Cost analysis of light water reactor power plants

International Nuclear Information System (INIS)

Mooz, W.E.

1978-06-01

A statistical analysis is presented of the capital costs of light water reactor (LWR) electrical power plants. The objective is twofold: to determine what factors are statistically related to capital costs and to produce a methodology for estimating these costs. The analysis in the study is based on the time and cost data that are available on U.S. nuclear power plants. Out of a total of about 60 operating plants, useful capital-cost data were available on only 39 plants. In addition, construction-time data were available on about 65 plants, and data on completed construction permit applications were available for about 132 plants. The cost data were first systematically adjusted to constant dollars. Then multivariate regression analyses were performed by using independent variables consisting of various physical and locational characteristics of the plants. The dependent variables analyzed were the time required to obtain a construction permit, the construction time, and the capital cost

Operational margin monitoring system for boiling water reactor power plants

International Nuclear Information System (INIS)

Fukutomi, S.; Takigawa, Y.

1992-01-01

This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

The low-temperature water-water reactor for district heating atomic power plant (DHPP)

International Nuclear Information System (INIS)

Skvortsov, S.A.; Sokolov, I.N.; Krauze, L.V.; Nikiporetz, Yu.G.; Philimonov, Yu.V.

1977-01-01

The district heating atomic power plant in the article is distinguished by the increased reliability and safety of operation that was provided by the use of following main principles: relatively low parameters of the coolant; the intergral arrangement of equipment and accordingly the minimum branching of the reactor circuit; the natural circulation of coolant of the primary circuit in the steady-state, transient and emergency regimes of reactor operation; the considerable reserves of cold water of the primary circuit in the reactor vessel, providing the emergency cooling; the application of two shells each of which is designed for the total working pressure, the second shell is made of prestressed reinforced concrete that eliminates its brittle failure. (M.S.)

Uranium-236 in light water reactor spent fuel recycled to an enriching plant

International Nuclear Information System (INIS)

de la Garza, A.

1977-01-01

The introduction of 236 U to an enriching plant by recycling spent fuel uranium results in enriched products containing 236 U, a parasitic neutron absorber in reactor fuel. Convenient approximate methodology determines 235 236 U, and total uranium flowsheets with associated separative work requirements in enriching plant operations for use by investigators of the light water reactor fuel cycle not having recourse to specialized multicomponent cascade technology. Application of the methodology has been made to compensation of an enriching plant product for 236 U content and to the value at an enriching plant of spent fuel uranium. The approximate methodology was also confirmed with more exact calculations and with some experience with 236 U in an enriching plant

Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

Energy Technology Data Exchange (ETDEWEB)

1993-05-13

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

SWR 1000: The new boiling water reactor power plant concept

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1989-10-01

Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule duration, and to simplify design. This document discusses construction approaches. 77 refs., 5 figs., 6 tabs

Multi-stage-flash desalination plants of relative small performance with integrated pressurized water reactors as a nuclear heat source

International Nuclear Information System (INIS)

Petersen, G.; Peltzer, M.

1977-01-01

In the Krupp-GKSS joint study MINIPLEX the requirements for seawater-desalination plants with a performance in the range of 10 000 to 80 000 m 3 distillate per day heated by a nuclear reactor are investigated. The reactor concept is similar to the Integrated Pressurized Water Reactor (IPWR) of the nuclear ship OTTO HAHN. The design study shows that IPWR systems have specific advantages up to 200 MWth compared to other reactor types at least being adapted for single- and dual-purpose desalination plants. The calculated costs of the desalinated water show that due to fuel cost advantages of reactors small and medium nuclear desalination plants are economically competetive with oil-fired plants since the steep rise of oil price in autumn 1973. (author)

Limiting factor analysis of high availability nuclear plants (boiling water reactors). Final report

International Nuclear Information System (INIS)

Frederick, L.G.; Brady, R.M.; Shor, S.W.W.; McCusker, J.T.; Alden, W.M.; Kovacs, S.

1979-08-01

The pertinent results are presented of a 16-month study conducted for Electric Power Research Institute by General Electric Company, Bechtel Power Corporation, and Philadelphia Electric Company. The study centered around the Peach Bottom 2 Atomic Power Station, but also included limited study of operations at 20 additional operating boiling water reactors. The purpose of the study was to identify and evaluate key factors limiting plant availability, and to identify potential improvements for eliminating or alleviating those limitations. The key limiting factors were found to be refueling activities; activities related to the reactor fuel; reactor scrams; activities related to 20 operating systems or major components; delays due to radiation, turbid water during refueling operations, facilities/working conditions, and dirt/foreign material; and general maintenance/repair of valves and piping. Existing programs to reduce the effect on plant unavailability are identified, and suggestions for further action are made

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1987-12-01

Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule durations, and to simplify design. This document discusses successes and problems in construction. 49 refs., 16 figs., 8 tabs

Optimization of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plant (FUWAC)

International Nuclear Information System (INIS)

2011-10-01

This report presents the results of the Coordinated Research Project (CRP) on Optimization of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plants (FUWAC, 2006-2009). It provides an overview of the results of the investigations into the current state of water chemistry practice and concerns in the primary circuit of water cooled power reactors including: corrosion of primary circuit materials; deposit composition and thickness on the fuel; crud induced power shift; fuel oxide growth and thickness; radioactivity buildup in the reactor coolant system (RCS). The FUWAC CRP is a follow-up to the DAWAC CRP (Data Processing Technologies and Diagnostics for Water Chemistry and Corrosion Control in Nuclear Power Plants 2001-2005). The DAWAC project improved the data processing technologies and diagnostics for water chemistry and corrosion control in nuclear power plants (NPPs). With the improved methods for controlling and monitoring water chemistry now available, it was felt that a review of the principles of water chemistry management should be undertaken in the light of new materials, more onerous operating conditions, emergent issues such as CIPS, also known as axial offset anomaly (AOA) and the ageing of operating power plant. In the framework of this CRP, water chemistry specialists from 16 nuclear utilities and research organizations, representing 15 countries, exchanged experimental and operational data, models and insights into water chemistry management. The CD-ROM attached to this IAEA-TECDOC includes the report itself, detailed progress reports of three Research Coordination Meetings (RCMs) (Annexes I-III) and the reports and presentations made during the project by the participants.

PWR type reactor plant

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1993-01-01

A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

Methods and technologies for cost reduction in the design of water cooled reactor power plants

International Nuclear Information System (INIS)

1991-05-01

The Specialists Meeting was organized in the framework of the IAEA International Working Group on Advanced Technologies for Water-Cooled Reactors. Its purpose was to provide an international forum for review and discussion on recent results in research and development on different methods and technologies of current and advanced water-cooled reactor power plants, which can lead to reduced investment and operation, maintenance and fuel-cycle costs of the plants. 27 specialists representing 10 countries and the IAEA took part in the meeting. 10 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1989-01-01

Under Contract No. AC03-86SF16565, Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule durations, and to simplify design. This document discusses design requirements. 36 refs., 57 figs., 56 tabs

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

International Nuclear Information System (INIS)

1995-08-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner's Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section

Recent computer applications in boiling water reactor power plants

International Nuclear Information System (INIS)

Hiraga, Shoji; Joge, Toshio; Kiyokawa, Kazuhiro; Kato, Kanji; Nigawara, Seiitsu

1976-01-01

Process computers in boiling water reactor power plants have won the position of important equipments for the calculation of the core and plant performances and for data logging. Their application technique is growing larger and larger every year. Here, two systems are introduced; plant diagnostic system and computerized control panel. The plant diagnostic system consists of the part processing the signals from a plant, the operation part mainly composed of a computer to diagnose the operating conditions of each system component using input signal, and the result display (CRT or typewriter). The concept on the indications on control panels in nuclear power plants is changing from ''Plant parameters and to be indicated on panel meters as much as possible'' to ''Only the data required for operation are to be indicated.'' Thus the computerized control panel is attracting attention, in which the process computer for processing the operating information and CRT display are introduced. The experimental model of that panel comprises and operator's console and a chief watchmen's console. Its functions are dialogic data access and the automatic selection of preferential information. (Wakatsuki, Y.)

A multi-stage-flash desalination plant of relative small performance with an integrated pressurized water reactor as a nuclear heat source

International Nuclear Information System (INIS)

Peltzer, M.; Petersen, G.

