus the capability to develop and qualify tools, procedures and repair techniques

World nuclear power generation market and prospects of industry reorganization

International Nuclear Information System (INIS)

Murakami, Tomoko

2007-01-01

In late years there are many trends placing nuclear energy with important energy in various countries in the world due to a remarkable rise to an energy price, importance of energy security and a surge of recognition to a global environment problem. Overseas nuclear industry's acquisition by a Japanese nuclear power plant maker and its capital or business tie-up with an overseas company, were announced in succession in 2006. A nuclear power plant maker has played an extremely important role supporting wide technology in all stages of a design, construction, operation and maintenance in a nuclear power generation business. After having surveyed the recent trend of world nuclear power generation situation, a background and the summary of these acquisition/tie-ups made were investigated and analyzed to consider the influence that movement of such an industry gives a world nuclear power generation market. (T. Tanaka)

Neutron dosimetry. Environmental monitoring in a BWR type reactor

International Nuclear Information System (INIS)

Tavera D, L.; Camacho L, M.E.

1991-01-01

The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

Residual stress analysis in BWR pressure vessel attachments

International Nuclear Information System (INIS)

Dexter, R.J.; Leung, C.P.; Pont, D.

1992-06-01

Residual stresses from welding processes can be the primary driving force for stress corrosion cracking (SCC) in BWR components. Thus, a better understanding of the causes and nature of these residual stresses can help assess and remedy SCC. Numerical welding simulation software, such as SYSWELD, and material property data have been used to quantify residual stresses for application to SCC assessments in BWR components. Furthermore, parametric studies using SYSWELD have revealed which variables significantly affect predicted residual stress. Overall, numerical modeling techniques can be used to evaluate residual stress for SCC assessments of BWR components and to identify and plan future SCC research

Nuclear weapons proliferation as a world order problem

International Nuclear Information System (INIS)

Falk, R.

1977-01-01

World-order concerns have intensified recently in light of mounting evidence that a weapons capability will soon be within easy reach of more and more governments and of certain nongovernmental groupings as well. One reliable source estimates that by 1985 as many as fifty countries could ''produce enough plutonium each year for at least several dozen nuclear explosives.'' In an even more immediate sense, ''economic competition among nuclear suppliers today could soon lead to a world in which twenty or more nations are but a few months from a nuclear weapons force.'' Three developments have created this ''world order'' sense of concern: (1) increased pace of civilian nuclear power deployment globally as a consequence of rising oil prices, unreliability of oil supplies, and reality of dwindling oil reserves in any case; (2) actuality of India's nuclear explosion in May 1974 which demonstrated vividly how any state that pursues a ''civilian'' program can also develop its own weapons capability; and (3) the intensification of competition for international nuclear sales which makes it increasingly evident that nonproliferation goals are no longer compatible with the pursuit of national commercial advantage; essentially, this reality has emerged from a break in the American monopoly over civilian nuclear technology and the willingness of French and German suppliers to provide all elements of the nuclear fuel cycle, including enrichment and reprocessing facilities,to any nation that feels it can afford to buy them; the German-Brazilian deal (worth at least $4 billion) has proven to be the equivalent in the commercial realm of India's ''peaceful'' nuclear explosion. Such developments disclose the alarming prospect that easier access to nuclear technology will make it relatively simple and thus more likely for a beleaguered government or a desperate political actor of any sort to acquire and possibly use nuclear weapons

Nuclear safeguards and security in a changing world

International Nuclear Information System (INIS)

Badolato, E.V.

1986-01-01

Two major crises of 1986 - the Chernobyl nuclear accident and international terrorism have had the effect of making what everyone does even more critically important for U.S. national security and for the security of the world. Chernobyl can be a starting point for efforts to make nuclear power systems safer and more benign. It also poses very basic questions for nuclear arms control activities. A fundamental objective of the Administration's arms control policy is to achieve substantial and equitable reductions in U.S. and Soviet nuclear forces with effective verification. However, Chernobyl served to remind the U.S. once again of the obsessive secretiveness of the Soviet Union and the difficulties of obtaining information on Soviet nuclear weapon activities. All of this points to the importance of developing improved monitoring technologies and obtaining Soviet agreement on on-site inspection. Nuclear safeguards and security developments in response to a changing world are the topic of discussion in this paper

World Health Organization on nuclear power

International Nuclear Information System (INIS)

Anon.

1977-01-01

A report published by the World Health Organization in cooperation with, and at the instigation of, the Belgian authorities, is summarised. The report was prepared by an international multidisciplinary working group, and concentrated on the somatic and genetic risks from ionising radiation, the environmental effects of nuclear power from the mining of uranium to the disposal of waste and the probability and consequences of accidents, sabotage and theft of nuclear materials. In general positive to nuclear power, the report nevertheless recommends for RESEARCH AND EVALUATION IN SEVERAL SECTORS: The duties of the authorities in providing full and open information on the consequences of the exploitation of nuclear power are emphasised. (JIW)

BWR Steam Dryer Alternating Stress Assessment Procedures

Energy Technology Data Exchange (ETDEWEB)

Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2016-12-01

This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

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

The situation of the nuclear energy in the world

International Nuclear Information System (INIS)

Souza, Jair Albo Marques de

1996-12-01

This work presents an overview of the nuclear energy in the world. It approaches the following main topics: kinds of nuclear power plants; operation experience of the nuclear plants; environmental and social aspects of the nuclear energy; economic aspects of the nuclear energy; development of the reactors technology and supply of the nuclear fuel

Study of transient rod extraction failure without RBM in a BWR; Estudio del transitorio error de extraccion de barra sin RBM en un BWR

Energy Technology Data Exchange (ETDEWEB)

Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

2015-09-15

The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

From the nuclear world

International Nuclear Information System (INIS)

Anon.

2015-01-01

This document gathers pieces of information concerning nuclear industry worldwide. The most relevant are the following ones. China has announced the construction between 2020 and 2025 of the biggest particle accelerator in the world. The Finn government has agreed with the project of a spent fuel storage center, it was the last administrative step before the launching of the construction works. The quality of the steel of the pressure vessel of the Olkiluoto EPR has been assessed by the Finn nuclear safety authority (STUK). The French nuclear safety authority (ASN) has launched a test program for assessing the resistance of the pressure vessel of the Flamanville EPR as a consequence of the recent discovery of defects in the composition of the steel. A robot called SX1 has been designed to measure radiation in a continuous way by strolling about in a nuclear facility. As a consequence of a tense relationship with Russia, the Turkish government has stopped the construction (by Rosatom) of the Akkuyu nuclear plant. The Belgian government and the Electrabel company have signed an agreement for a 10 year extension of the operating life of the Doel 1 and 2 reactors. Chinese authorities have approved the construction of 4 new third generation reactors. (A.C.)

Simplified distributed parameters BWR dynamic model for transient and stability analysis

International Nuclear Information System (INIS)

Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro

2006-01-01

This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR

A practical methodology of radiological protection for the reduction of hot particles in BWR type reactors

International Nuclear Information System (INIS)

Alvarez G, G.

1991-01-01

The purpose of this work, in general form, is to describe a practical method for reduction of hot particles generated as consequence of the operational activities of BWR nuclear reactors. This methodology provides a description of the localizations and/or probable activities of finding particles highly radioactive denominated hot particles. For this purpose it was developed a strategy based on the decontamination lineaments, as well as the manipulation, gathering, registration, contention, documentation, control and final disposition of the hot particles. In addition, some recommendations are reiterated and alternative, in order to gathering the hot particles in a dynamic way given to the activities of the personal occupationally exposed in highly radioactive areas. The structure of the methodology of hot particles is supported in the radiological controls based on the Code of Federal Regulation 10 CFR 20 as well as the applicable regulatory documents. It provides an idea based on administrative controls of radiological protection, in order to suggesting the responsibilities and necessary directing for the control of the hot particles required in nuclear plants of the BWR type. (author)

Core concept for long operating cycle simplified BWR (LSBWR)

International Nuclear Information System (INIS)

Kouji, Hiraiwa; Noriyuki, Yoshida; Mikihide, Nakamaru; Hideaki, Heki; Masanori, Aritomi

2002-01-01

An innovative core concept for a long operating cycle simplified BWR (LSBWR) is currently being developed under a Toshiba Corporation and Tokyo Institute of Technology joint study. In this core concept, the combination of enriched uranium oxide fuels and loose-pitched lattice is adopted for an easy application of natural circulation. A combination of enriched gadolinium and 0.7-times sized small bundle with peripheral-positioned gadolinium rod is also adopted as a key design concept for 15-year cycle operation. Based on three-dimensional nuclear and thermal hydraulic calculation, a nuclear design for fuel bundle has been determined. Core performance has been evaluated based on this bundle design and shows that thermal performance and reactivity characteristics meet core design criteria. Additionally, a control rod operation plan for an extension of control rod life has been successfully determined. (author)

Nuclear power in the developing world

International Nuclear Information System (INIS)

Sokolov, Y.

2005-01-01

Current trends in the interest in nuclear power development confirm important changes in opinions around the world about nuclear power's future. Much of the expansion of nuclear power in the sustainable development scenarios takes place in developing countries. For these countries to introduce nuclear power, they need to pass through three main steps: energy planning, infrastructure development and then deployment. The paper gives an overview of the IAEA's activity in this area. In order to meeting the energy needs of developed and developing countries, developing a global vision for nuclear energy, assessing and clarifying the afford ability and acceptability requirements for large-scale nuclear energy use in the 21st century in both developed and developed countries, facilitating international cooperation in developing different types of new generation nuclear energy systems which meet these requirement, and facilitating international discussions aimed at establishing enhanced institutional system acceptable to both developed and developing countries

Managing nuclear weapons in a changing world: Proceedings

International Nuclear Information System (INIS)

1992-01-01

The Center for Security and Technology Studies was established at the Lawrence Livermore National Laboratory to support long-range technical studies on issues of importance to US national security. An important goal of the Center is to bring together Laboratory staff and the broader outside community through a program of technical studies, visitors, symposia, seminars, workshops, and publications. With this in mind, the Center and LLNL's Defense Systems Program sponsored a conference on Managing Nuclear Weapons in a Changing World held on November 17--18,1992. The first day of the meeting focused on nuclear weapons issues in the major geographical areas of the world. On the second day, the conference participants discussed what could be done to manage, control, and account for nuclear weapons in this changing world. Each of the talks and the concluding panel discussion are being indexed as separate documents

Managing nuclear weapons in a changing world: Proceedings

Energy Technology Data Exchange (ETDEWEB)

1992-12-31

The Center for Security and Technology Studies was established at the Lawrence Livermore National Laboratory to support long-range technical studies on issues of importance to US national security. An important goal of the Center is to bring together Laboratory staff and the broader outside community through a program of technical studies, visitors, symposia, seminars, workshops, and publications. With this in mind, the Center and LLNL`s Defense Systems Program sponsored a conference on Managing Nuclear Weapons in a Changing World held on November 17--18,1992. The first day of the meeting focused on nuclear weapons issues in the major geographical areas of the world. On the second day, the conference participants discussed what could be done to manage, control, and account for nuclear weapons in this changing world. Each of the talks and the concluding panel discussion are being indexed as separate documents.

Evaluation of internal flooding in a BWR

International Nuclear Information System (INIS)

Shiu, K.; Papazoglou, I.A.; Sun, Y.H.; Anavim, E.; Ilberg, D.

1985-01-01

Flooding inside a nuclear power station is capable of concurrently disabling redundant safety systems. This paper presents the results of a recent review study performed on internally-generated floods inside a boiling water reactor (BWR) reactor building. The study evaluated the flood initiator frequency due to either maintenance or ruptures using Markovian models. A time phased event tree approach was adopted to quantify the core damage frequency based on the flood initiator frequency. It is found in the study that the contribution to the total core damage due to internal flooding events is not insignificant and is comparable to other transient contributors. The findings also indicate that the operator plays an important role in the prevention as well as the mitigation of a flooding event

Manufacturing technology and process for BWR fuel

International Nuclear Information System (INIS)

Kato, Shigeru

1996-01-01

Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

World nuclear atlas. A step toward energy transition

International Nuclear Information System (INIS)

Lepage, Corinne; Laborde, Xemartin

2015-01-01

Illustrated by more than 120 maps and figures, this book proposes an overview of the world nuclear industry, of its development, and of the various strategies chosen within the perspective of energy transition. It proposes an overview of the status of nuclear energy in the world (presentation of the nuclear energy, development during the X X century, uranium production, fuel production and processing, the nuclear reactor industry), addresses the main controversies (health and environmental impact, waste management, opacity of the information, major accidents), the new challenges faced by the nuclear sector (a difficult assessment of huge costs, competition with renewable energies, a competitive environment, a technological uncertainty, transparency and democracy), the solutions chosen by big countries (USA, China, India, Japan, Europe, the German energy transition), and proposes a focus on France which is the only country which chose an all-nuclear strategy (history, nuclear installations, main actors, the myth of the French energy independence, the post-Fukushima French fleet, the case of the Fessenheim reactor, the EPR in question, the challenge of waste storage with the Cigeo project, the debate on the nuclear cost)

ABB advanced BWR and PWR fuel

International Nuclear Information System (INIS)

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

1999-01-01

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

1999 Nuclear power world report

International Nuclear Information System (INIS)

Wesselmann, C.

2000-01-01

Last year, 1999, nuclear power plants were available for energy supply and under construction, respectively, in 33 countries. A total of 436 nuclear power plants with an aggregate net power of 350.228 MWe and an aggregate gross power of 366.988 MWe were in operation in 31 countries. Four units with an aggregate of 2.900 MWe, i.e. Civaux 2 in France, Kaiga 2 and Rajasthan 3 in India, and Wolsung-4 in the Republic of Korea, went critical for the first time or started commercial operation after having been synchronized with the power grid. After 26 years of operation, the BN 350 sodium cooled fast breeder was permanently decommissioned in Kazakhstan. The plant not only generated electricity (its capacity was 135 MWe) but also supplied process heat to a seawater desalination plant. In 1999, however, it did not contribute to the supply of electricity. In Sweden, unit 1 of the Barsebaeck nuclear power station (600 Mwe net) was decommissioned because of political decisions. This step entails financial compensation payments and substitute electricity generating capacity made available to the power plant operators. Net electricity generation in 1999 amounts to approx. 2.395 Twh, which marks a 100 TWh increase over the preceding year. Since the first generation of electricity from nuclear power in 1951, the cumulated world generation amounts to nearly 37.200 TWh of electricity, and experience in the operation of nuclear power plants has increased to 9414 years. Last year, 38 plants were under construction. This slight increase is due to the start of construction of a total of seven projects: Two each in Japan, the Republic of Korea and Taiwan, and one in China. Shares of nuclear power differ widely among the operator countries. They reach 75 per cent in France, 73 per cent in Lithuania, and 58 per cent in Belgium. With a share of approx. 20 per cent and more than 720 TWh, the US is the largest producer worldwide of electricity from nuclear power. As far as the aggregate

Analysis CFD for the hydrogen transport in the primary containment of a BWR; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR

Energy Technology Data Exchange (ETDEWEB)

Jimenez P, D. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: guerreroazteca_69@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

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.)