1976-01-01

In the Krupp-GKSS joint study MINIPLEX the requirements for seawater-desalination-plants with a performance in the range of 10,000 to 80,000 m 3 /d heated by a nuclear reactor are investigated. The reactor concept is similar to the integrated pressurized water reactor (IPWR) of the nuclear ship OTTO HAHN. The calculated costs of the desalinated water show, that due to the fuel cost advantages of reactors small and medium nuclear desalination plants are economically competetive with oil-fired plants since the steep rise of oil price in autumn 1973. (orig.) [de

Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

International Nuclear Information System (INIS)

1993-01-01

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+trademark during the US government's 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW t (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment

Modeling and simulation of pressurized water reactor power plant

International Nuclear Information System (INIS)

Wang, S.J.

1983-01-01

Two kinds of balance of plant (BOP) models of a pressurized water reactor (PWR) system are developed in this work - the detailed BOP model and the simple BOP model. The detailed model is used to simulate the normal operational performance of a whole BOP system. The simple model is used to combine with the NSSS model for a whole plant simulation. The trends of the steady state values of the detailed model are correct and the dynamic responses are reasonable. The simple BOP model approach starts the modelling work from the overall point of view. The response of the normalized turbine power and the feedwater inlet temperature to the steam generator of the simple model are compared with those of the detailed model. Both the steady state values and the dynamic responses are close to those of the detailed model. The simple BOP model is found adequate to represent the main performance of the BOP system. The simple balance of plant model was coupled with a NSSS model for a whole plant simulation. The NSSS model consists of the reactor core model, the steam generator model, and the coolant temperature control system. A closed loop whole plant simulation for an electric load perturbation was performed. The results are plausible. The coupling effect between the NSSS system and the BOP system was analyzed. The feedback of the BOP system has little effect on the steam generator performance, while the performance of the BOP system is strongly affected by the steam flow rate from the NSSS

Liquid radioactive waste processing system for pressurized water reactor plants

International Nuclear Information System (INIS)

Anon.

1976-01-01

This Standard sets forth design, construction, and performance requirements, with due consideration for operation, of the Liquid Radioactive Waste Processing System for pressurized water reactor plants for design basis inputs. For the purpose of this Standard, the Liquid Radioactive Waste Processing System begins at the interfaces with the reactor coolant pressure boundary and the interface valve(s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system, and at the point of recycle back to storage for reuse

Optimization of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plant (FUWAC). Additional Information

International Nuclear Information System (INIS)

2011-10-01

This report presents the results of the Coordinated Research Project (CRP) on Optimization of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plants (FUWAC, 2006-2009). It provides an overview of the results of the investigations into the current state of water chemistry practice and concerns in the primary circuit of water cooled power reactors including: corrosion of primary circuit materials; deposit composition and thickness on the fuel; crud induced power shift; fuel oxide growth and thickness; radioactivity buildup in the reactor coolant system (RCS). The FUWAC CRP is a follow-up to the DAWAC CRP (Data Processing Technologies and Diagnostics for Water Chemistry and Corrosion Control in Nuclear Power Plants 2001-2005). The DAWAC project improved the data processing technologies and diagnostics for water chemistry and corrosion control in nuclear power plants (NPPs). With the improved methods for controlling and monitoring water chemistry now available, it was felt that a review of the principles of water chemistry management should be undertaken in the light of new materials, more onerous operating conditions, emergent issues such as CIPS, also known as axial offset anomaly (AOA) and the ageing of operating power plant. In the framework of this CRP, water chemistry specialists from 16 nuclear utilities and research organizations, representing 15 countries, exchanged experimental and operational data, models and insights into water chemistry management. This CD-ROM attached to the printed IAEA-TECDOC includes the report itself, detailed progress reports of three Research Coordination Meetings (RCMs) (Annexes I-III) and the reports and presentations made during the project by the participants.

Next generation light water reactors

International Nuclear Information System (INIS)

Omoto, Akira

1992-01-01

In the countries where the new order of nuclear reactors has ceased, the development of the light water reactors of new type has been discussed, aiming at the revival of nuclear power. Also in Japan, since it is expected that light water reactors continue to be the main power reactor for long period, the technology of light water reactors of next generation has been discussed. For the development of nuclear power, extremely long lead time is required. The light water reactors of next generation now in consideration will continue to be operated till the middle of the next century, therefore, they must take in advance sufficiently the needs of the age. The improvement of the way men and the facilities should be, the simple design, the flexibility to the trend of fuel cycle and so on are required for the light water reactors of next generation. The trend of the development of next generation light water reactors is discussed. The construction of an ABWR was started in September, 1991, as No. 6 plant in Kashiwazaki Kariwa Power Station. (K.I.)

Pump selection and application in a pressurized water reactor electric generating plant

International Nuclear Information System (INIS)

Kitch, D.M.

1985-01-01

Various pump applications utilized in a nuclear pressurized water reactor electric generating plant are described. Emphasis is on pumps installed in the auxiliary systems of the primary nuclear steam supply system. Hydraulic and mechanical details, the ASME Code (Nuclear Design), materials, mechanical seals, shaft design, seismic qualification, and testing are addressed

Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

International Nuclear Information System (INIS)

Davis, G.A.

1992-01-01

Since 1985, ABB Combustion Engineering Nuclear Power (CENP) and Duke Engineering ampersand Services, Inc. (DE ampersand S) have been developing the next generation of pressurized water reactor (PWR) plant for worldwide deployment. The goal is to make available a pre-licensed, standardized plant design that can satisfy the need for a reliable and economic supply of electricity for residential, commercial and industrial use. To ensure that such a design is available when needed, it must be based on proven technology and established licensing criteria. These requirements dictate development of nuclear technology that is advanced, yet evolutionary in nature. This has been achieved with the System 80+ Standard Plant Design

Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

Energy Technology Data Exchange (ETDEWEB)

Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

Utility requirements for advanced light water reactors

International Nuclear Information System (INIS)

Machiels, A.; Gray, S.; Mulford, T.; Rodwell, E.

1996-01-01

The nuclear energy industry is actively engaged in developing advanced light water reactor (ALWR) designs for the next century. The new designs take advantage of the thousands of reactor-years of experience that have been accumulated by operating over 400 plants worldwide. The EPRI effort began in the early 1980's, when a survey of utility executives was conducted to determine their prerequisites for ordering nuclear power plants. The results were clear: new plants had to be simpler and safer, and have greater design margins, i.e., be more forgiving. The utility executives also supported making improvements to the established light water reactor technology, rather than trying to develop new reactor concepts. Finally, they wanted the option to build mid-size plants (∼600 MWe) in addition to full-size plants of more than 1200 MWe. 4 refs

New lineup of light water reactors

International Nuclear Information System (INIS)

Okamura, Kiyoshi; Oshima, Koichiro; Kitsukawa, Keisuke

2007-01-01

Toshiba is promoting technical studies for upcoming nuclear power plants based on its large accumulation of experience in boiling water reactor (BWR) design, manufacturing, construction, and maintenance. Our goal is to achieve higher reliability, lower life-cycle costs, and better competitiveness for nuclear power plants compared with other energy sources. In addition, we are developing a new light water reactor (LWR) lineup featuring the safest and most economical LWRs in the world as next-generation reactors almost at new construction and replacement in the Japanese and international markets expected to start from the 2020s. We are committed not only to developing BWRs with the world's highest performance but also to participating in the pressurized water reactor (PWR) market, taking advantage of the synergistic effect of both Toshiba's and Westinghouse's experience. (author)

Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

International Nuclear Information System (INIS)

1993-01-01

This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval

Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-09-15

This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

Status of the advanced boiling water reactor and simplified boiling water reactor

International Nuclear Information System (INIS)

Smith, P.F.

1992-01-01

This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

Catalogue and classification of technical safety rules for light-water reactors and reprocessing plants

International Nuclear Information System (INIS)

Bloser, M.; Fichtner, N.; Neider, R.

1975-08-01

This report on the cataloguing and classification of technical rules for land-based light-water reactors and reprocessing plants contains a list of classified rules. The reasons for the classification system used are given and discussed

Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR [pressurized-water-reactor] plants

International Nuclear Information System (INIS)

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

1988-01-01

Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs

Reactor water chemistry control

International Nuclear Information System (INIS)

Kundu, A.K.