BWR plant advanced central control panel PODIA

International Nuclear Information System (INIS)

Fujii, K.; Hayakawa, H.; Ikeda, Y.; Neda, T.; Suto, O.; Takamiya, S.

1983-01-01

BWR plant central control panels have become more and more enlarged and complicated recently due to the magnification of the scale of a plant and the requirement to reinforce safety. So, it is important to make communication between men and the complicated central control panel smooth. Toshiba has developed an advanced central control panel, named PODIA, which uses many computers and color CRTs, and PODIA is now in the stage of application to practical plants. In this article, the writers first touch upon control functions transition in the central control room, the PODIA position concerning the world-wide trend in this technology phase and the human engineering on the design. Then they present concrete design concepts for the control board and computer system which constitute PODIA

World's energy appetite may crave nuclear power

International Nuclear Information System (INIS)

Fulkerson, W.; Anderson, T.D.

1996-01-01

As scientists come to agree that global warming is a real phenomenon, it may be time to jumpstart the stalled nuclear industry. World population is expected to double by the end of the 21st century, and the lion's share of growth will be in developing nations. open-quotes More people and more economic activity will require more energy,close quotes say William Fulkerson, a senior fellow at the Joint Institute for Energy and the Environment in Knoxville, Tennessee, and Truman D. Anderson, formerly director of planning at Oak Ridge National Laboratory. There are only three viable options to fossil fuel plants, the authors say: nuclear fission, nuclear fusion, and such renewable energy sources as solar and wind. The advantages of nuclear energy are well known, the authors say. open-quotes It emits no greenhouse gases, and potentially it can be expanded almost without limit anywhere in the world, providing the controversies that surround it can be resolved.close quotes However, to garner public acceptance, a new generation of supersafe nuclear reactors, invulnerable to terrorism and conversion to weapons, will need to be developed, the authors say

Burnup credit feasibility for BWR spent fuel shipments

International Nuclear Information System (INIS)

Broadhead, B.L.

1990-01-01

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

From the nuclear world

International Nuclear Information System (INIS)

Anon.

2017-01-01

This document gathers pieces of information from around the world concerning nuclear industry. The most relevant ones are the following. AREVA NP and EDF have created a new society EDVANCE to combine their engineering teams in the fields of reactor core design and construction. The German Constitutional Court considers as illegal the nuclear fuel tax that was implemented in 2010 to balance public finance and fund the remediation of the Ass salt mine. In France on the Tricastin AREVA's site the ATLAS laboratory has opened its doors, it is the laboratory that will perform all the environmental and industrial analyses of this site. In Japan the reactors 3 and 4 of the Takahama power plant have resumed operations. Today 5 nuclear power reactors are operating on Japanese soil. The Indian government has announced its intention to build heavy water cooled nuclear reactors based on an Indian design. 22 reactors are operating in India representing a total of 6780 MW and 5 others are being built. According to the 'SMR Start' consortium public-private partnership contracts have to be promoted in order to launch the small modular reactor (SMR) technology. (A.C.)

Development of a computer program of fast calculation for the pre design of advanced nuclear fuel 10 x 10 for BWR type reactors

International Nuclear Information System (INIS)

Perusquia, R.; Montes, J.L.; Ortiz, J.J.

2005-01-01

In the National Institute of Nuclear Research (ININ) a methodology is developed to optimize the design of cells 10x10 of assemble fuels for reactors of water in boil or BWR. It was proposed a lineal calculation formula based on a coefficients matrix (of the change reason of the relative power due to changes in the enrichment of U-235) for estimate the relative powers by pin of a cell. With this it was developed the computer program of fast calculation named PreDiCeldas. The one which by means of a simple search algorithm allows to minimize the relative power peak maximum of cell or LPPF. This is achieved varying the distribution of U-235 inside the cell, maintaining in turn fixed its average enrichment. The accuracy in the estimation of the relative powers for pin is of the order from 1.9% when comparing it with results of the 'best estimate' HELIOS code. With the PreDiCeldas it was possible, at one minimum time of calculation, to re-design a reference cell diminishing the LPPF, to the beginning of the life, of 1.44 to a value of 1.31. With the cell design with low LPPF is sought to even design cycles but extensive that those reached at the moment in the BWR of the Laguna Verde Central. (Author)

Study of transient turbine shot without bypass in a BWR

International Nuclear Information System (INIS)

Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

2015-09-01

The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

The World Nuclear University Summer Institute

International Nuclear Information System (INIS)

Rivard, D.; McIntyre, M.

2007-01-01

The World Nuclear University (WNU) Summer Institute is a six weeks intensive training program aimed to develop a global leadership in the field of nuclear sciences and technologies. The topics covered include global setting, international regimes, technology innovation and nuclear industry operations. This event has been held annually since 2005. Mark McIntyre and Dominic Rivard attended this activity as a personal initiative. In this paper they will present the WNU and its Summer Institute, share their participation experience and discuss as well of some technical content covered during the Institute, highlighting the benefits this brought to their careers. (author)

Nuclear power in a changing world

International Nuclear Information System (INIS)

Taylor, J.

1996-01-01

Nuclear power has a future that will only be fully realised if it is shown to be a solution to some of the world's most pressing energy, and associated environmental, problems. This can only be done if nuclear power itself ceases to be perceived as a problem by the public, interest groups, governments and financial institutions. In public relations terms, this means removing the persistent distortions and misconceptions about the nuclear industry. Environmentally, it involves showing that nuclear power is the only alternative energy source which does not contribute to climate change, preserves rare minerals and recycles its raw materials. Governments must be persuaded to see that nuclear power is the only economic answer to the growing energy demand arising from increased industrialisation and population growth. Financiers need convincing that nuclear power is the investment of the future and generators that it is the lowest cost economic and environmental option. The future of nuclear power depends on meeting these challenges. (UK)

Examination of minor actinide annihilation by BWR core

International Nuclear Information System (INIS)

Hida, Kazuki

1995-01-01

From the viewpoint of reducing burden for disposing high level waste generated from spent fuel, the examination of recycling minor actinide (MA) to reactors and reducing its accumulation has been advanced. In this study, the possibility of annihilation in the case of recycling it to a BWR was examined. The main MAs are 237 Np, 241 Am, 243 Am, 242 Cm, and 244 Cm. However, as for Cm isotopes, the half life is short, the amount of generation is small, and the rate of neutron emission is high, therefore, those are disposed as waste, and 237 Np, 241 Am and 243 Am were taken as the objects of recycling. In order to grasp the basic characteristics in the case of recycling MAs to a BWR, MAs were added to UO 2 fuel, MOX fuel and HCR fuel and burned, and the nuclear conversion characteristics were examined. As the result, it was found that they were converted to short half life nuclides, and as the neutron spectra were softer, the rate of annihilation was higher. In the case of recycling MAs by concentrating to a specific reactor, reactivity loss, the degree of uranium enrichment required for compensating reactivity, and the rate of MA annihilation were calculated. Based on these data, the MA recycling system was set up, and the rate of MA annihilation was evaluated. This is reported. (K.I.)

Nuclear energy outlook: a GE perspective

International Nuclear Information System (INIS)

Fuller, J.

2006-01-01

Full text: Full text: As one of the world's leading suppliers of power generation and energy delivery technologies, GE Energy provides comprehensive solutions for coal, oil, natural gas and nuclear energy; renewable resources such as wind, solar and biogas, along with other alternative fuels. With the ever increasing demand for energy and pressures to decrease greenhouse gas emissions, global trends indicate a move towards building more base line nuclear generation capacity. As a reliable, cost-competitive option for commercial power generation, nuclear energy also addresses many of the issues the world faces when it comes to the environment. Since developing nuclear reactor technology in the 1950s, GE's Boiling Water Reactor (BWR) technology accounts for more than 90 operating plants in the world today. Building on that success, GE's ABWR design is now the first and only Generation 111 nuclear reactor in operation today. This advanced reactor technology, coupled with current construction experience and a qualified global supply chain, make ESBWR, GE's Generation III+ reactor design, an attractive option for owners considering adding nuclear generation capacity. In pursuit of new technologies, GE has teamed with Silex to develop, commercialize and license third generation laser enrichment technology. By acquiring the exclusive rights to develop and commercialize this technology, GE is positioned to support the anticipated global demands for enriched uranium. At GE, we are continuing to develop imaginative ideas and investing in products that are cost effective, increase productivity, limit greenhouse gas emissions, and improve safety and security for our customers

BWR vessel and internals project (BWRVIP)

International Nuclear Information System (INIS)

Bilanin, W.J.; Dyle, R.L.

1996-01-01

Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

Improvement for BWR operator training, 3

International Nuclear Information System (INIS)

Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

1990-01-01

BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

Assessment of two BWR accident management strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Petek, M.

1991-01-01

Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

Development of a BWR loading pattern design system based on modified genetic algorithms and knowledge

International Nuclear Information System (INIS)

Martin-del-Campo, Cecilia; Francois, Juan Luis; Avendano, Linda; Gonzalez, Mario

2004-01-01

An optimization system based on Genetic Algorithms (GAs), in combination with expert knowledge coded in heuristics rules, was developed for the design of optimized boiling water reactor (BWR) fuel loading patterns. The system was coded in a computer program named Loading Pattern Optimization System based on Genetic Algorithms, in which the optimization code uses GAs to select candidate solutions, and the core simulator code CM-PRESTO to evaluate them. A multi-objective function was built to maximize the cycle energy length while satisfying power and reactivity constraints used as BWR design parameters. Heuristic rules were applied to satisfy standard fuel management recommendations as the Control Cell Core and Low Leakage loading strategies, and octant symmetry. To test the system performance, an optimized cycle was designed and compared against an actual operating cycle of Laguna Verde Nuclear Power Plant, Unit I

Nuclear power and international cooperation - perceptions of the third world

International Nuclear Information System (INIS)

Khan, M.A.

1983-01-01

The views of the Third World that need to be given consideration in international nuclear policy-making are presented in the following topical sections: background summary of developing countries energy needs and sources, incentives for nuclear power development in developing countries, the need for nuclear cooperation, the Non-proliferation Treaty, erosion of confidence of the recipient states in the reliability of international cooperation agreements, and perceptions of the Third World regarding energy and proliferation

Nuclear energy education scenario around the world

International Nuclear Information System (INIS)

Barabas, Roberta de Carvalho; Sabundjian, Gaiane

2013-01-01

Nuclear energy has been used as a source of clean energy with many benefits. Nevertheless, it is still addressed with prejudice. The atomic bombing of Hiroshima and Nagasaki during World War II (1945), the Three Mile Island accident (1979), Chernobyl accident (1986), the crash of the cesium-137 in Goiana, Brazil (1987), and the recent accident in Fukushima (2011) may have been responsible for the negative image of nuclear energy. Researches on education have been conducted with students concerning the conceptual and practical issues of nuclear energy. This work aims to review the literature about nuclear energy education around the world in both, elementary school and high school. Since most educational researches on nuclear energy were published after 1980, this literature review covered the researches that have been published since 1980. The data were presented in chronological order. The results from the literature review provided a clear visualization of the global nuclear energy educational scenario, showing that the theme is still addressed with prejudice due to an incorrect view of nuclear energy and a limited view of its benefits. Concerning the science textbooks, the literature reports that the theme should be better addressed, encouraging students to research more about it. The data from this literature review will serve as a reference for a future proposal for a teaching training program for Brazilian science/physics high school teachers using a new teaching approach. (author)

Nuclear energy education scenario around the world

Energy Technology Data Exchange (ETDEWEB)

Barabas, Roberta de Carvalho; Sabundjian, Gaiane, E-mail: praroberta@uol.com.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

Nuclear energy has been used as a source of clean energy with many benefits. Nevertheless, it is still addressed with prejudice. The atomic bombing of Hiroshima and Nagasaki during World War II (1945), the Three Mile Island accident (1979), Chernobyl accident (1986), the crash of the cesium-137 in Goiana, Brazil (1987), and the recent accident in Fukushima (2011) may have been responsible for the negative image of nuclear energy. Researches on education have been conducted with students concerning the conceptual and practical issues of nuclear energy. This work aims to review the literature about nuclear energy education around the world in both, elementary school and high school. Since most educational researches on nuclear energy were published after 1980, this literature review covered the researches that have been published since 1980. The data were presented in chronological order. The results from the literature review provided a clear visualization of the global nuclear energy educational scenario, showing that the theme is still addressed with prejudice due to an incorrect view of nuclear energy and a limited view of its benefits. Concerning the science textbooks, the literature reports that the theme should be better addressed, encouraging students to research more about it. The data from this literature review will serve as a reference for a future proposal for a teaching training program for Brazilian science/physics high school teachers using a new teaching approach. (author)

Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor

International Nuclear Information System (INIS)

Escorcia O, D.; Salazar S, E.

2016-09-01

The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

Kinematics of two-phase mixture level motion in BWR pressure vessels

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Mallen, A.N.; Stritar, A.

1985-01-01

A model is presented for predicting two-phase mixture level elevations in BWR systems. The model accounts for the particular geometry and conditions in a BWR system during Small-Break Loss of Coolant Accidents. The model presented here is particularly suitable for efficient, high-speed simulations on small minicomputers. The model has been implemented and tested. Results are shown from BWR ATWS simulations

Analysis of results of AZTRAN and AZKIND codes for a BWR; Analisis de resultados de los codigos AZTRAN y AZKIND para un BWR

Energy Technology Data Exchange (ETDEWEB)

Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M., E-mail: gbo729@yahoo.com.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

An application of the process computer and CRT display system in BWR nuclear power station

International Nuclear Information System (INIS)

Goto, Seiichiro; Aoki, Retsu; Kawahara, Haruo; Sato, Takahisa

1975-01-01

A color CRT display system was combined with a process computer in some BWR nuclear power plants in Japan. Although the present control system uses the CRT display system only as an output device of the process computer, it has various advantages over conventional control panel as an efficient plant-operator interface. Various graphic displays are classified into four categories. The first is operational guide which includes the display of control rod worth minimizer and that of rod block monitor. The second is the display of the results of core performance calculation which include axial and radial distributions of power output, exit quality, channel flow rate, CHFR (critical heat flux ratio), FLPD (fraction of linear power density), etc. The third is the display of process variables and corresponding computational values. The readings of LPRM, control rod position and the process data concerning turbines and feed water system are included in this category. The fourth category includes the differential axial power distribution between base power distribution (obtained from TIP) and the reading of each LPRM detector, and the display of various input parameters being used by the process computer. Many photographs are presented to show examples of those applications. (Aoki, K.)

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.)

Improvement for BWR operator training

International Nuclear Information System (INIS)

Tsuchiya, Toshio; Masuda, Hisao; Isono, Tomoyuki; Noji, Kunio; Togo, Toshiki

1989-01-01

BWR Operator Training Center Corporation (BTC) was established in April 1971 for the purpose of training the operators from all BWR utilities in Japan. Since April 1974, more than 2600 operators and 1000 shift teams have been trained with the full-scope simulators in BTC up to the end of March 1988. To get the satisfactory results of the training, BTC has been making every effort to improve the facilities, the training materials, the instruction methods and the curricula. In this paper, such a series of recent improvements in the instruction methods and the curricula are presented that are effective to expand the knowledge and to improve the skills of middle or senior class operators. (author)

The prospects for the world nuclear energy market

International Nuclear Information System (INIS)

Anon.

1976-01-01

Over the last few years projections of nuclear power generating capacity growth for the next two decades have progressively decreased. Dwindling load growth, increasing load lead time, costs of delays and high cost inflation, industrial recession and fuel cycle delays are discussed as the main causes of the setback. The state of the fuel cycle business in the world market is examined and data are presented and discussed for predicted world supply and demand. Nuclear plans and fuel policies and requirements are then examined for individual countries. (U.K.)

Way to LWR nuclear power stations taking root

International Nuclear Information System (INIS)

Toyota, Masatoshi

1994-01-01

In this report, as for BWR nuclear power stations, the countermeasures to initial troubles, the circumstances of making bold start for the improvement and standardization after the remarkable lowering of the rate of operation, the development of A-BWR aiming at the most excellent BWR plant and the effort to reduce the construction cost are looked back, and the opinion on advancing nuclear energy development hereafter is mentioned. As the initial troubles, the thermal fatigue of reactor feedwater nozzles, the cracking of return water nozzles in control rod drive, the damage of fuel elements and others occurred. The circumstances till starting the project of improvement and standardization, and the investigation by the ad hoc committee and the results are reported. The feasibility study on A-BWR was carried out from July, 1978 to October, 1979. The basic design and the research and development for verification and the optimization design were carried out till December, 1985, as the result, the improvement of the plant performance was obtained. The effort for reducing the construction cost was successful. (K.I.)

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

Energy Technology Data Exchange (ETDEWEB)

Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-10-01

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

Application of gadolinia credit to cask transportation of BWR-STEP3 SFAs

International Nuclear Information System (INIS)

Kikuchi, Tsukasa; Mitsuhashi, Ishi; Ito, Dai-ichiro; Nakamura, Yu

2003-01-01

Instead of the fresh-fuel assumption, the application of gadolinia credit to cask transportation of BWR SFAs is studied. Its efficacy for BWR-STEP2 SFAs had already been estimated. This paper reports on the application of gadolinia credit to cask transportation of BWR-STEP3 SFAs. (author)

The World Nuclear Industry Status Report 2016

International Nuclear Information System (INIS)

Schneider, Mycle; Froggatt, Antony; Hazemann, Julie; Katsuta, Tadahiro; Ramana, M.V.; Fairlie, Ian; Maltini, Fulcieri; Thomas, Steve; Kaaberger, Tomas

2016-07-01

The World Nuclear Industry Status Report 2015 provides a comprehensive overview of nuclear power plant data, including information on operation, production and construction. The WNISR assesses the status of new-build programs in current nuclear countries as well as in potential newcomer countries. Nuclear power generation in the world increased by 1.3%, entirely due to a 31% increase in China. Ten reactors started up in 2015-more than in any other year since 1990-of which eight were in China. Construction on all of them started prior to the Fukushima disaster. Eight construction starts in the world in 2015-to which China contributed six-down from 15 in 2010 of which 10 were in China. No construction starts in the world in the first half of 2016. The number of units under construction is declining for the third year in a row, from 67 reactors at the end of 2013 to 58 by mid-2016, of which 21 are in China. China spent over US$100 billion on renewables in 2015, while investment decisions for six nuclear reactors amounted to US$18 billion. Eight early closure decisions taken in Japan, Sweden, Switzerland, Taiwan and the U.S. Nuclear phase-out announcements in the U.S. (California) and Taiwan. In nine of the 14 building countries all projects are delayed, mostly by several years. Six projects have been listed for over a decade, of which three for over 30 years. China is no exception here, at least 10 of 21 units under construction are delayed. With the exception of United Arab Emirates and Belarus, all potential newcomer countries delayed construction decisions. Chile suspended and Indonesia abandoned nuclear plans. AREVA has accumulated US$11 billion in losses over the past five years. French government decides euro 5.6 billion bailout and breaks up the company. Share value 95 percent below 2007 peak value. State utility EDF struggles with US$ 41.5 billion debt, downgraded by S and P. Chinese utility CGN, EDF partner for Hinkley Point C, loses 60% of its share value

Data list of nuclear power plants in Japan

International Nuclear Information System (INIS)

Izumi, Fumio; Ito, Noboru; Higuchi, Suminori; Kobayashi, Kensuke; Tobioka, Toshiaki

1987-03-01

The PPD (Nuclear Power Plant Data Base) has been under development in JAERI since 1983 as a six-year program to provide useful information for reactor safety regulation and reactor safety research. Information source of the PPD is mainly based on SAR's (Safety Analysis Reports) of 35 nuclear power plants which are operating, under construction or under licensing review in Japan. The report BWR edition consists of several data lists stored in the PPD, concerning safety design such as performances, equipments and installations of 18 BWR plants in Japan. The informations are based on SAR Attachment Chapter 8 ''Nuclear Reactor Facility Safety Design''. (author)

Decontamination and disposal of radioactive wastes from nuclear facilities

International Nuclear Information System (INIS)

Dlouhy, Z.

1978-01-01

A survey and characteristics are given of the main sources of wastes from the operation of nuclear installations. The amounts are compared of liquid and gaseous wastes from PWR and BWR reactors. The main trends of radioactive waste processing in the world are described. In Czechoslovakia, two methods of waste fixation have been developed: vacuum cementation and bituminization. The demands are summed up on radioactive waste storage sites and it is stated that there are a number of suitable localities, namely abolished granite quarries with a very deep ground water level and a low-permeable overburden and exhausted quarries of kaolinitic clays, which meet all criteria and secure the safe disposal of wastes from Czechoslovak nuclear power plants up to the year 2020. (Z.M.)

Progress to a nuclear-weapon-free world through tactical nuclear arms control

International Nuclear Information System (INIS)

Matseiko, Youri

1999-01-01

At a time when, after some years of passivity, nuclear disarmament is becoming more of an urgent item on the international agenda and receiving accordingly more attention on the part of politicians and non-governmental experts. This is partly reflected in the Report of the Canberra Commission, and the statement on nuclear weapons by international generals and admirals. At the same time some developments such as uncertainties with START II ratification and the process of NATO enlargement make the task of nuclear arms control even more demanding. What is needed now is to pursue at last without any further delay negotiations on effective measures relating to the cessation of the nuclear arms race and to nuclear disarmament. And these effective measures must include both strategic and tactical nuclear weapons. Only a sustained commitment at the highest political level will legitimate serious discussions of the elimination option and ensure that resources and personnel are devoted to finding solutions to the problems associated with moving to zero, and to crafting appropriate transition strategies. In the absence of such a commitment, the nations of the world may never reach the point at which the desirability and feasibility of a nuclear-free world can be evaluated with greater certainty. This Pugwash Conference is trying to make a modest contribution in helping to make possible such a vitally important commitment

The bomb, the dark side of the nuclear world

International Nuclear Information System (INIS)

Collin, J.M.

2009-01-01

The reality of the nuclear world can be summarized in few words: a world arsenal of 26000 bombs, enough uranium and plutonium and the know-how to make much more, multiple non-proliferation and weapons limitation treaties which have troubles regulating the diffusion of this technology, indelible environmental and sanitary marks left by 2059 tests, governments secretly wishing to assume this supreme power attribute, a black market, spies and dealers, but also: opponents, political leaders, local representatives and non-governmental organizations who militate for a nuclear weapon-free world. However, this burning question paradoxically remains obscure to citizens and its obscure aspect is relayed by media and politicians. This book aims at decoding the wheels of the international nuclear weapons situation: from the five official nuclear powers to the proliferation actors, from the defense policies to the risks of accidents and the stakes of disarmament, from the bomb fabrication to its devastating effects. It shows how this ultimate weapon has durably pervaded the defense policies and strategies of countries who own it, and how difficult it will be to reconsider this situation

From the nuclear stalemate to a nuclear-weapon free world. In memory of Klaus Fuchs

International Nuclear Information System (INIS)

Flach, Guenter; Fuchs-Kittowski, Klaus

2012-01-01

The following topics were dealt with: The first soviet atomic bomb and Klaus Fuchs, in illusory worlds of Andrei Sakharov, Edward Teller, and Klaus Fuchs, Klaus Fuchs as grandfather of the hydrogen bomb, memories of and thinking about Klaus Fuchs, the Scottish years of Klaus Fuchs 1937-1941, Klaus Fuchs in the mirror of the Venona documents, Gernot Zippe and the ultracentrifuge or east-west technology transfer in the cold war, secret impulses for the soviet nuclear project, responsibility of knowledge with anti-facism, philosophy, and science as well as peace as the first human right in the work of Klaus Fuchs, the request of Klaus Fuchs for a lasting peace, Klaus Fuchs in Daniel Granin's roman ''Escape to Russia'', ways to a nuclear-weapon free world, Otto Hahn and the declarations of Mainau and Goettingen, nuclear winter, initiatives of the GDR for the prohibition of weapons of mass destruction, nuclear weapons in negative entropy, militarism and antimilitarism of the nuclear age, contributions of the young Klaus Fuchs to statistical physics, nuclear disarmament and the peaceful use of nuclear energy, the responsibility of the scientists for a socially effective and efficient energy change, Berlin-Bucher contributions to a world free of biological weapons. (HSI)

Trouble at the world's nuclear dustbin

International Nuclear Information System (INIS)

Adkins, J.

1984-01-01

Radioactive discharges from the Windscale nuclear fuel reprocessing plant in England's lake district temporarily closed 15 miles of shoreline and continues to raise safety questions in this recreational area. The plant receives high-level radioactive wastes and spent fuel from 36 power plants around the world for reprocessing. The site is also a storage point for 1550 tons of oxide waste waiting for additional reprocessing capacity. Pipelines carry 2.2 million gallons of low-level wastes into the Irish Sea each day. Five hundred pounds of weapons grade plutonium also entered the sea from a World War II munitions depot. Accidents have also contributed to the radioactive debris that has accumulated on sandy beaches. Pressure from Greenpeace and the Barrow Action Group helped to expedite an extensive cleanup program, but activity on the beaches is still highly restricted. British Nuclear Fuels remains undaunted by its negative public relations problems. 4 figures

State of the art of second international exercise on benchmarks in BWR reactors

International Nuclear Information System (INIS)

Verdu, G.; Munoz-Cobo, J. L.; Palomo, M. J.; Escriva, A.; Ginestar, D.

1998-01-01

This is a second in series of Benchmarks based on data from operating Swedish BWRs. The first one concerned measurements made in cycles 14,15 16 and 17 at Ringhals 1 Nuclear Power Plant and addressed predictive power of analytical tools used in BWR stability analysis. Part of the data was disclosed only after participants had provided their results. This work has been published in the report: NEA/NSC/DOC(96)22, November 1996. In this report it was recognised that there is a need for better qualification of the applied noise analysis methods. A follow up Benchmark was thus proposed dedicated to the analysis of time series data and including the evaluation of both global and regional stability of Forsmarks 1 and 2 Nuclear Power Plant. In this second Benchmark have participated Forsmarks Kraftgrupp AB,NEA Nuclear Science Committee, CSN Consejo de Seguridad Nuclear and Department of Chemical and Nuclear Engineering of Polytechnic University of Valencia. (Author)

1984 availability of the world's nuclear power plants

International Nuclear Information System (INIS)

Szeless, A.; Oszuszky, F.

1985-01-01

This survey of the availability of the world's nuclear power plants in 1984 coveres 250 units (the CMEA countries excluded) with an aggregate 184,500 MWe, which are arranged by types of reactor and geographic distribution. The utilization of nuclear power plant capacity attained an average of 66% in 1984, which is an increase by 3 percentage points in the utilization of capacity over the previous year's level (63%). Capacity utilization in pressurized water reactors 1984 (69%) was 4 percentage points higher than it was in boiling water reactors (65%). The ranking list of the world's nuclear power plants is headed by one heavy water reactor and one gas cooled reactor each (98%), followed by five generating units, i.e., two pressurized water reactors, two boiling water reactors, and one heavy water reactor (all 95%). The best German nuclear power plants were Grafenrheinfeld (89%, position No. 20) and Unterweser and Stade (88%, position No. 25). (orig.) [de

Nuclear Power Reactors in the World. 2013 Ed

International Nuclear Information System (INIS)

2013-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to nuclear power reactors in IAEA Member States. This thirty-third edition of Reference Data Series No. 2 provides a detailed comparison of various statistics through 31 December 2012. The tables and figures contain the following information: - General statistics on nuclear reactors in IAEA Member States; - Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; - Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA's Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. The IAEA collects data through designated national correspondents in Member States

Nuclear physics accelerator facilities of the world

International Nuclear Information System (INIS)

1991-12-01

this report is intended to provide a convenient summary of the world's major nuclear physics accelerator facility with emphasis on those facilities supported by the US Department of Energy (DOE). Previous editions of this report have contained only DOE facilities. However, as the extent of global collaborations in nuclear physics grows, gathering summary information on the world's nuclear physics accelerator facilities in one place is useful. Therefore, the present report adds facilities operated by the National Science Foundation (NSF) as well as the leading foreign facilities, with emphasis on foreign facilities that have significant outside user programs. The principal motivation for building and operating these facilities is, of course, basic research in nuclear physics. The scientific objectives for this research were recently reviewed by the DOE/NSF Nuclear Science Advisory Committee, who developed a long range plan, Nuclei, Nucleons, and Quarks -- Nuclear Science in the 1990's. Their report begins as follows: The central thrust of nuclear science is the study of strongly interacting matter and of the forces that govern its structure and dynamics; this agenda ranges from large- scale collective nuclear behavior through the motions of individual nucleons and mesons, atomic nuclei, to the underlying distribution of quarks and gluons. It extends to conditions at the extremes of temperature and density which are of significance to astrophysics and cosmology and are conducive to the creation of new forms of strongly interacting matter; and another important focus is on the study of the electroweak force, which plays an important role in nuclear stability, and on precision tests of fundamental interactions. The present report provides brief descriptions of the accelerator facilities available for carrying out this agenda and their research programs

Development of new irradiation facility for BWR safety research

International Nuclear Information System (INIS)

Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

2013-01-01

In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

Development of NUFREQ-N, an analytical model for the stability analysis of nuclear coupled density-wave oscillations in boiling water nuclear reactors

International Nuclear Information System (INIS)

Park, G.C.