2010-01-01

Tarapur Atomic Power Station - 1 and 2 (TAPS) is a twin unit Boiling Water Reactors (BWRs) built in 1960's and operating presently at 160MWe. TAPS -1 and 2 are one of the vintage reactors operating in the world and belongs to earlier generation of BWRs has completed 40 years of successful, commercial and safe operation. In 1980s, both the reactors were de-rated from 660MWth to 530MWth due to leaks in the Secondary Steam Generators (SSGs). In BWR the feed water acts as the primary coolant which dissipates the fission heat and thermalises the fast neutrons generated in the core due to nuclear fission reaction and under goes boiling in the Reactor Pressure Vessel (RPV) to produce steam. Under the high reactor temperature and pressure, RPV and the primary system materials are highly susceptible to corrosion. In order to avoid local concentration of the chemicals in the RPV of BWR, chemical additives are not recommended for corrosion prevention of the system materials. So to prevent corrosion of the RPV and the primary system materials, corrosion resistant materials like stainless steel (of grade SS304, SS304L and SS316LN) is used as the structural material for most of the primary system components. In case of feed water system, main pipe lines are of carbon steel and the heater shell materials are of carbon steel lined with SS whereas the feed water heater tubes are of SS-304. In addition to the choice of materials, another equally important factor for corrosion prevention and corrosion mitigation of the system materials is maintaining highly pure water quality and strict water chemistry regime for both the feed water and the primary coolant, during operation and shutdown of the reactor. This also helps in controlled migration of corrosion product to and from the reactor core and to reduce radiation field build up across the primary system materials. Experience in this field over four decades added to the incorporation of modern techniques in detection of low

State space modeling of reactor core in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

Dynamic analysis of the condensate feedwater system in boiling water reactor plants

International Nuclear Information System (INIS)

Tanji, J.; Omori, T.

1982-01-01

The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

Siting of light-water reactor power plants in the Federal Republic of Germany

International Nuclear Information System (INIS)

Kohler, H.A.G.

1975-01-01

The nuclear power plant site requirements formulated for environment protection in Germany allow nuclear power plants to be built at any site provided these requirements are duly taken into account in preparing and monitoring the site and in the design of the proposed power plant. After a brief discussion of light water reactor power plant sites, prevailing practice in site planning, site selection criteria, licensing procedure and used criteria, rules and guidelines, this paper reports on some considerations taken into account by the expert advisers and by the licensing authorities and future site planning. (orig.) [de

Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design

International Nuclear Information System (INIS)

Galvin, Mark R.; Todreas, Neil E.; Conway, Larry E.

2003-01-01

New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented

Normal and compact spent fuel storage in light water reactor power plants

International Nuclear Information System (INIS)

Kuenel, R.R.

1978-01-01

The compact storage of light water reactor spent fuel is a safe, cheap and reliable contribution towards overcoming the momentarily existing shortage in spent fuel reprocessing. The technical concept is described and physical behaviour discussed. The introduction of compact storage racks in nuclear power plants increases the capacity from 100 to about 240 %. The increase in decay heat is not more than about 14%, the increase in activity inventory and hazard potential does not exceed 20%. In most cases the existing power plant equipment fulfils the new requirements. (author)

Safety problems of nuclear power plants with channel-type graphite boiling water reactors

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Vasilevskij, V.P.; Volkov, V.P.; Gavrilov, P.A.; Kramerov, A.Ya.; Kuznetsov, S.P.; Kunegin, E.P.; Rybakov, N.Z.

1977-01-01

Construction of nuclear power plants in a highly populated region near large industrial centres necessitates to pay a special attention to their nuclear and radiation safety. Safety problems of nuclear reactor operation are discussed, in particular, they are: reliable stoppage of fission chain reaction at any emergency cases; reliable core cooling with failure of various equipment; emergency core cooling with breached pipes of a circulating circuit; and prevention of radioactive coolant release outside the nuclear power plant in amount exceeding the values adopted. Channel-type water boiling reactors incorporate specific features requiring a new approach to safety operation of a reactor and a nuclear power plant. These include primarily a rather large steam volume in the coolant circuit, large amount of accumulated heat, void reactivity coefficient. Channel-type reactors characterized by fair neutron balance and flexible fuel cycle, have a series of advantages alleviating the problem of ensuring their safety. The possibility of reliable control over the state of each channel allows to replace failed fuel elements by the new ones, when operating on-load, to increase the number of circulating loops and reduce the diameter of main pipelines, simplifies significantly the problem of channel emergency cooling and localization of a radioactive coolant release from a breached circuit. The concept of channel-type reactors is based on the solution of three main problems. First, plant safety should be assured in emergency switch off of separate units and, if possible, energy conditions should be maintained, this is of particular importance considering the increase in unit power. Second, the system of safety and emergency cooling should eliminate a great many failures of fuel elements in case of potential breaches of any tube in the circulating circuit. Finally, rugged boxes and localizing devices should be provided to exclude damage of structural elements of the nuclear power

A review of boiling water reactor water chemistry: Science, technology, and performance

International Nuclear Information System (INIS)

Fox, M.J.

1989-02-01

Boiling water reactor (BWR) water chemistry (science, technology, and performance) has been reviewed with an emphasis on the relationships between BWR water quality and corrosion fuel performance, and radiation buildup. A comparison of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.56, the Boiling Water Reactor Owners Group (BWROG) Water Chemistry Guidelines, and Plant Technical Specifications showed that the BWROG Guidelines are more stringent than the NRC Regulatory Guide, which is almost identical to Plant Technical Specifications. Plant performance with respect to BWR water chemistry has shown dramatic improvements in recent years. Up until 1979 BWRs experienced an average of 3.0 water chemistry incidents per reactor-year. Since 1979 the water chemistry technical specifications have been violated an average of only 0.2 times per reactor-year, with the most recent data from 1986-1987 showing only 0.05 violations per reactor-year. The data clearly demonstrate the industry-wide commitment to improving water quality in BWRs. In addition to improving water quality, domestic BWRs are beginning to switch to hydrogen water chemistry (HWC), a remedy for intergranular stress corrosion cracking. Three domestic BWRs are presently operating on HWC, and fourteen more have either performed HWC mini tests or are in various stages of HWC implementation. This report includes a detailed review of HWC science and technology as well as areas in which further research on BWR chemistry may be needed. 43 refs., 30 figs., 8 tabs

Realistic thermal transient margin analysis of 'MONJU' based on plant performance measurements. Reactor vessel outlet nozzle and evaporator feed water inlet tube sheet of the manual reactor plant trip

International Nuclear Information System (INIS)

Yamada, Fumiaki; Mori, Takero

2005-01-01

In order to develop technologies and achieve safe and stable operation of Monju' as well as realize optimized design and construction of safe and economically competitive fast breeder reactors, the authors are evaluating design approach applied to 'Monju' based on actually measured behavioral data during plant operations. This report uses actual measured characteristic data of 'Monju' during a plant trip test obtained at a commissioning stage with up to 40% power output and introduces plant thermal hydraulic behavior analysis in a representative thermal transient event, i.e. a manual plant trip. Thermal transient driven loads incurred by the reactor vessel outlet nozzle and by the evaporator feed water inlet tube sheet were further derived by structural analyses and were compared with the previously derived values in the design stage and with the limit values. Though the reactor vessel outlet nozzle was exposed to larger temperature change in the trip test than the analytical prediction, the newly calculated mechanical load was about 50% of the previous evaluation in the design stage. Also, the newly analyzed mechanical load incurred by the evaporator feed water inlet tube sheet in this event had a large margin against the limit value of cumulative damage cycle fraction, although the observed temperature disturbance in a steam blow test was wilder than the analytical prediction. Thus we concluded that the Monju' plant has an assured safety margin against thermal transient in plant trip events. (author)

Tritium issues in commercial pressurized water reactors

International Nuclear Information System (INIS)

Jones, G.

2008-01-01

Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

Plant life management processes and practices for heavy water reactors

International Nuclear Information System (INIS)

Kang, K.-S.; Cleveland, J.; Clark, C.R.

2006-01-01

In general, heavy water reactor (HWR) nuclear power plant (NPP) owners would like to keep their NPPs in service as long as they can be operated safely and economically. Their decisions are depending on essentially business model. They involve the consideration of a number of factors, such as the material condition of the plant, comparison with current safety standards, the socio-political climate and asset management/ business planning considerations. Continued plant operation, including operation beyond design life, called 'long term operation, depends, among other things, on the material condition of the plant. This is influenced significantly by the effectiveness of ageing management. Key attributes of an effective plant life management program include a focus on important systems, structure and components (SSCs) which are susceptible to ageing degradation, a balance of proactive and reactive ageing management programmes, and a team approach that ensures the co-ordination of and communication between all relevant nuclear power plant and external programmes. Most HWR NPP owners/operators use a mix of maintenance, surveillance and inspection (MSI) programs as the primary means of managing ageing. Often these programs are experienced-based and/or time-based and may not be optimised for detecting and/or managing ageing effects. From time-to-time, operational history has shown that this practice can be too reactive, as it leads to dealing with ageing effects (degradation of SSCs) after they have been detected. In many cases premature and/or undetected ageing cannot be traced back to one specific reason or an explicit error. The root cause is often a lack of communication, documentation and/or co-ordination between design, commissioning, operation or maintenance organizations. This lack of effective communication and interfacing frequently arises because, with the exception of major SSCs, such as the fuel channels or steam generators, there is a lack of explicit

Method and device for forecasting remaining lifetime for material constituting light water reactor plant

International Nuclear Information System (INIS)

Anzai, Hideya; Nakada, Kiyotomo; Shimanuki, Sei; Kida, Toshitaka; Fuse, Motomasa; Shigenaka, Naoto; Kuniya, Jiro; Izumiya, Masakiyo; Hattori, Shigeo; Saito, Takashi.