1983-01-01

A state-of-the-art one-dimensional thermal-hydraulic model has been developed to be used for the linear analysis of nuclear-coupled density-wave oscillations in a boiling water nuclear reactor (BWR). The model accounts for phasic slip, distributed spacers, subcooled boiling, space/time-dependent power distributions and distributed heated wall dynamics. In addition to a parallel channel stability analysis, a detailed model was derived for the BWR loop analysis of both the natural and forced circulation modes of operation. In its final form, this model constitutes a multi-input, multi-output (MIMO) linear system, which features a general nodal neutron kinetics model. Kinetics parameters for use in the kinetics model have been obtained by utilizing self-consistent nodal data and power distributions. The stability characteristics of a typical BWR/4 has been investigated with the Nyquist criterion. The computer implementation of this mode, NUFREQ-N, was used for the parametric study of a typical BWR/4 and comparison were made with existing in-core and out-of-core data. Also, NUFREQ-N was used to analyze the expected stability characteristics of a typical BWR/4. The parametric results revealed important factors influencing BWR stability margin. It was found that NUFREQ-N generally agreed well with out-of-core data. This was especially true for the predicted power-to-flow transfer function, which is the most important transfer function in thermal-hydraulic stability analysis

Study of transient rod extraction failure without RBM in a BWR

International Nuclear Information System (INIS)

Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

2015-09-01

The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

BWR type reactor

International Nuclear Information System (INIS)

Okano, Shigeru.

1992-01-01

In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

Preliminary study on characteristics of equilibrium thorium fuel cycle of BWR

International Nuclear Information System (INIS)

Waris, A.; Kurniadi, R.; Su'ud, Z.; Permana, S.

2007-01-01

One of the main objectives behind the transuranium recycling ideas is not merely to utilize natural resource that is uranium much more efficiently, but to reduce the environmental impact of the radio-toxicity of the nuclear spent fuel. Beside uranium resource, there is thorium which has three times abundance compared to that of uranium which can be utilized as nuclear fuel. On top of that thorium is believed to have less radio-toxicity of spent fuel since its produce smaller amount of higher actinides compared to that of uranium. However, the studies on the thorium utilization in nuclear reactor in particular in light water reactors (LWR) are not performed intensively yet. Therefore, the aim of the present study is to evaluate the characteristics of thorium fuel cycle in LWR, especially boiling water reactor (BWR). To conduct the comprehensive investigations we have employed the equilibrium burnup model (1-3). The equilibrium burnup model is an alternative powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor1). We have employed 1368 nuclides in the equilibrium burnup calculation where 129 of them are heavy metals (HMs). This burnup code then is coupled with SRAC cell calculation code by using PIJ module to compose an equilibrium-cell burnup code. For cell calculation, 26 HMs, 66 fission products (FPs) and one pseudo FP have been utilized. The JENDL 3.2 library has been used in this study. References: 1. A. Waris and H. Sekimoto, 'Characteristics of several equilibrium fuel cycles of PWR', J. Nucl. Sci. Technol., 38, p.517-526, 2001 2. A. Waris, H. Sekimoto, and G. Kastchiev, Influence of Moderator-to-Fuel Volume Ratio on Pu and MA Recycling in Equilibrium Fuel Cycles of

A world class nuclear research reactor complex for South Africa's nuclear future

International Nuclear Information System (INIS)

Keshaw, Jeetesh

2008-01-01

South Africa recently made public its rather ambitious goals pertaining to nuclear energy developments in a Draft Policy and Strategy issued for public comment. Not much attention was given to an important tool for nuclear energy research and development, namely a well equipped and maintained research reactor, which on its own does not do justice to its potential, unless it is fitted with all the ancillaries and human resources as most first world countries have. In South Africa's case it is suggested to establish at least one Nuclear Energy Research and Development Centre at such a research reactor, where almost all nuclear energy related research can be carried out on par with some of the best in the world. The purpose of this work is to propose how this could be done, and motivate why it is important that it be done with great urgency, and with full involvement of young professionals, if South Africa wishes to face up to the challenges mentioned in the Draft Strategy and Policy. (authors)

Analysis CFD for the hydrogen transport in the primary containment of a BWR

International Nuclear Information System (INIS)

Jimenez P, D. A.; Del Valle G, E.; Gomez T, A. M.

2014-10-01

This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

The World Nuclear Industry Status Report 2010-2011. Nuclear Power in a Post-Fukushima World. 25 years after the Chernobyl accident

International Nuclear Information System (INIS)

Schneider, Mycle; Froggatt, Antony; Thomas, Steve; Hazemann, Julie; Mastny, Lisa

2011-04-01

The report provides the reader with the basic quantitative and qualitative facts about nuclear power plants in operation, under construction, and in planning phases throughout the world. It assesses the economic performance of past and current nuclear projects and compares their development to that of leading renewable energy sources. An extensive annex provides a country-by-country analysis of nuclear programs around the world. The report also includes the first published overview of reactions to the catastrophe in Japan. But developments even prior to March 11, when the Fukushima crisis began, illustrate that the international nuclear industry has been unable to stop the slow decline of nuclear energy. Not enough new units are coming online, and the world's reactor fleet is aging quickly. Moreover, it is now evident that nuclear power development cannot keep up with the pace of its renewable energy competitors. Annual renewables capacity additions have been outpacing nuclear start-ups for 15 years. In the United States, the share of renewables in new capacity additions skyrocketed from 2 percent in 2004 to 55 percent in 2009, with no new nuclear coming on line. In 2010, for the first time, worldwide cumulated installed capacity of wind turbines (193 GW), small hydro (80 GW, excluding large hydro) biomass and waste-to-energy plants (65 GW), and solar power (43 GW) reached 381 GW, outpacing the installed nuclear capacity of 375 GW prior to the Fukushima disaster. Total investment in renewable energy technologies has been estimated at $243 billion in 2010. As of April 1, 2011, there were 437 nuclear reactors operating in the world-seven fewer than in 2002. The International Atomic Energy Agency (IAEA) currently lists 64 reactors as 'under construction' in 14 countries. By comparison, at the peak of the industry's growth phase in 1979, there were 233 reactors being built concurrently. In 2008, for the first time since the beginning of the nuclear age, no new unit was

The new competition in the world market for nuclear reactors

International Nuclear Information System (INIS)

Finon, Dominique

2014-01-01

The current revival in the world market for nuclear reactors, notwithstanding Fukushima, completes the re-composition of the world's nuclear industry that started in the early 1990's and which has displaced nuclear power's centre of gravity towards Asia. In this new context, the capability to provide full-fledged financing for the buyers and to set up consortia that may include the operator have become major advantages at this stage, relegating to a lower order the ability to supply reactors with a high level of safety. (author)

Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen; Estrategias de mitigacion de la corrosion intergranular en sistemas de reactores de agua en ebullicion (BWR). Accion combinada de la quimica del hidrogeno y del oxigeno

Energy Technology Data Exchange (ETDEWEB)

Verdugo, M.

2015-07-01

Inter-Granular Stress Corrosion cracking (IGSCC) in austenitic stainless steel and in austenitic nickel-based alloys has been the subject of many studies the aim of which was to resolve one of the main problems faced by BWR nuclear power plants since the 1960s. This corrosion phenomenon is the result of the combined action of three factors: sensitization of the material, high local stresses and an aggressive medium. This paper deals with these factors separately and analyzes the oxidative chemistry of BWR reactors (aggressivity of the medium) as one the main causes if IGSCC. (Author)

General model for Pc-based simulation of PWR and BWR plant components

Energy Technology Data Exchange (ETDEWEB)

Ratemi, W M; Abomustafa, A M [Faculty of enginnering, alfateh univerity Tripoli, (Libyan Arab Jamahiriya)

1995-10-01

In this paper, we present a basic mathematical model derived from physical principles to suit the simulation of PWR-components such as pressurizer, intact steam generator, ruptured steam generator, and the reactor component of a BWR-plant. In our development, we produced an NMMS-package for nuclear modular modelling simulation. Such package is installed on a personal computer and it is designed to be user friendly through color graphics windows interfacing. The package works under three environments, namely, pre-processor, simulation, and post-processor. Our analysis of results using cross graphing technique for steam generator tube rupture (SGTR) accident, yielded a new proposal for on-line monitoring of control strategy of SGTR-accident for nuclear or conventional power plant. 4 figs.

Interpretation of incore noise measurements in BWR's

International Nuclear Information System (INIS)

Dam, H. van

1982-01-01

A survey is given of the main incentives for power reactor noise research and the differences and similarities of noise in power and zero power systems are touched on. The basic characteristics of the adjoint method in reactor noise theory are treated. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurement of the reactor transfer function, which is demonstrated by results from measurements on a BWR in the Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

Interpretation of incore noise measurements in BWR's

International Nuclear Information System (INIS)

Dam, H. van

1983-01-01

A survey is given of the main incentives for power reactor noise research, and the differences and similarities of noise in power and zero power systems are shown. After a short outline of historical developments the basic characteristics of the adjoint method in reactor noise theory are dealt with. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies, which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurements on a BWR in The Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

International Nuclear Information System (INIS)

Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

2012-01-01

The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

Radiation buildup and control in BWR recirculation piping

International Nuclear Information System (INIS)

Meyer, W.; Wood, R.M.; Rao, T.V.; Vook, R.W.

1987-01-01

Boiling water nuclear reactors (BWRs) employ stainless steel (Types 304 or 316 NG) pipes in which high-purity water at temperatures of ∼ 275 0 C are circulated. Various components of the system, such as valves and bearings, often contain hard facing metal alloys such as Stellite-6. These components, along with the stainless steel tubing and feedwater, serve as sources of 59 Co. This cobalt, along with other soluble and insoluble impurities, is carried along with the circulating water to the reactor core where it is converted to radioactive 60 Co. After reentering the circulating water, the 60 Co can be incorporated into a complex corrosion layer in the form of CoCr 2 O 4 and/or CoFe 2 O 4 . The presence of even small amounts of 60 Co on the walls of BWR cooling systems is the dominant contributor to inplant radiation levels. Thus BWR owners and their agents are expending significant time and resources in efforts to reduce both the rate and amount of 60 Co buildup. The object of this research is twofold: (a) to form a thin diffusion barrier against the outward migration of cobalt from a cobalt-containing surface and (b) to prevent the growth of a 60 Co-containing corrosion film. The latter goal was the more important since most of the radioactive cobalt will originate from sources other than the stainless steel piping itself

Status and Trends of Nuclear Power World-wide

International Nuclear Information System (INIS)

Gueorguiev, B.; Spiegelberg-Planer, R.

1996-01-01

The reliable and adequate supply of energy, especially electricity, is necessary not only for economic development but to enhance the quality of life. Nuclear power is a proven technology which already supplies about 17% of the world''s electricity generation. In 1995, seven countries produce more than 40% of their electricity from nuclear power plants: Lithuanian, France, Belgium, Sweden, Bulgaria, Slovak and Hungary. It is quite clear that many countries are heavily reliant on nuclear power and are well beyond the point where nuclear power could be replaced by some other source, so, nuclear power remains one of the few technologically proven, economically promising and environmentally benign energy sources. An important factor in the continued development of nuclear power is the extent to which nuclear generated electricity remains economically competitive. Factors such as plant availability, standardisation of systems, components and equipment, as well as the cost of equipment to meet safety and environmental regulations play also an important role in determining the relative competitiveness of nuclear power plants. Many operating organizations have already impressive results in the reduction of plant unavailability. The number of nuclear power plants currently operating with annual availability factor exceeding 85% is increasing. Good performance of some operators should establish performance targets for operators everywhere. The International Atomic Energy Agency (IAEA) has the only international and almost complete information system, the Power Reactor Information System (PRIS) with nuclear power plant status and performance data. This paper presents the current status of nuclear power plants, according to information contained in the IAEA. It discusses the plant performance indicators available in PRIS and the improvement trend in the performance of nuclear power plants based on these indicators. It also presents the future trends of nuclear power focusing

Summary of nuclear fuel reprocessing activities around the world

International Nuclear Information System (INIS)

Mellinger, P.J.; Harmon, K.M.; Lakey, L.T.

1984-11-01

This review of international practices for nuclear fuel reprocessing was prepared to provide a nontechnical summary of the current status of nuclear fuel reprocessing activities around the world. The sources of information are widely varied

U.S. Department Of Energy's nuclear engineering education research: highlights of recent and current research-II. 7. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

International Nuclear Information System (INIS)

Edwards, Robert; Huang, Zhengyu

2001-01-01

The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been recently expanded for BWR out-of-phase behavior. Out-of-phase oscillation is a phenomenon that occurs at BWRs. During this kind of event, half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. The HRS will be used for development and validation of stability monitoring and control techniques as part of an ongoing U.S. Department of Energy Nuclear Engineering Education and Research grant. The Penn State TRIGA reactor is used to simulate BWR fundamental mode power dynamics. The first harmonic mode power, together with detailed thermal hydraulics of boiling channels of both fundamental mode and first harmonic mode, is simulated digitally in real time with a computer. Simulations of boiling channels provide reactivity feedback to the TRIGA reactor, and the TRIGA reactor's power response is in turn fed into the channel simulations and the first harmonic mode power simulation. The combination of reactor power response and the simulated first harmonic power response with spatial distribution functions thus mimics the stability phenomena actually encountered in BWRs. The digital simulations of the boiling channels are performed by solving conservation equations for different regions in the channel with C-MEX S-functions. A fast three-dimensional (3-D) reactor power display of modal BWR power distribution was implemented using MATLAB graphics capability. Fundamental mode, first harmonic, together with the total power distribution over the reactor cross section, are displayed. Because of the large amount of computation for BWR boiling channel simulation and real-time data processing and graph generation, one computer is not sufficient to handle these jobs in the hybrid reactor simulation environment. A new three-computer setup has been

Nuclear power : world and Australia - a long-term view

Energy Technology Data Exchange (ETDEWEB)

Ford, G W.K.