1994-01-01

A pressure vessel of a light water type reactor comprises a crack development sensor at the inside and a crack development monitor at the outside to monitor the development of cracks detected by the crack progress sensor. In addition, the reactor also comprises, at the outside thereof, a dissolved oxygen meter, a dissolved hydrogen peroxide meter and a conductivity meter for reactor water. A computer is connected, on line, to the crack development monitor, the dissolved oxygen meter, the dissolved hydrogen peroxide meter and the conductivity meter. A crack development rate measured by the crack development monitor, as well as the dissolved oxygen concentration, the dissolved peroxide hydrogen concentration and the conductivity of reactor water measured at the outside of the reactor by the dissolved oxygen meter, the dissolved hydrogen peroxide meter and the conductivity meter are inputted to the computer. The computer calculates the effective dissolved oxygen concentration for each portion of the plant based on these measured values. Further, the period of time till the crack reaches a predetermined limit value is calculated based on the measured values. Then, the period of time is displayed as a remaining life time of the materials due to stress corrosion crackings. (I.N.)

Safety of light water reactors. Risks of nuclear technology

International Nuclear Information System (INIS)

Veser, Anke; Schlueter, Franz-Hermann; Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen; Kessler, Guenter

2012-01-01

The book on the safety of light-water reactors includes the following chapters: Part I: Physical and technical safety concept of actual German and future European light-water reactors: (1) Worldwide operated nuclear power plants in 2011, (2) Some reactor physical fundamentals. (3) Nuclear power plants in Germany. (4) Radioactive exposure due to nuclear power plants. (5) Safety concept of light-water reactors. (6) Probabilistic analyses and risk studies. (7) Design of light-water reactors against external incidents. (8) Risk comparison of nuclear power plants and other energy systems. (9) Evaluation of risk studies using the improved (new) safety concept for LWR. (19) The severe reactor accidents of Three Mile Island, Chernobyl and Fukushima. Part II: Safety of German LWR in case of a postulated aircraft impact. (11) Literature. (12) Review of requirements and actual design. (13) Incident scenarios. (14) Load approach for aircraft impact. (15) Demonstration of the structural behavior in case of aircraft impact. (16) Special considerations. (17) Evaluation of the safety state of German and foreign nuclear power plants. Part III: ROSOS as example for a computer-based decision making support system for the severe accident management. (19) Literature. (20) Radiological fundamentals, accident management, modeling of the radiological situation. (21) The decision making support system RODOS. (22) RODOS and the Fukushima accident. (23) Recent developments in the radiological emergency management in the European frame.

Supercritical-pressure, once-through cycle light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi

2001-01-01

The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density. (author)

Water chemistry in boiling water reactors - A Leibstadt-specific overview

International Nuclear Information System (INIS)

Sarott, F.-A.

2005-01-01

The boiling water reactor (BWR) consists of two main water circuits: the water-steam cycle and the main cooling water system. In the introduction, the goals and tasks of the BWR plant chemistry are described. The most important objectives are the prevention of system degradation by corrosion and the minimisation of radiation fields. Then a short description of the BWR operation principle, including the water steam cycle, the transport of various impurities by the steam, removing impurities from the condensate, the reactor water clean-up system, the balance of plant and the main cooling water system, is given. Subsequently, the focus is set on the water-steam cycle chemistry. In order to fulfil the somewhat contradictory requirements, the chemical parameters must be well balanced. This is achieved by the water chemistry control method called 'normal water chemistry'. Other additional methods are used for the solution to different problems. The 'zinc addition method' is applied to reduce high radiation levels around the recirculation loops. The 'hydrogen water chemistry method' and the 'noble metal chemical addition method' are used to protect the reactor core components and piping made of stainless steel against stress corrosion cracking. This phenomenon has been observed for about 40 years and is partly due to the strong oxidising conditions in the BWR water. Both mitigation methods are used by the majority of the BWR plants all over the world (including the two Swiss NPPs Muehleberg and Leibstadt). (author)

Expert system for maintenance management of a boiling water reactor power plant

International Nuclear Information System (INIS)

Hong Shen; Liou, L.W.; Levine, S.; Ray, A.; Detamore, M.

1992-01-01

An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, which must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components

Nuclear reactor plants and control systems therefor

International Nuclear Information System (INIS)

de Boer, G.A.; de Hex, M.

1976-01-01

A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

Boiling water reactor life extension monitoring

International Nuclear Information System (INIS)

Stancavage, P.

1991-01-01

In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

Advances in Nuclear Power Plant Water Chemistry in Reducing Radiation Exposure

International Nuclear Information System (INIS)

Febrianto

2005-01-01

Water quality in light water reactor in Pressurized Water Reactor as well as in Boiling Water Reactor has being gradually improved since the beginning, to reduce corrosion risk and radiation exposure level. Corrosion problem which occurred to both type of reactors can reduce the plants availability, increase the operation and maintenance cost and increase the radiation exposure. Corrosion and radiation exposure risk in both reactor rare different. BWR type reactor has more experiences in corrosion problem because at the type of reactor lets water to boil in the core, while at PWR type reactor, water is kept not to boil. The BWR reactor has also higher radiation exposure rather than the PWR one. Many collaborative efforts of plants manufacturers and plant operator utilities have been done to reduce the radiation exposure level and corrosion risk. (author)

Areva's water chemistry guidebook with chemistry guidelines for next generation plants (AREVA EPRTM reactors)

International Nuclear Information System (INIS)

Ryckelynck, N.; Chahma, F.; Caris, N.; Guillermier, P.; Brun, C.; Caron-Charles, M.; Lamanna, L.; Fandrich, J.; Jaeggy, M.; Stellwag, B.

2012-09-01

Over the years, AREVA globally has maintained a strong expertise in LWR water chemistry and has been focused on minimizing short-term and long-term detrimental effects of chemistry for startup, operation and shutdown chemistry for all key plant components (material integrity and reliability, promote optimal thermal performances, etc.) and fuel. Also AREVA is focused on minimizing contamination and equipment/plant dose rates. Current Industry Guidelines (EPRI, VGB, etc.) provide utilities with selected chemistry guidance for the current operating fleet. With the next generation of PWR plants (e.g. AREVA's EPR TM reactor), materials of construction and design have been optimized based on industry lessons learned over the last 50+ years. To support the next generation design, AREVA water chemistry experts, have subsequently developed a Chemistry Guidebook with chemistry guidelines based on an analysis of the current international practices, plant operating experience, R and D data and calculation codes now available and/or developed by AREVA. The AREVA LWR chemistry Guidebook can be used to help resolve utility and safety authority questions and addresses regulation requirement questions/issues for next generation plants. The Chemistry Guidebook provides water chemistry guidelines for primary coolant, secondary side circuit and auxiliary systems during startup, normal operation and shutdown conditions. It also includes conditioning and impurity limits, along with monitoring locations and frequency requirements. The Chemistry Guidebook Guidelines will be used as a design reference for AREVA's next generation plants (e.g. EPR TM reactor). (authors)

Potential uses of high gradient magnetic filtration for high-temperature water purification in boiling water reactors

International Nuclear Information System (INIS)

Elliott, H.H.; Holloway, J.H.; Abbott, D.G.

1979-01-01

Studies of various high-temperature filter devices indicate a potentially positive impact for high gradient magnetic filtration on boiling water reactor radiation level reduction. Test results on in-plant water composition and impurity crystallography are presented for several typical boiling water reactors (BWRs) on plant streams where high-temperature filtration may be particularly beneficial. An experimental model on the removal of red iron oxide (hematite) from simulated reactor water with a high gradient magnetic filter is presented, as well as the scale-up parameters used to predict the filtration efficiency on various high temperature, in-plant streams. Numerical examples are given to illustrate the crud removal potential of high gradient magnetic filters installed at alternative stream locations under typical, steady-state, plant operating conditions

Electrochemistry of Water-Cooled Nuclear Reactors

International Nuclear Information System (INIS)

Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy; Amit Jain Han Sang Kim; Vishisht Gupta; Jonathan Pitt

2006-01-01

This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or ''radiation fields'' around the primary loop and the vessel, as a function of the operating parameters and the water chemistry

Data list of nuclear power plants of pressurized-water reactor type in Japan

International Nuclear Information System (INIS)

Izumi, Fumio; Harayama, Yasuo

1981-08-01

This report has collected and compiled the data concerning performances, equipments and installations for nuclear power plants of the pressurized-water reactor type in Japan. The data used in the report are based on informations that were collected before December in 1980. The report is edited by modifing changes of the data appeared after publication of 1979 edition (JAERI-M 8947), and extending the data-package to cover new plants proposed thereafter. All data have been processed and tabulated with a computer program FREP, which has been developed as an exclusive use of data processing. (author)

Hydrogen water chemistry for boiling water reactors

International Nuclear Information System (INIS)

Cowan, R.L.; Cowan, R.L.; Kass, J.N.; Law, R.J.

1985-01-01

Hydrogen Water Chemistry (HWC) is now a practical countermeasure for intergranular stress corrosion cracking (IGSCC) susceptibility of reactor structural materials in Boiling Water Reactors (BWRs). The concept, which involves adding hydrogen to the feedwater to suppress the formation of oxidizing species in the reactor, has been extensively studied in both the laboratory and in several operating plants. The Dresden-2 Unit of Commonwealth Edison Company has completed operation for one full 18-month fuel cycle under HWC conditions. The specifications, procedures, equipment, instrumentation and surveillance programs needed for commercial application of the technology are available now. This paper provides a review of the benefits to be obtained, the side affects, and the special operational considerations needed for commercial implementation of HWC. Technological and management ''Lessons Learned'' from work conducted to date are also described

Reacting flow simulations of supercritical water oxidation of PCB-contaminated transformer oil in a pilot plant reactor

Directory of Open Access Journals (Sweden)

V. Marulanda

2011-06-01

Full Text Available The scale-up of a supercritical water oxidation process, based on recent advancements in kinetic aspects, reactor configuration and optimal operational conditions, depends on the research and development of simulation tools, which allow the designer not only to understand the complex multiphysics phenomena that describe the system, but also to optimize the operational parameters to attain the best profit for the process and guarantee its safe operation. Accordingly, this paper reports a multiphysics simulation with the CFD software Comsol Multiphysics 3.3 of a pilot plant reactor for the supercritical water oxidation of a heavily PCB-contaminated mineral transformer oil. The proposed model was based on available information for the kinetic aspects of the complex mixture and the optimal operational conditions obtained in a lab-scale continuous supercritical water oxidation unit. The pilot plant simulation results indicate that it is not feasible to scale-up directly the optimal operational conditions obtained in the isothermal lab-scale experiments, due to the excess heat released by the exothermic oxidation reactions that result in outlet temperatures higher than 600°C, even at reactor inlet temperatures as low as 400°C. Consequently, different alternatives such as decreasing organic flowrates or a new reactor set-up with multiple oxidant injections should be considered to guarantee a safe operation.

Reactor core cooling device for nuclear power plant

International Nuclear Information System (INIS)

Tsuda, Masahiko.

1992-01-01

The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

ULTRA SCWR+: Practical advanced water reactor concepts

International Nuclear Information System (INIS)

Duffey, Romney; Khartabil, Hussam; Kuran, Sermet; Zhou, Tracy; Pioro, Igor

2008-01-01

Modern thermal power plants now utilize supercritical steam cycles with thermal efficiencies of over 45%. Recent developments have lead to Ultra-SuperCritical (USC) systems, which adopt reheat turbines that can attain efficiencies of over 50%. Because these turbines are already developed, demonstrated and deployed worldwide, and use existing and traditional steam cycle technology, the simplest nuclear advance is to utilize these proven thermal cycle conditions by coupling this turbine type to a reactor. This development direction is fundamentally counter to the usual approach of adopting high-temperature gas-cooled (helium-cooled) reactor cycles, for which turbines have yet to be demonstrated on commercial scale unlike the supercritical steam turbines. The ULTRA (Ultra-supercritical Light water Thermal ReActor) SCWR+ concept adopts the fundamental design approach of matching a water and steam-cooled reactor to the ultra-supercritical steam cycle, adopting the existing and planned thermal power plant turbines. The HP and IP sections are fed with conditions of 25 MPa/625degC and 7 MPa/700degC, respectively, to achieve operating plant thermal efficiencies in excess of 50%, with a direct turbine cycle. By using such low-pressure reheated steam, this concept also adopts technology that was explored and used many years ago in existing water reactors, with the potential to produce large quantities of low cost heat, which can be used for other industrial and district processes. Pressure-Tube (PT) reactors are suitable for adoption of this design approach and, in addition, have other advantages that will significantly improve water-cooled reactor technology. These additional advantages include enhanced safety and improved resource utilization and proliferation resistance. This paper describes the PT-SCWR+ concept and its potential enhancements. (author)

Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

International Nuclear Information System (INIS)

Duce, S.W.; Marley, A.W.; Freeman, A.L.

1989-12-01

Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

Advanced light water reactors for the nineties

International Nuclear Information System (INIS)

Ross, F.A.; Sugnet, W.R.

1987-01-01

The EPRI/Industry advanced light water reactor (ALWR) program and the US Department of Energy ALWR program are closely coordinated to meet the common objective which is the availability of improved and simplified light water reactor plants that may be ordered in the next decade to meet new or replacement capacity requirements. The EPRI/Industry objectives, program participants, and foreign participants, utility requirements document, its organization and content, small plant conceptual design program, the DOE ALWR program, design verification program, General Electric ABWR design features, Combustion Engineering system design, mid-size plant development, General Electric SBWR objectives, Westinghouse/Burns and Roe design objectives, construction improvement, and improved instrumentation and control are discussed in the paper

Once-through cycle, supercritical-pressure light water cooled reactor concept

Energy Technology Data Exchange (ETDEWEB)

Oka, Y.; Koshizuka, S. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab

2001-07-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

Once-through cycle, supercritical-pressure light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Y.; Koshizuka, S.

2001-01-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

Water desalination using different capacity reactors options

International Nuclear Information System (INIS)

Alonso, G.; Vargas, S.; Del Valle, E.; Ramirez, R.

2010-01-01

The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity, cogeneration of potable water production and nuclear electricity is an option to be assessed. In this paper we will perform an economical comparison for cogeneration using a big reactor, the AP1000, and a medium size reactor, the IRIS, both of them are PWR type reactors and will be coupled to the desalination plant using the same method. For this cogeneration case we will assess the best reactor option that can cover both needs using the maximum potable water production for two different desalination methods: Multistage Flash Distillation and Multi-effect Distillation. (authors)

The materials concept in German light water reactors. A contribution to plant safety, economic performance and damage prevention

International Nuclear Information System (INIS)

Ilg, Ulf

2008-01-01

Major decisions taken as early as in the planning and construction phases of nuclear power plants may influence overall plant life. Component quality at the beginning of plant life is determined very much also by a balanced inclusion of the 'design, choice of materials, manufacturing and inspection' elements. One example of the holistic treatment of design, choice of material, and manufacture of important safety-related components in pressurized water reactors is the reactor pressure vessel (RPV) in which the ferritic compound tubes, with inside claddings, for the control rod drive nozzles are screwed into the vessel top. Also the choice of Incoloy 800 for the steam generator tubes, and the design of the main coolant pipes with inside claddings as seamless pipe bends / straight pipes with integrated nozzles connected to mixed welds with austenitic pipes are other special design features of the Siemens/KWU plants. A demonstrably high quality standard by international comparison to this day has been exhibited by the austenitic RPV internals of boiling water reactors, which were made of a low-carbon Nb-stabilized austenitic steel grade by optimum manufacturing technologies. The same material is used for backfitting austenitic pipes. Reliable and safe operation of German nuclear power plants has been demonstrated for more than 4 decades. One major element in this performance is the materials concept adopted in Germany also in the interest of damage prevention. (orig.)

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapter 1, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Jain, Vikas; Nayak, A.; Dhiman, M.; Kulkarni, P. P.; Vijayan, P. K.; Vaze, K. K. [Bhabha Atomic Research Centre, Mumbai (India)

2013-10-15

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

Directory of Open Access Journals (Sweden)

VIKAS JAIN

2013-10-01

Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

RETRAN sensitivity studies of light water reactor transients. Final report

International Nuclear Information System (INIS)

Burrell, N.S.; Gose, G.C.; Harrison, J.F.; Sawtelle, G.R.

1977-06-01

This report presents the results of sensitivity studies performed using the RETRAN/RELAP4 transient analysis code to identify critical parameters and models which influence light water reactor transient predictions. Various plant transients for both boiling water reactors and pressurized water reactors are examined. These studies represent the first detailed evaluation of the RETRAN/RELAP4 transient code capability in predicting a variety of plant transient responses. The wide range of transients analyzed in conjunction with the parameter and modeling studies performed identify several sensitive areas as well as areas requiring future study and model development

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

International Nuclear Information System (INIS)

1989-02-01

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

A capital cost reduction study on the fast breeder reactor plant

International Nuclear Information System (INIS)

Taniyama, H.; Kamei, M.; Moriyama, M.