1989-01-01

Developments in world and Australian activities relating to nuclear power and the nuclear fuel cycle are reviewed. Main issues addressed include environment, energy sources, uranium mining, enrichment, reactor design, fuel reprocessing and waste disposal. The benefits for Australia through its involvement in all stages of the nuclear fuel cycle are also discussed.

Nuclear energy

International Nuclear Information System (INIS)

Hesketh, Ross.

1985-01-01

The subject is treated under the headings: nuclear energy -what is it; fusion (principles; practice); fission (principles); reactor types and systems (fast (neutron) reactors as breeders; fast reactors; thermal reactors; graphite-moderated thermal reactors; the CANDU reactor; light water reactors - the BWR and the PWR); the nuclear fuel cycle (waste storage; fuel element manufacture; enrichment processes; uranium mining); safety and risk assessment; the nuclear power industry and the economy (regulating authorities; economics; advantages and disadvantages). (U.K.)

Pre-study of dynamic loads on the internals caused by a large pipe break in a BWR; Foerstudie av stroemningsinducerade laster paa interndelar vid brott i huvudcirkulationskretsarna i BWR

Energy Technology Data Exchange (ETDEWEB)

Marcinkiewicz, Jerzy; Lindgren, Anders [Det Norske Veritas Nuclear Technology AB, Stockholm (Sweden)

2002-12-01

Det Norske Veritas Nuclear Technology has performed a literature study of dynamic load on a BWR (Boiling Water Reactor) internals caused by a large pipe break. The goal of the study was to improve the knowledge about the physics of phenomena occurring in the RPV (Reactor Pressure Vessel) after pipe break in the main circulation system and also to make a review of calculation methods, models and computer programs including their capabilities when calculating the dynamic loads. The report presents description of relevant parts of a BWR, initial and boundary conditions, and phenomena determining the loads - rapid depressurization and propagation of pressure wave (including none-equilibrium). Furthermore, the report generally describes possible methodologies for calculating the dynamic loads on internals after the pipe break and the experiences from calculations the dynamic loads with different methods (computer programs) including comparisons with experimental data. Fluid-Structure Interaction methodology and its importance for calculation of dynamic loads on reactor internals is discussed based on experimental data. A very intensive research program for studying and calculating the dynamic loads on internals after pipe breaks has been performed in USA and Germany during the seventies and the eighties. Several computer programs have been developed and a number of large-scale experiments have been performed to calibrate the calculation methods. In spite of the fact that all experiments were performed for PWR several experiences should be valid also for BWR. These experiences, connected mainly to capabilities of computer programs calculating dynamic loads, are discussed in the report.

Nuclear astrophysics of worlds in the string landscape

International Nuclear Information System (INIS)

Hogan, Craig J.

2006-01-01

Motivated by landscape models in string theory, cosmic nuclear evolution is analyzed allowing the standard model Higgs expectation value w to take values different from that in our world (w≡1), while holding the Yukawa couplings fixed. Thresholds are estimated, and astrophysical consequences are described, for several sensitive dependences of nuclear behavior on w. The dependence of the neutron-proton mass difference on w is estimated based on recent calculations of strong isospin symmetry breaking, and is used to derive the threshold of neutron-stable worlds, w≅0.6±0.2. The effect of a stable neutron on nuclear evolution in the big bang and stars is shown to lead to radical differences from our world, such as a predominance of heavy r-process and s-process nuclei and a lack of normal galaxies, stars, and planets. Rough estimates are reviewed of w thresholds for deuteron stability and the pp and pep reactions dominant in many stars. A simple model of nuclear resonances is used to estimate the w dependence of overall carbon and oxygen production during normal stellar nucleosynthesis; carbon production is estimated to change by a fraction ≅15(1-w). Radical changes in astrophysical behavior seem to require changes in w of more than a few percent, even for the most sensitive phenomena

BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

International Nuclear Information System (INIS)

McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

1986-02-01

This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

Transmutation of minor actinide using thorium fueled BWR core

International Nuclear Information System (INIS)

Susilo, Jati

2002-01-01

One of the methods to conduct transmutation of minor actinide is the use of BWR with thorium fuel. Thorium fuel has a specific behaviour of producing a little secondary minor actinides. Transmutation of minor actinide is done by loading it in the BWR with thorium fuel through two methods, namely close recycle and accumulation recycle. The calculation of minor actinide composition produced, weigh of minor actinide transmuted, and percentage of reminder transmutation was carried SRAC. The calculations were done to equivalent cell modeling from one fuel rod of BWR. The results show that minor actinide transmutation is more effective using thorium fuel than uranium fuel, through both close recycle and accumulation recycle. Minor actinide transmutation weight show that the same value for those recycle for 5th recycle. And most of all minor actinide produced from 5 unit BWR uranium fuel can transmuted in the 6 t h of close recycle. And, the minimal value of excess reactivity of the core is 12,15 % Δk/k, that is possible value for core operation

Decay ratio studies in BWR and PWR using wavelet

International Nuclear Information System (INIS)

Ciftcioglu, Oe.

1996-10-01

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

World electricity generation, nuclear power, and oil markets

International Nuclear Information System (INIS)

1990-01-01

Striking changes have characterized the world's production and use of energy over the past 15 years. Most prominent have been the wide price fluctuations, politicization of world oil prices and supply, along with profound changes in patterns of production and consumption. This report, based on a study by energy analysts at Science Concepts, Inc., in the United States, traces changes in world energy supply since 1973-74 - the time of the first oil ''price shocks''. In so doing, it identifies important lessons for the future. The study focused in particular on the role of the electric power sector because the growth in fuel use in it has been accomplished without oil. Instead, the growth has directly displaced oil. In the pre-1973 era, the world relied increasingly on oil for many energy applications, including the production of electricity. By 1973, more than on-fourth of the world's electricity was produced by burning oil. By 1987, however, despite a large increase in electric demand, the use of oil was reigned back to generating less than 10% of the world's electricity. Nuclear power played a major role in this turnaround. From 1973-87, analysts at Science Concepts found, nuclear power displaced the burning of 11.7 billion barrels of oil world-wide and avoided US $323 billion in oil purchases

Experimental investigation of control absorber blade effects in a modern 10x10 BWR assembly

Energy Technology Data Exchange (ETDEWEB)

Jatuff, F.; Grimm, P.; Murphy, M.; Luethi, A.; Seiler, R.; Joneja, O.; Meister, A.; Geemert, R. van; Brogli, R.; Chawla, R. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Williams, T. [EGL Laufenburg (Switzerland); Helmersson, S. [Westinghouse Atom (Sweden)

2001-03-01

The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades. In Boiling Water Reactors (BWR) is important for safety assessment, and for achieving a flexible operation during the cycle. Characteristics which are affected strongly include the power distribution for controlled core regions and its impact on linear heat generation rate margins, as well as the build-up of plutonium, and its influence on core excess reactivity and the reactivity worth of the shutdown system. PSI and the Swiss Nuclear Utilities (UAK) are conducting an experimental reactor physics programme related to modern Light Water Reactor (LWR) fuel assemblies, as employed in the Swiss nuclear power plants: the so-called. LWR-PROTEUS Phase I project. A significant part of this project has been devoted to the characterization of highly heterogeneous BWR fuel elements in the presence of absorber blades. The paper presents typical results for the performance of modern lattice codes in the estimation of controlled assembly reaction rate distributions, the sensitivity to the geometrical and material characterization, and a preliminary comparison of reflected-test-zone calculations with experimental reaction rate distributions measured in a Westinghouse SVEA-96+ assembly under full-density water moderation conditions in the presence of Westinghouse boron-carbide absorber blades. (author)

Experimental investigation of control absorber blade effects in a modern 10x10 BWR assembly

International Nuclear Information System (INIS)

Jatuff, F.; Grimm, P.; Murphy, M.; Luethi, A.; Seiler, R.; Joneja, O.; Meister, A.; Geemert, R. van; Brogli, R.; Chawla, R.; Williams, T.; Helmersson, S.

2001-01-01

The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades. In Boiling Water Reactors (BWR) is important for safety assessment, and for achieving a flexible operation during the cycle. Characteristics which are affected strongly include the power distribution for controlled core regions and its impact on linear heat generation rate margins, as well as the build-up of plutonium, and its influence on core excess reactivity and the reactivity worth of the shutdown system. PSI and the Swiss Nuclear Utilities (UAK) are conducting an experimental reactor physics programme related to modern Light Water Reactor (LWR) fuel assemblies, as employed in the Swiss nuclear power plants: the so-called. LWR-PROTEUS Phase I project. A significant part of this project has been devoted to the characterization of highly heterogeneous BWR fuel elements in the presence of absorber blades. The paper presents typical results for the performance of modern lattice codes in the estimation of controlled assembly reaction rate distributions, the sensitivity to the geometrical and material characterization, and a preliminary comparison of reflected-test-zone calculations with experimental reaction rate distributions measured in a Westinghouse SVEA-96+ assembly under full-density water moderation conditions in the presence of Westinghouse boron-carbide absorber blades. (author)

BWR radiation exposure--experience and projection

International Nuclear Information System (INIS)

Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

1979-01-01

The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

The BWR Hybrid 4 control rod

International Nuclear Information System (INIS)

Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

1988-01-01

The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

Energy Technology Data Exchange (ETDEWEB)

Moranchel y R, M.

2012-07-01

Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

Advanced technology for BWR operator training simulator

International Nuclear Information System (INIS)

Shibuya, Akira; Fujita, Eimitsu; Nakao, Toshihiko; Nakabaru, Mitsugu; Asaoka, Kouchi.

1991-01-01

This paper describes an operator training simulator for BWR nuclear power plants which went into service recently. The simulator is a full scope replica type simulator which faithfully replicates the control room environment of the reference plant with six main control panels and twelve auxiliary ones. In comparison with earlier simulators, the scope of the simulation is significantly extended in both width and depth. The simulation model is also refined in order to include operator training according to sympton-based emergency procedure guidelines to mitigate the results in accident cases. In particular, the core model and the calculational model of the radiation intensity distribution, if radioactive materials were released, are improved. As for simulator control capabilities by which efficient and effective training can be achieved, various advanced designs are adopted allowing easy use of the simulators. (author)

Technology developments for Japanese BWR MOX fuel utilization

International Nuclear Information System (INIS)

Oguma, M.; Mochida, T.; Nomata, T.; Asahi, K.

1997-01-01

The Long-Term Program for Research, Development and Utilization of Nuclear Energy established by the Atomic Energy Commission of Japan asserts that Japan will promote systematic utilization of MOX fuel in LWRs. Based on this Japanese nuclear energy policy, we have been pushing development of MOX fuel technology aimed at future full scale utilization of this fuel in BWRs. In this paper, the main R and D topics are described from three subject areas, MOX core and fuel design, MOX fuel irradiation behaviour, and MOX fuel fabrication technology. For the first area, we explain the compatibility of MOX fuel with UO 2 core, the feasibility of the full MOX core, and the adaptability of MOX design methods based on a mock-up criticality experiment. In the second, we outline the Tsuruga MOX irradiation program and the DOMO program, and suggest that MOX fuel behaviour is comparable to ordinary BWR UO 2 fuel behaviour. In the third, we examine the development of a fully automated MOX bundle assembling apparatus and its features. (author). 14 refs, 11 figs, 3 tabs

A nonlinear 3D real-time model for simulation of BWR nuclear power plants

International Nuclear Information System (INIS)

Ercan, Y.

1982-02-01

A nonlinear transient model for BWR nuclear power plants which consists of a 3D-core (subdivided into a number of superboxes, and with parallel flow and subcooled boiling), a top plenum, steam removal and feed water systems and main coolant recirculation pumps is given. The model describes the local core and global plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. The model which forms the basis for the digital code GARLIC-B (Er et al. 82) is aimed to be used on an on-site process computer in parallel to the actual reactor process (or even in predictive mode). Thus, special measures had to be taken into account in order to increase the computational speed and reduce the necessary computer storage. This could be achieved by - separating the neutron and power kinetics from the xenon-iodine dynamics, - treating the neutron kinetics and most of the thermodynamics and hydrodynamics in a pseudostationary way, - developing a special coupling coefficient concept to describe the neutron diffusion, calculating the coupling coefficients from a basic neutron kinetics code, - combining coarse mesh elements into superboxes, taking advantage of the symmetry properties of the core and - applying a sparse matrix technique for solving the resulting algebraic power equation system. (orig.) [de

Turbine protecting device in a BWR type nuclear power plant

International Nuclear Information System (INIS)

Kasuga, Hajime; Oka, Yoko.

1984-01-01

Purpose: To prevent highly humid steams from flowing into the turbine upon abnormal reduction in the reactor water level in order to ensure the turbine soundness, as well as in order to trip the turbine with no undesired effect on the reactor. Constitution: A protection device comprising a judging device and a timer are disposed in a BWR type reactor, in order to control a water level signal from a reactor water level gage. If the reactor water level is reduced during rated power operation, steams are kept to be generated due to decay heat although reactor is scramed. When a signal from the reactor water level detector is inputted to the protection device, a trip signal is outputted by way of a judging device after 15 second by means of the timer, when the main steam check valve is closed to trip the turbine. With this delay of time, abrupt increase in the pressure of the reactor due to sudden shutdown can be prevented. (Nakamoto, H)

Cooperative Security: A New Paradigm For A World Without Nuclear Weapons?

Directory of Open Access Journals (Sweden)

Marc Finaud

2013-11-01

Full Text Available If there is a loose consensus on aiming at a world free of nuclear weapons in the future, there are clear oppositions as to the timeframe as well as the means for achieving this goal. The approach to nuclear disarmament followed to date has only yielded limited success because it has been conceived in isolation from global and regional security environments and threat perceptions. A new paradigm should thus be sought in order to reconcile nuclear powers’ security doctrines with global aspirations for a safer world, and ensure that nuclear powers derive their security less from others’ insecurity but from mutually beneficial cooperative security. This should not become a pretext for preserving nuclear weapons for ever. It will on the contrary require parallel tracks addressing the initial motivations for acquiring nuclear weapons and other weapons of mass destruction (WMD, in particular in the context of regional conflicts, as well as dealing with the current issues necessarily related to nuclear disarmament (missile defence, weaponization of space, conventional imbalances and future weapon systems. Ultimately, in a globalised nuclear-weapon free world, state security will not require nuclear weapons because it will be inserted into a broader network encompass­ing all aspects of security addressed in cooperative and multilateral approaches.

World-wide termination of nuclear energy application

International Nuclear Information System (INIS)

Quirin, W.

1991-01-01

It is easy to require the widely discussed termination of nuclear energy application, but it is hardly possible to realise it, unless one is prepared to accept enormous economic and ecological problems. The article investigates, whether the other energy carriers or energy saving methods, respectively, would be in a position to replace the nuclear energy. Thereby the aspects of securing the supply and its economy are of considerable importance. The author describes furthermore the effects of terminating nuclear energy on the growing world population and the economy of trading countries. Ecological problems that may also be aggravated are dealt with, too. (orig.) [de

Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Ortiz S, J.J

1998-10-01

In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

Nuclear Power Reactors in the World. 2014 Ed

International Nuclear Information System (INIS)

2014-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to nuclear power reactors in IAEA Member States. This thirty-fourth edition of Reference Data Series No. 2 provides a detailed comparison of various statistics up to and including 31 December 2013. The tables and figures contain the following information: — General statistics on nuclear reactors in IAEA Member States; — Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; — Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. The IAEA collects this data through designated national correspondents in Member States

The World Nuclear Industry Status Report 2017

International Nuclear Information System (INIS)

Schneider, Mycle; Froggatt, Antony; Hazemann, Julie; Katsuta, Tadahiro; Ramana, M.V.; Rodriguez, Juan C.; Ruedinger, Andreas; Stienne, Agnes

2017-09-01

The World Nuclear Industry Status Report 2017 (WNISR2017) provides a comprehensive overview of nuclear power plant data, including information on operation, production and construction. The WNISR assesses the status of new-build programs in current nuclear countries as well as in potential newcomer countries. The WNISR2017 edition includes a new assessment from an equity analyst view of the financial crisis of the nuclear sector and some of its biggest industrial players. The Fukushima Status Report provides not only an update on onsite and offsite issues six years after the beginning of the catastrophe, but also the latest official and new independent cost evaluations of the disaster. Focus chapters provide in-depth analysis of France, Japan, South Korea, the United Kingdom and the United States. The Nuclear Power vs. Renewable Energy chapter provides global comparative data on investment, capacity, and generation from nuclear, wind and solar energy. Finally, Annex 1 presents a country-by-country overview of all other countries operating nuclear power plants

FIST small break accident analysis with BWR TRACBO2-pretest predictions

International Nuclear Information System (INIS)

Alamgir, M.; Sutherland, W.A.

1984-01-01

The BWR Full Integral Simulation Test (FIST) program includes experimental simulation and analytical evaluation of BWR thermal-hydraulic phenomena during transient events. One such event is a small size break in the suction line of one of the recirculation pumps. The results from a test simulating this transient in the FIST facility are compared with a system analysis using the Transient Reactor Analysis Code (TRACB02). This comparison demonstrates BWR-TRAC capability for small break analyses and provides detailed understanding of the phenomena

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.)

Nuclear desalination in the Arab world - Part II: Advanced inherent and passive safe nuclear reactors

International Nuclear Information System (INIS)

Karameldin, A.; Samer S. Mekhemar

2004-01-01

Rapid increases in population levels have led to greater demands for fresh water and electricity in the Arab World. Different types of energies are needed to contribute to bridging the gap between increased demand and production. Increased levels of safeguards in nuclear power plants have became reliable due to their large operational experience, which now exceeds 11,000 years of operation. Thus, the nuclear power industry should be attracting greater attention. World electricity production from nuclear power has risen from 1.7% in 1970 to 17%-20% today. This ratio had increased in June 2002 to reach more than 30%, 33% and 42% in Europe, Japan, and South Korea respectively. In the Arab World, both the public acceptance and economic viability of nuclear power as a major source of energy are greatly dependent on the achievement of a high level of safety and environmental protection. An assessment of the recent generation of advanced reactor safety criteria requirements has been carried out. The promising reactor designs adapted for the Arab world and other similar developing countries are those that profit from the enhanced and passive safety features of the new generation of reactors, with a stronger focus on the effective use of intrinsic characteristics, simplified plant design, and easy construction, operation and maintenance. In addition, selected advanced reactors with a full spectrum from small to large capacities, and from evolutionary to radical types, which have inherent and passive safety features, are discussed. The relevant economic assessment of these reactors adapted for water/electricity cogeneration have been carried out and compared with non-nuclear desalination methods. This assessment indicates that, water/electricity cogeneration by the nuclear method with advanced inherent and passive safe nuclear power plants, is viable and competitive. (author)

Application of tearing modulus stability concepts to nuclear piping. Final report. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Cotter, K.H.; Chang, H.Y.; Zahoor, A.

1982-02-01

The recently developed tearing modulus stability concept was successfully applied to several boiling water reactor (BWR) and pressurized water reactor (PWR) piping systems. Circumferentially oriented through-the-thickness cracks were postulated at numerous locations in each system. For each location, the simplified tearing stability methods developed in USNRC Report NUREG/CR-0838 were used to determine crack stability. The J-T diagram was used to present the results of the computations. The piping systems considered included Type 304 stainless steel as well as A106 carbon steel materials. These systems were analyzed using the piping analysis computer code MINK.

A BWR Safety and Operability Improvements

International Nuclear Information System (INIS)

Sawyer, Craig D.

1993-01-01

The A BWR is the culmination of 30 years of design, development and operating experience of BWRs around the world. It represents across the board improvements is safety, operation and maintenance practices (O and M), economics, radiation exposure and rad waste generation. More than ten years and $20m5 went into the design and development of its new features, and it is now under construction in Japan. This paper concentrates on the safety and operability improvements. In the safety area, more than a decade improvement in core damage frequency (CDFR) has been assessed by formal PIRA techniques, with CDFR less than 10 -6 /year. Severe accident mitigation has also been formally addressed in the design. Plant operations were simplified by incorporation of better materials, optimum use of redundancy in mechanical and electrical equipment so that on-line maintenance can be performed, by better arrangements which account for required maintenance practices, and by an advanced control room

From the nuclear world

International Nuclear Information System (INIS)

Anon.

2016-01-01

This document gathers pieces of information concerning nuclear industry worldwide. The most relevant are the following ones. CGN (China General Nuclear) will launched in 2017 the construction of a prototype of a small transportable modular reactor whose purpose is to produce electricity on a remote place like an island or aboard a boat for long-term missions. The Wylfa reactor (490 MWe) was decommissioned on December 30., 2015. Wylfa was the last Magnox type reactor operating in the world. In France a campaign of information and iodine drug dispatching has been launched for people living near nuclear power plants. The global cost of the CIGEO project whose aim is the disposal of high-level radioactive wastes has been estimated to 25 billions euros including construction costs, operating costs over a 100 year period and dismantling costs. The European Commission has warned France that the financial provisions made for the dismantling of nuclear facilities and the processing of the consequent wastes are not sufficient to cover the future costs. 4 reactors with a power of 1400 MWe each, are being built on the Barakah site in Abu Dhabi, works are on time and the first unit may operate end 2016. Wikileaks has accused AREVA of not taking all necessary measures for the protection of its employees at the Bakouma mine. AREVA denies the charges and affirms that regulations and safety requirements are the same as for its French sites whatever the country. The initiative 'Nuclear for Climate' gathering pro-nuclear associations worldwide, intends to remind the international community that nuclear energy is an important tool to fight climate change. The French site for the disposal of low-level radioactive wastes is facing saturation in the very short term while the volume of such wastes is expected to soar in the next decades as the dismantling programmes will gain in importance. A new policy for the management of such wastes is needed. (A.C.)

Fault tree analysis on BWR core spray system

International Nuclear Information System (INIS)

Watanabe, Norio

1982-06-01

Fault Trees which describe the failure modes for the Core Spray System function in the Browns Ferry Nuclear Plant (BWR 1065MWe) were developed qualitatively and quantitatively. The unavailability for the Core Spray System was estimated to be 1.2 x 10 - 3 /demand. It was found that the miscalibration of four reactor pressure sensors or the failure to open of the two inboard valves (FCV 75-25 and 75-53) could reduce system reliability significantly. It was recommended that the pressure sensors would be calibrated independently. The introduction of the redundant inboard valves could improve the system reliability. Thus this analysis method was verified useful for system analysis. The detailed test and maintenance manual and the informations on the control logic circuits of each active component are necessary for further analysis. (author)

World nuclear fuel cycle requirements 1989

International Nuclear Information System (INIS)

1989-01-01

This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under two nuclear supply scenarios. These two scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries in the World Outside Centrally Planned Economic Areas (WOCA). A No New Orders scenarios is also presented for the Unites States. This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the WOCA projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel; discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2020 for the Lower and Upper Reference cases and through 2036, the last year in which spent fuel is discharged, for the No New Orders case

U.S. congressional attitudes and policies affecting nuclear power development in the world

International Nuclear Information System (INIS)

McCormack, M.

1976-01-01

The world future for nuclear power is even now being formed by policies and decisions of many governments and international organizations. Congressman McCormack looks to the United States for revived and stronger leadership in strengthening the web of institutions and international relations to permit the world to reap the benefits of nuclear power without a destabilizing spread of nuclear weapons. He says Congress will have a major role in shaping that nuclear future. The tensions between Congress and the executive branch that are part of the U.S. system of separation of powers can help to test and strengthen future policy on international nuclear power. The point of no return along the course of nuclear evolution is approaching and the author asks: will we press on to create an acceptable balance between benefits of nuclear power and the risk that expanded use may increase proliferation--or will we turn back toward nuclear isolationism. Mr. McCormack opts for vigorous legislative, executive and diplomatic initiatives to sustain U.S. nuclear leadership so that we can accelerate and influence world measures to prevent proliferation while developing uranium and thorium as future world energy resources

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.)

Simplified compact containment BWR plant

International Nuclear Information System (INIS)

Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

2004-01-01

The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

BWR alloy 182 stress Corrosion Cracking Experience

International Nuclear Information System (INIS)

Horn, R.M.; Hickling, J.

2002-01-01

Modern Boiling Water Reactors (BWR) have successfully operated for more than three decades. Over that time frame, different materials issues have continued to arise, leading to comprehensive efforts to understand the root cause while concurrently developing different mitigation strategies to address near-term, continued operation, as well as provide long-term paths to extended plant life. These activities have led to methods to inspect components to quantify the extent of degradation, appropriate methods of analysis to quantify structural margin, repair designs (or strategies to replace the component function) and improved materials for current and future application. The primary materials issue has been the occurrence of stress corrosion cracking (SCC). While this phenomenon has been primarily associated with austenitic stainless steel, it has also been found in nickel-base weldments used to join piping and reactor internal components to the reactor pressure vessel consistent with fabrication practices throughout the nuclear industry. The objective of this paper is to focus on the history and learning gained regarding Alloy 182 weld metal. The paper will discuss the chronology of weld metal cracking in piping components as well as in reactor internal components. The BWR industry has pro-actively developed inspection processes and procedures that have been successfully used to interrogate different locations for the existence of cracking. The recognition of the potential for cracking has also led to extensive studies to understand cracking behavior. Among other things, work has been performed to characterize crack growth rates in both oxygenated and hydrogenated environments. The latter may also be relevant to PWR systems. These data, along with the understanding of stress corrosion cracking processes, have led to extensive implementation of appropriate mitigation measures. (authors)

The JAERI code system for evaluation of BWR ECCS performance

International Nuclear Information System (INIS)

Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

1982-12-01

Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

Nuclear power for developing countries

International Nuclear Information System (INIS)

Hirschmann, H.; Vennemann, J.

1980-01-01

The paper describes the energy policy quandary of developing countries and explains why nuclear power plants of a suitable size - the KKW 200 MW BWR nuclear power plant for electric power and/or process steam generation is briefly presented here - have an economic advantage over fossil-fuelled power plants. (HP) [de

From the nuclear world, no.5

International Nuclear Information System (INIS)

Anon.

2016-01-01

This document gathers information from around the world and concerning nuclear industry. The most relevant is the following. The British government has given its agreement for the construction of 2 EPR by EDF Energy at Hinkley Point. The Hinkley Point project will generate more than 1700 jobs in France. The EPR being built in Finland will operate in 2018. The electrical car is sustainable only in the countries where low-carbon electricity production is important which is the case of France thanks to nuclear energy. In the framework of the ICERR cooperation, 2 research reactors of CEA: Isis (Saclay) and RJH (being built at Cadarache) and their experimental facilities will be used by Slovenia, Tunisia and Morocco through joint research projects. In China 6 AP1000 units will be built on 3 sites in the Yangtze region. China and Turkey have signed an agreement concerning the organisation of nuclear safety authority. In Turkey a site for a 3. nuclear power plant is being selected. (A.C.)

The situation of the nuclear energy in the world (Oct. 1991)

International Nuclear Information System (INIS)

1991-10-01

This work presents an overview of the nuclear energy in the world. It approaches the following main topics: kinds of nuclear power plants; operation experience of the nuclear plants; environmental and social aspects of the nuclear energy; economic aspects of the nuclear energy; development of the reactors technology and supply of the nuclear fuel

The situation of the nuclear energy in the world (Sep. 1992)

International Nuclear Information System (INIS)

Souza, Jair Albo Marques de

1992-09-01

This work presents an overview of the nuclear energy in the world. It approaches the following main topics: kinds of nuclear power plants; operation experience of the nuclear plants; environmental and social aspects of the nuclear energy; economic aspects of the nuclear energy; development of the reactors technology and supply of the nuclear fuel

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: a route out of world crisis

International Nuclear Information System (INIS)

Jeffs, Eric.