1991-01-01

A capital cost reduction study has been performed for large fast breeder reactor designs. The primary objective of this study is to show a trend of capital cost reduction between FBR plants at the prototype stage, the demonstration stage, and the future commercialization stage. For the FBR plant at the demonstration stage a construction cost comparison with a light water reactor has also been performed, and the target cost of FBR of below 1.5 times that of the light water reactor cost was achieved. To extend the capital cost reduction study, a feasibility study was made to achieve a capital cost of an FBR less than that of a light water reactor. The recommended design is shown as a future commercialization FBR design concept. (author)

Seawater desalination plant using nuclear heating reactor coupled with MED process

Institute of Scientific and Technical Information of China (English)

无

2000-01-01

A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. This seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. The intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10～200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m3/d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented.

Seawater desalination plant using nuclear heating reactor coupled with MED process

International Nuclear Information System (INIS)

Wu Shaorong; Dong Duo; Zhang Dafang; Wang Xiuzhen

2000-01-01

A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. this seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. the intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10-200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m 3 /d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented

Functional systems of a pressurized water reactor

International Nuclear Information System (INIS)

Heinzel, V.

1982-01-01

The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

International Nuclear Information System (INIS)

Bozoki, G.; Kohut, P.; Fitzpatrick, R.

1989-02-01

This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs

SBWR: A simplified boiling water reactor

International Nuclear Information System (INIS)

Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

1987-01-01

An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

International Nuclear Information System (INIS)

Woo, H.H.; Lu, S.C.

1981-01-01

Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design

Materials for advanced water cooled reactors

International Nuclear Information System (INIS)

1992-09-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

US Advanced Light Water Reactor Program; overall objective

International Nuclear Information System (INIS)

Klug, N.

1989-01-01

The overall objective of the US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) program is to perform coordinated programs of the nuclear industry and DOE to insure the availability of licensed, improved, and simplified light water reactor standard plant designs that may be ordered in the 1990's to help meet the US electrical power demand. The discussion includes plans to meet program objectives and the design certification program. DOE is currently supporting the development of conceptual designs, configurations, arrangements, construction methods/plans, and proof test key design features for the General Electric ASBWR and the Westinghouse AP600. Key features of each are summarized. Principal milestones related to licensing of large standard plants, simplified mid-size plant development, and plant lifetime improvement are noted

Design of virtual SCADA simulation system for pressurized water reactor

International Nuclear Information System (INIS)

Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

2016-01-01

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

Design of virtual SCADA simulation system for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

2016-02-08

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

Application of expert system to evaluating reactor water cleanup system performance

International Nuclear Information System (INIS)

Maeda, Katsuji; Nakamura, Masahiro; Nagasawa, Katsumi; Fushiki, Sumiyuki.

1991-01-01

Expert systems employing artificial intelligence (AI) have been developed for finding and elucidating causes of anomalies and malfunctions, presenting pertinent recommendation for countermeasures and for making precautionary diagnosis. On the other hand, further improvements in reliabilities for chemical control are required to promote BWR plant reliability and advancement. Especially, it is necessary to maintain the reactor water purity in high quality to minimize stress corrosion cracking (SCC) in primary cooling system, fuel performance degradation and radiation buildup. The reactor water quality is controlled by the reactor water cleanup (RWCU) system. So, it is very important to maintain the RWCU performance, in order to keep good reactor water quality. This paper describes an expert system used for evaluating RWCU system performance in BWR plants. (author)

Materials technologies of light water reactors

International Nuclear Information System (INIS)

Begley, R.

1984-01-01

Satisfactory materials performance is a key element in achieving reliable operation of light water reactors. Outstanding performance under rigorous operational conditions has been exhibited by pressure boundary components, core internals, fuel cladding, and other critical components of these systems. Corrosion and stress corrosion phenomena have, however, had an impact on plant availability, most notably relating to pipe cracking in BWR systems and steam generator corrosion in PWR systems. These experiences have stimulated extensive development activities by the nuclear industry in improved NDE techniques, investigation of corrosion phenomena, as well as improved materials and repair processes. This paper reviews key materials performance aspects of light water reactors with particular emphasis on the progress which has been made in modeling of corrosion phenomena, control of the plant operating environment, advanced material development, and application of sophisticated repair procedures. Implementation of this technology provides the basis for improved plant availability

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapters 2--13, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

Analysis of reactor trips involving balance-of-plant failures

International Nuclear Information System (INIS)

Seth, S.; Skinner, L.; Ettlinger, L.; Lay, R.

1986-01-01

The relatively high frequency of plant transients leading to reactor trips at nuclear power plants in the US is of economic and safety concern to the industry. A majority of such transients is due to failures in the balance-of-plant (BOP) systems. As a part of a study conducted for the US Nuclear Regulatory Commission, Mitre has carried out a further analysis of the BOP failures associated with reactor trips. The major objectives of the analysis were to examine plant-to-plant variations in BOP-related trips, to understand the causes of failures, and to determine the extent of any associated safety system challenges. The analysis was based on the Licensee Event Reports submitted on all commercial light water reactors during the 2-yr period, 1984-1985

Adapting a reactor safety assessment system for specific plants

International Nuclear Information System (INIS)

Ballard, T.L.; Cordes, G.A.

1991-01-01

The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

Replacement of outboard main steam isolation valves in a boiling water reactor plant

Energy Technology Data Exchange (ETDEWEB)

Schlereth, J.R.; Pennington, D.

1996-12-01

Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

Replacement of outboard main steam isolation valves in a boiling water reactor plant

International Nuclear Information System (INIS)

Schlereth, J.R.; Pennington, D.

1996-01-01

Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it's Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components

Assessment of EPRI water chemistry guidelines for new nuclear power plants

International Nuclear Information System (INIS)

Reid Richard; Kim Karen; McCree, Anisa; Eaker, Richard; Sawochka, Steve; Giannelli, Joe

2012-09-01

Water chemistry control technologies for nuclear power plants have been significantly enhanced over the past few decades to improve material and equipment reliability and fuel performance, and to minimize radionuclide production and transport. Chemistry Guidelines have been developed by the Electric Power Research Institute (EPRI) for currently operating plants and have been intermittently revised over the past twenty-five years for the protection of systems and components and for radiation management. As new plants are being designed for improved safety and increased power production, it is important to ensure that the designs consider implementation of state-of-the-art, industry developed water chemistry controls. In parallel, the industry will need to consider and update water chemistry guidelines as well as plant startup and operational strategies based on the advanced plant designs. EPRI has performed assessments of water chemistry control guidance or assumptions provided in design and licensing documents for several advanced plant designs. These designs include: Westinghouse AP1000 Pressurized Water Reactor AREVA US-EPR Pressurized Water Reactor Mitsubishi Nuclear Energy Systems/Mitsubishi Heavy Industries Advanced Pressurized Water Reactor Korea Hydro and Nuclear Power APR1400 Pressurized Water Reactor Toshiba Advanced Boiling Water Reactor (ABWR) General Electric-Hitachi Economic Simplified Boiling Water Reactor (ESBWR) The intent of these assessments was to identify key design differences in each of the new plant designs relative to the current operating fleet and to identify differences in water chemistry specifications or design assumptions provided in design and licensing documents for the plants in comparison to current EPRI Water Chemistry Guidelines. This paper provides a summary of the key results of these assessments. The fundamental design and operation of the advanced plants is similar to the currently operating fleet. As such, the new plants are

Integrated plant safety assessment: Systematic Evaluation Program. LaCrosse Boiling Water Reactor, Dairyland Power Cooperative, Docket No. 50-409

International Nuclear Information System (INIS)

1983-04-01

The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the La Crosse Boiling Water Reactor, operated by Dairyland Power Cooperative. The La Crosse plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addresed. Equipment and procedural changes have been identified as a result of the review

Systems design of direct-cycle supercritical-water-cooled fast reactors

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

1995-01-01

The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

Advanced light water reactor plants System 80+trademark design certification program. Annual progress report, October 1, 1994 - September 30, 1995

International Nuclear Information System (INIS)

1998-01-01

The purpose of this report is to provide the status of the progress that was made towards Design Certification of System 80+trademark during the US government's 1995 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2, and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems

Advanced light water reactor plants System 80+trademark design certification program. Annual progress report, October 1, 1995 - September 30, 1996

International Nuclear Information System (INIS)

1996-01-01

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+trademark during the US government's 1996 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2 and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems

Method and plant to remote tritium from the cooling water of a nuclear reactor

International Nuclear Information System (INIS)

O'Brien, C.J.