1981-01-01

Presentations at the Eleventh World Energy Conference in September 1980 are highlighted, with the emphasis on nuclear energy. High oil prices have adversely affected the economies of many countries, but especially the less developed countries (LDC). The extent to which nuclear power can help the LDCs in the near future is limited by the size of their electricity grids. In the more developed countries, the discussion of nuclear energy is dominated by perceived need and public acceptance. The crisis of confidence in nuclear energy is not completely global. Both France and the COMECON countries have ambitious long-range plans, including nuclear combined heat and power units, and the more advanced developing countries, such as Argentina, Brazil, Korea and Mexico, are also pressing ahead with nuclear programs. (NDH)

Status update of the BWR cask simulator

Energy Technology Data Exchange (ETDEWEB)

Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-09-01

The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

Stability of transition to a world without nuclear weapons: Technical problems of verification

International Nuclear Information System (INIS)

Zhigalov, V.

1998-01-01

A serious psychological barrier to acceptance of the concept for achieving the nuclear-weapon-free world is fear of facing the prospect that one or more nations or extremist political groups might develop their own nuclear weapons. Actually this is a question of stability of the nuclear-weapon-free world. From this point of view the most effective system of verification is an absolute necessity. This system must ensure detection of so called undeclared nuclear activity at early stage. Scientists of Russian nuclear centers are working today on solving this problem. This paper is considered to be a comprehensive attempt to analyze the technical and organizational aspects of the problems of transition to a nuclear-weapons-free world, setting aside the difficulties of resolving purely political problems

World Nuclear University School of Uranium Production: Eight years' experience

International Nuclear Information System (INIS)

Trojacek, J.

2014-01-01

The World Nuclear University School of Uranium Production was established by DIAMO, state enterprise in 2006 year under the auspices of the World Nuclear University in London in partnership with international nuclear organizations – OECD/NEA and IAEA. Using the expertise and infrastructure of DIAMO State Enterprise, in conjuction with national and international universities, scientific institutions, regulatory authorities and other individual experts, the “school” covers its mission with the aim to provide world-class training on all aspects of uranium production cycle to equip operators, regulators and executives with the knowledge and expertise needed to provide expanded, environmentally-sound uranium mining throughout the world: • to educate students on all aspects of uranium production cycle including exploration, planning, development, operation, remediation and closure of uranium production facilities; • to improve the state of the art of uranium exploration, mining and mine remediation through research and development; • to provide a forum for the exchange of information on the latest uranium mining technologies and experiences – best practices.

Advanced methods for BWR transient and stability analysis

Energy Technology Data Exchange (ETDEWEB)

Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)

2008-07-01

The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

Advanced methods for BWR transient and stability analysis

International Nuclear Information System (INIS)

Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

2008-01-01

The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

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.)

Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1992-03-01

The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

The nuclear spread: a Third World view

International Nuclear Information System (INIS)

Kapur, A.

1980-01-01

The view of the Third World of nuclear power and international relations is contrasted with that of the superpowers. Aspects considered include the Non-Proliferation Treaty, factors for and against proliferation, regional and international environments, vertical versus horizontal proliferation. (U.K.)

BWR stability: analysis of cladding temperature for high amplitude oscillations - 146

International Nuclear Information System (INIS)

Pohl, P.; Wehle, F.

2010-01-01

Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge on BWR instabilities and possible consequences to fuel rod integrity. The objective of this paper is to present a simplified stability tool, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. In case of high amplitude oscillations a cyclical dryout and rewetting process at the fuel rod may take place, which leads in turn to rapid changes of the heat transfer from the fuel rod to the coolant. The application of this stability tool allows for a conservative determination of the fuel rod cladding temperature in case of high amplitude oscillations during the dryout / re-wet phase. Moreover, it reveals in good agreement to experimental findings the stabilizing effect of the reverse bundle inlet flow, which might be obtained for large oscillation amplitudes. (authors)

Neutron activation analysis and activity in the vessel steel of a BWR reactor for their study without radiological risks in microscopy and spectrometry; Analisis de activacion neutronica y actividad en el acero de la vasija de un reactor nuclear tipo BWR para su estudio sin riesgos radiologicos en microscopia y espectrometria

Energy Technology Data Exchange (ETDEWEB)

Moranchel, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Unidad Profesional Adolfo Lopez Mateos, Zacatenco, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.mx [IAEA, Department of Technical Cooperation, Division for Latin America, Room B1109 Wagramerstrasse 5, PO Box 100, A-1400, Vienna (Austria)

2012-07-01

The vessel material of nuclear reactors is subject to irradiation damage induced by the bombardment of neutrons coming from the reactor core. Neutrons are classified as fast and thermal, which produce different effects. Fast neutrons cause damage to the material by dislocation or displacement of atoms in the crystal structure, while the effect of thermal neutrons is a nuclear transmutation that can significantly change the properties of the material. The type and intensity of damage is based on the characteristics of the material, the flow of neutrons and the modes of neutrons interaction with the atomic structures of the material, among others. This work, alluding to nuclear transmutation, makes an analysis of neutron activation of all isotopes in a steel boiling water nuclear reactor (BWR) vessel. An analytical expression is obtained in order to model activity of steel, on the basis of the weight percentage of its atomic components. Its activity is theoretically estimated in a witness sample of the same material as that of the vessel, placed within the nuclear reactor since the beginning of its commercial operation in April 1995, up to August 2010. It was theoretically determined that the witness sample, with a 0.56 g mass (1 x 1 x 0.07 cm{sup 3} dimensions or equivalent) does not present a radiological risks during the stage of preparation, observation and analysis of it in electron microscopy and X-ray diffraction equipment s. The theoretical results were checked experimentally by measuring the activity of the sample by means of gamma spectrometry, measurement of the exposure levels around the sample, as well as the induced level to whole body and limbs, using thermo-luminescent dosimetry (TLD). As a result of the theoretical analysis, new chemical elements are predicted, as a result of the activation phenomena and radioactive decay, whose presence can be a fundamental factor of change in the properties of the vessel. This work is a preamble to the

Assessment of the Prony's method for BWR stability analysis

International Nuclear Information System (INIS)

Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Palacios-Hernandez, Javier C.

2011-01-01

Highlights: → This paper describes a method to determine the degree of stability of a BWR. → Performance comparison between Prony's and common AR techniques is presented. → Benchmark data and actual BWR transient data are used for comparison. → DR and f results are presented and discussed. → The Prony's method is shown to be a robust technique for BWR stability. - Abstract: It is known that Boiling Water Reactors are susceptible to present power oscillations in regions of high power and low coolant flow, in the power-flow operational map. It is possible to fall in one of such instability regions during reactor startup, since both power and coolant flow are being increased but not proportionally. One other possibility for falling into those areas is the occurrence of a trip of recirculation pumps. Stability monitoring in such cases can be difficult, because the amount or quality of power signal data required for calculation of the stability key parameters may not be enough to provide reliable results in an adequate time range. In this work, the Prony's Method is presented as one complementary alternative to determine the degree of stability of a BWR, through time series data. This analysis method can provide information about decay ratio and oscillation frequency from power signals obtained during transient events. However, so far not many applications in Boiling Water Reactors operation have been reported and supported to establish the scope of using such analysis for actual transient events. This work presents first a comparison of decay ratio and frequency oscillation results obtained by Prony's method and those results obtained by the participants of the Forsmark 1 and 2 Boiling Water Reactor Stability Benchmark using diverse techniques. Then, a comparison of decay ratio and frequency oscillation results is performed for four real BWR transient event data, using Prony's method and two other techniques based on an autoregressive modeling. The four

Development and future perspective of nuclear power plants. Current status and future prospect of world nuclear power plants

International Nuclear Information System (INIS)

Kobayashi, Masaharu

2013-01-01

Fukushima Daiichi NPS accidents occurred on 11 March 2011 brought about great effects on nuclear development not only in Japan but also in the world. In Japan restart of operation of periodically inspected nuclear power plants (NPPs) could not be allowed except Oi NPPs two units and most parties except Liberal Democratic Party (LDP) pledged to possibly phasing out nuclear power at House of Councillors election in July and public opinion was mostly against nuclear power after the accident. LDP clearly stated that, with the inauguration of new government last December, Japan would not pursuing the policy of the prior government of possibly phasing out nuclear power by the 2030s, but would instead make a 'zero-base' review of energy policy. Germany decided to close eight reactors immediately and remaining nine by the end of 2022. For many countries, nuclear power would play an important role in achieving energy security and sustainable development goals. In 2011 NPPs 6 units started operation with 2 units under construction, and in 2012 NPPs 3 units started operation with 7 units under construction. At present there are now over 400 NPPs operating in 31 countries and world trend seemed nuclear development was continued and number of countries newly deploying NPPs was increasing as much as eighteen. This article presented current status and future prospect of world NPPs in details. Japan would like to share its experiences and information obtained from the accident with the world and also promote NPPs overseas to meet the world's expectations. (T. Tanaka)

Great expectations. Projections of nuclear power around the world

International Nuclear Information System (INIS)

McDonald, Alan; Rogner, Hans-Holger; Gritsevskyi, Andrii

2009-01-01

In its 2008 edition of Energy, Electricity and Nuclear Power Estimates for the Period to 2030, the International Atomic Energy Agency (IAEA) has again revised its projections for nuclear power upwards. Every year since 1981 the IAEA has published 2 updated projections for the world's nuclear power generating capacity, a low projection and a high projection. The low projection is a down to earth, business-as-usual projection. The high projection takes into account government and corporate announcements about longer-term plans for nuclear investments as well as potential new national policies, e.g., to combat climate change. The results for the 2008 projections are presented. In the low projection, the projected nuclear power capacity in 2030 is 473 GW(e), some 27% higher than today's 372 GW(e). In the high projection, nuclear capacity in 2030 is 748 GW(e), double today's capacity. But while projections for nuclear power's future rose, its share of the world's electricity generation today dropped from 15% in 2006 to 14% in 2007. The main reason is that while total global electricity generation rose 4.8% from 2007 to 2008, nuclear electricity actually dropped slightly. The overall message from the IAEA's 2008 edition of Energy, Electricity and Nuclear Power Estimates for the Period to 2030 is that global electricity use will grow significantly, that nuclear power will have to expand more rapidly than it has done recently in order to maintain its share, and that nuclear power can meet the challenge. (orig.)

Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

1993-02-01

This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (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.)

World nuclear generating capacity and uranium requirements to 2005

International Nuclear Information System (INIS)

Anon.

1991-01-01

The outlook for the world nuclear power industry through 2005 is more positive than some may believe. Installed nuclear electric generating capacity is forecast to grow at an average rate of 2.4 percent per year, and reach 448 gigawatts electric (GWe) by 2005. Consequently, annual world uranium requirements also will grow, reaching over 200 million pounds equivalent U 3 O 8 by 2005. This article presents data and summarizes installed nuclear generating capacity and charts its increase as a function of time through the year 2005. This data is also charted by reactor type as well as reactor status: under construction, planned, or estimated future construction. In a similar fashion, the data is also charted by country and continent. Historical and projected data is also given for capacity factor

Regression analysis of nuclear plant capacity factors

International Nuclear Information System (INIS)

Stocks, K.J.; Faulkner, J.I.

1980-07-01

Operating data on all commercial nuclear power plants of the PWR, HWR, BWR and GCR types in the Western World are analysed statistically to determine whether the explanatory variables size, year of operation, vintage and reactor supplier are significant in accounting for the variation in capacity factor. The results are compared with a number of previous studies which analysed only United States reactors. The possibility of specification errors affecting the results is also examined. Although, in general, the variables considered are statistically significant, they explain only a small portion of the variation in the capacity factor. The equations thus obtained should certainly not be used to predict the lifetime performance of future large reactors

Highlights of the IAEA scientific forum: Nuclear science: Physics helping the world

International Nuclear Information System (INIS)

Richter, B.

2005-01-01

The conclusions drawn at the Scientific Forum during the 49th regular Session of the IAEA General Conference entitled 'Nuclear Physics: Helping the World' are as follows. Physics is indeed helping the world and the IAEA, applications of nuclear are continuing to grow, nuclear power is likely to increase dramatically, work on 'Beyond Kyoto' should begin and nuclear should be part of any 'Clean Development Mechanism'. IAEA should look again at the role of R and D in safeguards, IAEA should look at safety issues in newly nuclear countries, internationalizing the fuel cycle is right IF one can do it

Analysis of results of AZTRAN and AZKIND codes for a BWR

International Nuclear Information System (INIS)

Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L.; Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M.

2016-09-01

This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

Nuclear power for Third World countries: A necessary evil

International Nuclear Information System (INIS)

Ratsch, U.

1984-01-01

The possible role of nuclear energy for typical spheres of life and energy-related services in the Third World is discussed, starting alternatives to nuclear energy in each case. The sequence of the life spheres shows a categorization of the energy need as rural and urban. (DG) [de

2009 nuclear power world report

International Nuclear Information System (INIS)

Anon.

2010-01-01

At the end of 2009, 437 nuclear power plants were available for energy supply in 30 countries of the world. This is 1 plant less than at the end of 2008. The aggregate gross power of the plants amounted to approx. 391.5 GWe, the aggregate net power, to 371.3 GWe. This capacity numbers are a little bit less than one year before (gross: 392.6 GWe, net: 372.2 GWe). Two units were commissioned in 2009; 1 unit in India (Rajasthan 5) and 1 unit in Japan (Tomari 3). Three nuclear power plant were shut down permanently in 2009 in Japan (Hamaoka 1 and Hamaoka 2) and in Lithuania (Ignalina 2). 52 nuclear generating units, i.e. 9 plants more than at the end of 2008, were under construction in late 2009 in 14 countries with an aggregate gross power of approx. 51.2 GWe. Worldwide, some 80 new nuclear power plants are in the concrete project design, planning, and licensing phases; in some of these cases license applications have been submitted or contracts have already been signed. Some 130 further projects are planned. Net electricity generation in nuclear power plants worldwide in 2009 achieved another reasonable ranking level of approx. 2,558 billion kWh (2008: approx. 2,628 billion kWh). Since the first generation of electricity in a nuclear power plant in the EBR-I fast breeder (USA) on December 20, 1951, cumulated net production has reached approx. 60,500 billion kWh, and operating experience has grown to some 13,950 reactor years. (orig.)

Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

International Nuclear Information System (INIS)

Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

1991-01-01

The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

Nuclear energy - stabilising factor in the world economy

International Nuclear Information System (INIS)

Legassov, V.; Feoktistov, L.; Kouzmine, I.