1976-01-01

A method is proposed for the extraction of tritium from the cooling water of a nuclear reactor, based on the principle of concentrating the tritium by a multi-stage transfer process. The cooling water is brought into contact in each stage with basic, labile, hydrogen-containing material with high pH value, whereby the tritium is transfered into an intermediate solid product and can be separated off. The technical details of the plant are described. Cellulose materials, such as cotton and wood as well as protein-containing material, such as muscle tissue are mentioned as examples of materials with a high affinity to tritium, greater than the affinity of water to tritium. They extract tritium from the cooling water. (HK) [de

Fuel Summary Report: Shippingport Light Water Breeder Reactor

International Nuclear Information System (INIS)

Illum, D.B.; Olson, G.L.; McCardell, R.K.

1999-01-01

The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749

Systematic methodology for diagnosis of water hammer in LWR power plants

International Nuclear Information System (INIS)

Safwat, H.H.; Arastu, A.H.; Husaini, S.M.

1990-01-01

The paper gives the dimensions of the knowledge base that is necessary to carry out a diagnosis of water hammer susceptibility/root cause analyses for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) nuclear power plant systems. After introducing some fundamentals, water hammer phenomena are described. Situations where each phenomenon is encountered are given and analytical models capable of simulating the phenomena are referenced. Water hammer events in operating plants and their inclusion in the knowledge base is discussed. The diagnostic methodology is presented through an application on a system in a typical light water reactor plant. The methodology presented serves as a possible foundation for the creation of an expert water hammer diagnosis system. (orig.)

Safety considerations concerning light water reactors in Sweden

International Nuclear Information System (INIS)

Nilsson, T.

1977-01-01

In 1975 the Swedish Nuclear Power Inspectorate was commissioned by the Government to perform a Reactor Safety Study concerning commercial light water reactors. The study will contain an account of: - rules and regulations for reactor designs; - operation experience of the Swedish nuclear power plants with international comparisons; - the development of reactor designs during the last 10 years; - demands and conditions for inspection and inspection methods; - nuclear power plant operation organization; - training of operators; and - the results of research into nuclear safety. The study is scheduled for completion by July 1st, 1977, however, this paper gives a summary of the results of the Reactor Safety Study already available. The paper contains detailed statistics concerning safety related occurrences and reactor scrams in Sweden from July 1st, 1974 until the beginning of 1977

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

Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

International Nuclear Information System (INIS)

Ware, A.G.

1988-01-01

This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

Chemistry of the water in thermal power plants

International Nuclear Information System (INIS)

Freier, R.K.

1984-01-01

This textbook and practical manual gives a comprehensive review of the scientific knowledge of water as operating substance and of the chemistry of water in thermal power plants. The fundamentals of water chemistry and of the conventional and nuclear water/steam circuit are described. The contents of the chapters are: 1. The atom, 2. The chemical bond, 3. The dissolving capacity of water, 4. Operational parameters and their measurement, 5. Corrosion, 6. The water/steam coolant loop of conventional plants (WSC), 7. The pressurized water reactor (PWR), 8. The boiling water reactor (BWR), 9. The total and partial desalination properties of ion exchangers, 10. The cooling water, 11. The failure of Harrisburg in a simple presentation. (HK) [de

Study on dual plant concept for the next generation boiling water reactors

International Nuclear Information System (INIS)

Sato, Takashi; Oikawa, Hirohide

1999-01-01

The paper presents the study results on the basic concept of dual BWRs. For the convenience, we call the concept here as Trial Study on BWR dual concept (TSBWR dual). The concept is general and applicable to all BWRs which have internal recirculation pumps (RIP). The TSBWR dual is a plant concept of dual BWRs contained in a same secondary containment building. The plant output is from 2 x l,350 MWe up to 2 x 1,700 MWe. This concept is mainly aiming at safety improvement and cost savings of the next generation BWRs. The TSBWR dual has two RPVs and two dry wells (DW). It has, however, only one wet well (WW) and only one R/B. The WW and the R/B are shared by the dual reactors. The operating floor is also shared by the two reactors. The TSBWR dual has both passive safety systems and active safety systems. They are also shared between the two reactors. A lot of sharing between the dual reactors enables significant cost savings accompanied by the power increase up to 3,400 MWe. Although the TSBWR dual consists of two reactors, the simplified cylindrical configuration of the key structures and reduction of the R/B height can minimize the plant construction period. The TSBWR dual provides a concept with which we can challenge to construct a dual BWR plant in the near future. (author)

Coagulant recovery from water treatment plant sludge and reuse in post-treatment of UASB reactor effluent treating municipal wastewater.

Science.gov (United States)

Nair, Abhilash T; Ahammed, M Mansoor

2014-09-01

In the present study, feasibility of recovering the coagulant from water treatment plant sludge with sulphuric acid and reusing it in post-treatment of upflow anaerobic sludge blanket (UASB) reactor effluent treating municipal wastewater were studied. The optimum conditions for coagulant recovery from water treatment plant sludge were investigated using response surface methodology (RSM). Sludge obtained from plants that use polyaluminium chloride (PACl) and alum coagulant was utilised for the study. Effect of three variables, pH, solid content and mixing time was studied using a Box-Behnken statistical experimental design. RSM model was developed based on the experimental aluminium recovery, and the response plots were developed. Results of the study showed significant effects of all the three variables and their interactions in the recovery process. The optimum aluminium recovery of 73.26 and 62.73 % from PACl sludge and alum sludge, respectively, was obtained at pH of 2.0, solid content of 0.5 % and mixing time of 30 min. The recovered coagulant solution had elevated concentrations of certain metals and chemical oxygen demand (COD) which raised concern about its reuse potential in water treatment. Hence, the coagulant recovered from PACl sludge was reused as coagulant for post-treatment of UASB reactor effluent treating municipal wastewater. The recovered coagulant gave 71 % COD, 80 % turbidity, 89 % phosphate, 77 % suspended solids and 99.5 % total coliform removal at 25 mg Al/L. Fresh PACl also gave similar performance but at higher dose of 40 mg Al/L. The results suggest that coagulant can be recovered from water treatment plant sludge and can be used to treat UASB reactor effluent treating municipal wastewater which can reduce the consumption of fresh coagulant in wastewater treatment.

Development of next-generation light water reactor

International Nuclear Information System (INIS)

Ishibashi, Fumihiko; Yasuoka, Makoto

2010-01-01

The Next-Generation Light Water Reactor Development Program, a national project in Japan, was inaugurated in April 2008. The primary objective of this program is to meet the need for the replacement of existing nuclear power plants in Japan after 2030. With the aim of setting a global standard design, the reactor to be developed offers greatly improved safety, reliability, and economic efficiency through several innovative technologies, including a reactor core system with uranium enrichment of 5 to 10%, a seismic isolation system, long-life materials, advanced water chemistry, innovative construction techniques, optimized passive and active safety systems, innovative digital technologies, and so on. In the first three years, a plant design concept with these innovative features is to be established and the effectiveness of the program will be reevaluated. The major part of the program will be completed in 2015. Toshiba is actively engaged in both design studies and technology development as a founding member of this program. (author)

Operation management of the prototype heavy water reactor 'Fugen'

International Nuclear Information System (INIS)

Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi

1983-01-01

The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported. (Kako, I.)

Design measures in evolutionary water cooled reactors to optimize for economic viability

International Nuclear Information System (INIS)

Oh, S.J.; Yu, S.K.W.; Appell, B.

1999-01-01

Since the mid 1980s, there have been various efforts to develop evolutionary water cooled reactors based on the current operating plant experience. To sustain and improve the economic viability, particular attention has been paid to the following aspects in developing evolutionary water cooled reactors: design simplification and increased operating margins, standardization in design as well as construction and operation, integration of operating plant insights, and consideration of safety, operability and constructability during the design stage. This paper reviews each item and discusses several examples from some of the evolutionary water cooled reactors being developed. (author)

Regulatory requirements for desalination plant coupled with nuclear reactor plant

International Nuclear Information System (INIS)

Yune, Young Gill; Kim, Woong Sik; Jo, Jong Chull; Kim, Hho Jung; Song, Jae Myung

2005-01-01

A small-to-medium sized reactor has been developed for multi-purposes such as seawater desalination, ship propulsion, and district heating since early 1990s in Korea. Now, the construction of its scaled-down research reactor, equipped with a seawater desalination plant, is planned to demonstrate the safety and performance of the design of the multi-purpose reactor. And the licensing application of the research reactor is expected in the near future. Therefore, a development of regulatory requirements/guides for a desalination plant coupled with a nuclear reactor plant is necessary for the preparation of the forthcoming licensing review of the research reactor. In this paper, the following contents are presented: the design of the desalination plant, domestic and foreign regulatory requirements relevant to desalination plants, and a draft of regulatory requirements/guides for a desalination plant coupled with a nuclear reactor plant

A review of the UKAEA interest in heavy water reactors

International Nuclear Information System (INIS)

Symes, R.J.