1986-01-01

One of the most important factors for international stability is the development of the economy, reducing the risk of local armed conflicts which could escalate into world-wide nuclear war. Economic progress which plays such a vital part is in turn heavily dependent on energy supplies. The article takes a brief look at the role of nuclear power in this context. (B.M.S.)

Data list of nuclear power plants in Japan

International Nuclear Information System (INIS)

Izumi, Fumio; Horikami, Kunihiko; Kobayashi, Kensuke; Namatame, Ken.

1993-01-01

The development of the database called PPD (Nuclear Power Plant Database) has started in 1983 at JAERI as a six-year program to provide useful information for reactor safety regulation and reactor safety research. In 1988 the program has been accomplished, and since then the data in the database has been updating and adding. Information source of the PPD is based on SAR's (Safety Analysis Report) of 47 nuclear power plants which are operating, under construction or under licensing review in Japan. The report, BWR edition, consists of lists of major data stored in the PPD, relating to safety design of 25 BWR plants in Japan. (author)

Nuclear Power, Nuclear Fuel Cycle and Sustainable Development in a Changing World

International Nuclear Information System (INIS)

Arakawa, Yoshitaka

2000-01-01

Important changes concerning nuclear energy are coming to the fore, such as economic competitiveness compared to other energy resources, requirement for severe measures to mitigate man-made greenhouse gas (GHG) emission, due to the rise of energy demand in Central and Eastern Europe and Asia and to the greater public concern with respect to the nuclear safety, particularly related to spent fuel and radioactive waste disposal. Global safety culture, as well as well focused nuclear research and development programs for safer and more efficient nuclear technology manifest themselves in a stronger and effective way. Information and data on nuclear technology and safety are disseminated to the public in timely, accurate and understandable fashion. Nuclear power is an important contributor to the world's electricity needs. In 1999, it supplied roughly one sixth of global electricity. The largest regional percentage of electricity generated through nuclear power last year was in western Europe (30%). The nuclear power shares in France, Belgium and Sweden were 75%, 58% and 47%, respectively. In North America, the nuclear share was 20% for the USA and 12% for Canada. In Asia, the highest figures were 43% for the Republic of Korea and 36% for Japan. In 1998, twenty-three nations produced uranium of which, the ten biggest producers (Australia, Canada, Kazakhstan, Namibia, Niger, the Russian Federation, South Africa, Ukraine, USA and Uzbekistan) supplied over 90% of the world's output. In 1998, world uranium production provided only about 59% of world reactor requirements. In OECD countries, the 1998 production could only satisfy 39% of the demand. The rest of the requirements were satisfied by secondary sources including civilian and military stockpiles, uranium reprocessing and re-enrichment of depleted uranium. With regard to the nuclear fuel industry, an increase in fuel burnup, higher thermal rates, longer fuel cycle and the use of mixed uranium-plutonium oxide (MOX

The World Nuclear Industry Status Report 2014

International Nuclear Information System (INIS)

Schneider, Mycle; Froggatt, Antony; Ayukawa, Yurika; Burnie, Shaun; Piria, Raffaele; Thomas, Steve; Hazemann, Julie; Suzuki, Tatsujiro

2014-07-01

The World Nuclear Industry Status Report 2014 provides a comprehensive overview of nuclear power plant data, including information on operation, production and construction. The WNISR assesses the status of new-build programs in current nuclear countries as well as in potential newcomer countries. A 20-page chapter on nuclear economics looks at the rapidly changing market conditions for nuclear power plants, whether operating, under construction, or in the planning stage. Reactor vendor strategies and the 'Hinkley Point C Deal' are analyzed in particular. The performance on financial markets of major utilities is documented. The WNISR2013 featured for the first time a Fukushima Status Report that triggered widespread media and analyst attention. The 2014 edition entirely updates that Fukushima chapter. The Nuclear Power vs. Renewable Energy chapter that provides comparative data on investment, capacity, and generation has been greatly extended by a section on system issues. How does nuclear power perform in systems with high renewable energy share? Is this the end of traditional baseload/ peak-load concepts? Finally, the 45-page Annex 1 provides a country-by-country overview of all 31 countries operating nuclear power plants, with extended Focus sections on China, Japan, and the United States

Dry well cooling systems in BWR type nuclear power plants

International Nuclear Information System (INIS)

Hanamura, Ikuo; Tada, Kenji.

1986-01-01

Purpose: To prevent the damages of pipeways due to salt damages at the surface of control rod drives in BWR type reactors. Constitution: In control rod drives and the lowermost area in the dry well in which surface corrosion and pitching have been resulted by the salt contents in air due to the increase in the humidity accompanying the lowering of the temperature, a blower is disposed to the upstream of the cooling coils and a portion of high temperature air returned to the lower cooler is replaced with a low temperature feed air to increase the feed temperature in the area. Further, by upwardly turning the downwarded feed air drawing port in which cold feed air has so far been descended as it is, the descendance of the cold air is suppressed. As a result, temperature lowering in the driving mechanisms and the lower area can be prevented to obtain a predetermined temperature, whereby the dewing on the surface can be prevented and thereby preventing the occurrence of corrosion and pitching. (Horiuchi, T.)

Review of world nuclear power programs

International Nuclear Information System (INIS)

Rippon, S.

1978-01-01

Political, economic and environmental decisions are still affecting the growth of nuclear power generation throughout the world, but there are signs that proven past performance and increasing prices of energy from conventional sources are becoming overriding factors. In the USA, some uncertainty has been created by the moratorium on reprocessing enacted by the 1978 Non-Proliferation Act. The uncertainty has spread to importers of nuclear fuel and technology, but there is increasing international acceptance of the International Nuclear Fuel Cycle Evaluation. Isotope separation capacity to be installed by the US government, by the joint European organization Urenco and by the French Eurodif should ease the supply of fuel. There is little progress on waste-management policy in the USA, but in Germany a company (DWK) has been formed and a site for a rocksalt repository has been designated at Gorleben. In the UK the Windscale inquiry has a positive significance for nuclear energy generally as well as for reprocessing. An inquiry in Sweden has also come to positive conclusions. The status of nuclear energy in these and a number of other countries is discussed in terms of political and national policies. (N.D.H.)

World nuclear fuel cycle requirements 1990

International Nuclear Information System (INIS)

1990-01-01

This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under three nuclear supply scenarios. Two of these scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries with free market economies (FME countries). A No New Orders scenario is presented only for the United States. These nuclear supply scenarios are described in Commercial Nuclear Power 1990: Prospects for the United States and the World (DOE/EIA-0438(90)). This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the FME projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2030 for the Lower and Upper Reference cases and through 2040, the last year in which spent fuel is discharged, for the No New Orders case. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management

Necessity of nuclear energy in energetic world context

International Nuclear Information System (INIS)

Lopez Rodriguez, M.

1981-01-01

Different opinions on nuclear energy make the middle citizen feel confounded and wonder hundreds of questions to wwhich an easy reply is not found. May be if nuclear energy is really necessary, the first of these questions, without noticing that necessity is a vague concept with a double interpretation. To some, those support a total change in the actual society into more primitive situations, the energy pattern the world has chosen -both the East and West models- is annoying, and they consider a pattern based on ''soft energies''to be the solution to the social scheme they imagined. To others, those who think on an economic, industrial and social development in the countries, it should be based on a strong energy pattern, which could supply what the world needs more and more, nuclear energy is, at least nowadays, an unavoidable necessity and an inevitable option. The document shown has been prepared on the conclusions of the most recent works on the subject, and it is deduced from all of them what everybody considers to be the future energy demand for the year 2000 and its distribution into energy sources, nuclear energy includes. The two basic parameters for tAe valuation of this demand are the increasing of population and gross national product. Available energy resources are mentioned on the document and, mainly, the nuclear capacity of each country. (author) [es

Cooperative control scheme for an HVDC system connected to an isolated BWR nuclear power plant

International Nuclear Information System (INIS)

Sakurai, T.; Goto, K.; Kawai, T.; Matori, I.; Nakao, T.; Watanabe, A.

1983-01-01

This paper describes a cooperative control system to achieve stable operation of an isolated BWR nuclear plant linked to an HVDC system. In the proposed control system, under normal conditions the power plant is controlled according to the generating power reference and the generator frequency deviation is adjusted by converter power control. Such frequency control is also effective in the case of AC-DC system faults. In addition to the frequency control, an overload control is provided with the HVDC system, where the DC transmission power in the sound poles is increased due to a fault detection signal from the faulty pole. Effects of the above mentioned control systems were studied using digital dynamic programs. The sets of simulation results confirmed that in the case of a DC single pole fault, the plant is able to continue operation without any use of the turbine speed control units even for a restarting failure in the faulty pole. In case of a DC two pole fault, the plant is able to continue operation, being assisted by turbine speed control units when restarting in the faulty poles succeeds. In case of an AC three-line to ground fault near the AC terminal of the converter at the sending or receiving end, the system is able to continue stable operation, being supplemented by the turbine control unit when the faulty section of the AC system is isolated by a main or back-up relaying system

Nuclear Policy and World Order: Why Denuclearization. World Order Models Project. Occasional Paper Number Two.

Science.gov (United States)

Falk, Richard A.

The monograph examines the relationship of nuclear power to world order. The major purpose of the document is to stimulate research, education, dialogue, and political action for a just and peaceful world order. The document is presented in five chapters. Chapter I stresses the need for a system of global security to counteract dangers brought…

Material operating behaviour of ABB BWR control rods

International Nuclear Information System (INIS)

Rebensdorff, B.; Bart, G.

2000-01-01

The BWR control rods made by ABB use boron carbide (B 4 C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B 4 C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

Requirements of coolants in nuclear reactors

International Nuclear Information System (INIS)

Abass, O. A. M.

2014-11-01

This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

General Electric's training program for BWR chemists

International Nuclear Information System (INIS)

Osborn, R.N.; Lim, W.

1981-01-01

This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

BWR SFAT, gross-defect verification of spent BWR fuel. Final report on Task FIN A563 on the Finnish Support Programme to IAEA Safeguards including BWR SFAT User Manual

International Nuclear Information System (INIS)

Tarvainen, M.; Paakkunainen, M.; Tiitta, A.; Sarparanta, K.

1994-04-01

A measurement instrument called Spent Fuel Attribute Tester, SFAT, has been designed, fabricated and taken into use by the IAEA in gross defect verification of spent BWR fuel assemblies. The equipment consists of an underwater measurement head connected with cables to a control unit on the bridge of the fuel handling machine as well as to a PMCA for measurement of the gamma spectra. The BWR SFAT is optimized for the AFR interim storage, TVO KPA-STORE, of the TVO Power Company in Olkiluoto, Finland. It has a shape and it is moved like a fuel assembly using the fuel handling machine. No fuel movements are needed. Spent fuel specific radiation from the fission product 137 Cs at the gamma-ray energy of 662 keV is detected above the assemblies in the storage rack using a NaI(Tl) detector. In the design and in licensing the requirements of the IAEA, operator and the safety authority have been taken into account. The BWR SFAT allows modifications for other LWR fuel types with minor changes. The work has been carried out under the task FIN A 563 of the Finnish Support Programme to IAEA Safeguards. (orig.) (9 refs., 22 figs.)

BWR spent fuel transport and storage system for KKL: TN trademark 52L, TN trademark 97L, TN trademark 24 BHL

International Nuclear Information System (INIS)

Sicard, D.; Verdier, A.; Monsigny, P.A.

2004-01-01

The LEIBSTADT (KKL) nuclear power plant in Switzerland has opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN trademark a52L and TN trademark 97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TNae24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN trademark 52L and TN trademark 97L casks by the KKL and ZWILAG operators

World nuclear fuel cycle requirements, 1988

International Nuclear Information System (INIS)

1988-01-01

This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the (WOCA) World Outside Centrally Planned Economic Areas projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix E includes aggregated domestic spent fuel projections through the year 2020 for the Lower and Upper References cases and through 2037, the last year in which spent fuel is discharged, for the No New Orders case. Annual projections of spent fuel discharges through the year 2037 for individual US reactors in the No New Orders cases are included for the first time in Appendix H. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management

World nuclear fuel cycle requirements 1985

International Nuclear Information System (INIS)

Moden, R.; O'Brien, B.; Sanders, L.; Steinberg, H.

1985-01-01

Projections of uranium requirements (both yellowcake and enrichment services) and spent fuel discharges are presented, corresponding to the nuclear power plant capacity projections presented in ''Commercial Nuclear Power 1984: Prospects for the United States and the World'' (DOE/EIA-0438(85)) and the ''Annual Energy Outlook 1984:'' (DOE/EIA-0383(84)). Domestic projections are provided through the year 2020, with foreign projections through 2000. The domestic projections through 1995 are consistent with the integrated energy forecasts in the ''Annual Energy Outlook 1984.'' Projections of capacity beyond 1995 are not part of an integrated energy foreccast; the methodology for their development is explained in ''Commercial Nuclear Power 1984.'' A range of estimates is provided in order to capture the uncertainty inherent in such forward projections. The methodology and assumptions are also stated. A glossary is provided. Two appendixes present additional material. This report is of particular interest to analysts involved in long-term planning for the disposition of radioactive waste generated from the nuclear fuel cycle. 14 figs., 18 tabs

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-01

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

Recent technology for BWR operator training simulators

International Nuclear Information System (INIS)

Sato, Takao; Hashimoto, Shigeo; Kato, Kanji; Mizuno, Toshiyuki; Asaoka, Koichi.

1990-01-01

As one of the important factors for maintaining the high capacity ratio in Japanese nuclear power stations, the contribution of excellent operators is pointed out. BWR Operation Training Center has trained many operators using two full scope simulators for operation training modeling BWRs. But in order to meet the demands of the recent increase of training needs and the upgrading of the contents, it was decided to install the third simulator, and Hitachi Ltd. received the order to construct the main part, and delivered it. This simulator obtained the good reputation as its range of simulation is wide, and the characteristics resemble very well those of the actual plants. Besides, various new designs were adopted in the control of the simulator, and its handling became very easy. Japanese nuclear power plants are operated at constant power output, and the unexpected stop is very rare, therefore the chance of operating the plants by operators is very few. Accordingly, the training using the simulators which can simulate the behavior of the plants with computers, and can freely generate abnormal phenomena has become increasingly important. The mode and positioning of the simulators for operation training, the full scope simulator BTC-3 and so on are reported. (K.I.)