1983-01-01

The chapter commences with a brief account of the history of heavy water production and then begins the story of the British use of this moderator in power reactors. This is equated with the introduction and development of the tube reactor as a distinct and important form of reactor construction in contrast with the perhaps better known vessel design that has tended to dominate reactor engineering to date. The account thus includes a succession of reactor designs including the gas and steam cooled heavy water systems in addition to the steam-generating heavy water reactor. The SGHWR was demonstrated by the construction of a substantial prototype, which continues in operation as a flexible and reliable electricity-generating plant. It was also, for a time, identified as the system to be used for Britain's third reactor programme. Today the successful Canadian CANDU power reactors represent the only penetration of heavy water reactor technology into large scale electricity generation. The range of research and experimental reactors using heavy water in their cores is reviewed. (author)

Graphite-water steam-generating reactor in the USSR

Energy Technology Data Exchange (ETDEWEB)

Dollezhal, N A [AN SSSR, Moscow

1981-10-01

One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

Core design concepts for high performance light water reactors

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.

2007-01-01

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

Coupling of RO-MSF hybrid desalination plants with nuclear reactors

International Nuclear Information System (INIS)

Al-Sulaiman, Khalil; Al-Mutaz, Ibrahim S.

1999-01-01

Full text.Reverse osmosis (RO) and multistage flash (MSF) desalination are the most widely commercial available processes. MSF utilizes stream in the brine heater as a primary source of energy. RO is derived mainly by electricity that pumps the feed water against the mambranes. Steam and electricity and be produced easily by nuclear reactors. Nuclear reactors may be coupled with deslination plants (MSF, RO or combined (hybrid) RO/MSF configuration). This integrated plant will be capable of producing power and water at reasonable cost. The capital and operating cost will be reduced and the excess power can be efficiently utilized. Maintenance and operating cost will drop significantly. In this paper, a techno-economic study of hybrid reverses osmosis /multistage flash desalination will be carried. The proposed configuration (hybrid RO/MSF) coupled with nuclear reactor is considered the most appropriate candidate system for the application of dual-purpose nuclear desalination plants. the design parameters for such a desalination hybrid system will be the applied pressure and recovery for reverse osmosis plant and the number of stages and the heat transfer areas for multistage flash plant

Analysis of water hammer phenomena in RBMK-1500 reactor main circulation circuit

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.; Vaisnoras, M.

2006-01-01

Water hammer can occur in any thermal-hydraulic systems. Water hammer can reach pressure levels far exceeding the pressure range of a pipe given by the manufacturer, and it can lead to the failure of the pipeline integrity. In the past three decades, since a large number of water hammer events occurred in the light-water- reactor power plants, a number of comprehensive studies on the phenomena associated with water hammer events have been performed. There are three basic types of severe water hammer occurring at power plants that can result in significant plant damage: rapid valve operation events; void-induced water hammer; condensation-induced water hammer. Correct prediction of water hammer transients, is therefore of paramount importance for the safe operation of the plant. Therefore verifying of computer codes capability to simulate water hammer type transients is very important issue at performing of safety analyses for nuclear power plants. Verification of RELAP5/MOD3.3 code capability to simulate water hammer type transients employing the experimental investigations is presented. Experience gained from benchmarking analyses has been used at development of the detail RELAP5 code RBMK-1500 model for simulation of water hammer effects in reactor main circulation circuit. Analysis of reactor cooling system shows, that water hammers can occur in main circulation circuit of RBMK-1500 reactor in cases of: (1) Guillotine break of the inlet piping upstream of the Group Distribution Header and (2) Guillotine break of the pressure piping upstream the Main Circulation Pump check valve. Analysis of above mentioned accident scenarios is presented in this paper. First scenario of the accident potentially is more dangerous, because the pressure pulses influence not only the reactor cooling circuit, but also the piping of safety related system (Emergency Core Cooling System pipeline) connected to affected Group Distribution Header. The performed analysis using RELAP5 code

Fusion power plant for water desalination and reuse

International Nuclear Information System (INIS)

Borisov, A.A.; Desjatov, A.V.; Izvolsky, I.M.; Serikov, A.G.; Smirnov, V.P.; Smirnov, Yu.N.; Shatalov, G.E.; Sheludjakov, S.V.; Vasiliev, N.N.; Velikhov, E.P.

2001-01-01

Development of industry and agriculture demands a huge fresh water consumption. Exhaust of water sources together with pollution arises a difficult problem of population, industry, and agriculture water supply. Request for additional water supply in next 50 years is expected from industrial and agricultural sectors of many countries in the world. The presented study of fusion power plant for water desalination and reuse is aimed to widen a range of possible fusion industrial applications. Fusion offers a safe, long-term source of energy with abundant resources and major environmental advantages. Thus fusion can provide an attractive energy option to society in the next century. Fusion power tokamak reactor based on RF DEMO-S project [Proc. ISFNT-5 (2000) in press; Conceptual study of RF DEMO-S fusion reactor (2000)] was chosen as an energy source. A steady state operation mode is considered with thermal power of 4.0 GW. The reactor has to operate in steady-state plasma mode with high fraction of bootstrap current. Average plant availability of ∼0.7 is required. A conventional type of water cooled blanket is the first choice, helium or lithium coolants are under consideration. Desalination plant includes two units: reverse osmosis and distillation. Heat to electricity conversion schemes is optimized fresh water production and satisfy internal plant electricity demand The plant freshwater capacity is ∼6000000 m 3 per day. Fusion power plant of this capacity can provide a region of a million populations with fresh water, heat and electricity

Fusion power plant for water desalination and reuse

Energy Technology Data Exchange (ETDEWEB)

Borisov, A.A.; Desjatov, A.V.; Izvolsky, I.M.; Serikov, A.G.; Smirnov, V.P.; Smirnov, Yu.N.; Shatalov, G.E.; Sheludjakov, S.V.; Vasiliev, N.N. E-mail: vasiliev@nfi.kiae.ru; Velikhov, E.P

2001-11-01

Development of industry and agriculture demands a huge fresh water consumption. Exhaust of water sources together with pollution arises a difficult problem of population, industry, and agriculture water supply. Request for additional water supply in next 50 years is expected from industrial and agricultural sectors of many countries in the world. The presented study of fusion power plant for water desalination and reuse is aimed to widen a range of possible fusion industrial applications. Fusion offers a safe, long-term source of energy with abundant resources and major environmental advantages. Thus fusion can provide an attractive energy option to society in the next century. Fusion power tokamak reactor based on RF DEMO-S project [Proc. ISFNT-5 (2000) in press; Conceptual study of RF DEMO-S fusion reactor (2000)] was chosen as an energy source. A steady state operation mode is considered with thermal power of 4.0 GW. The reactor has to operate in steady-state plasma mode with high fraction of bootstrap current. Average plant availability of {approx}0.7 is required. A conventional type of water cooled blanket is the first choice, helium or lithium coolants are under consideration. Desalination plant includes two units: reverse osmosis and distillation. Heat to electricity conversion schemes is optimized fresh water production and satisfy internal plant electricity demand The plant freshwater capacity is {approx}6000000 m{sup 3} per day. Fusion power plant of this capacity can provide a region of a million populations with fresh water, heat and electricity.

Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1994--September 30, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

The purpose of this report is to provide the status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1995 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2, and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems.

Heavy water cycle in the CANDU reactor

International Nuclear Information System (INIS)

Nanis, R.

2000-01-01

Hydrogen atom has two isotopes: deuterium 1 H 2 and tritium 1 H 3 . The deuterium oxide D 2 O is called heavy water due to its density of 1105.2 Kg/m 3 . Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D 2 O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D 2 O management is required to preserve it. (author)

European supercritical water cooled reactor

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.; Marsault, P.; Bittermann, D.; Maraczy, C.; Laurien, E.; Lycklama a Nijeholt, J.A.; Anglart, H.; Andreani, M.; Ruzickova, M.; Toivonen, A.

2011-01-01

Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 o C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-08-01

This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

International Nuclear Information System (INIS)

1995-08-01

This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory

Water cooled reactor technology: Safety research abstracts no. 1

International Nuclear Information System (INIS)

1990-01-01

The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

The future 700 MWe pressurized heavy water reactor

International Nuclear Information System (INIS)

Bhardwaj, S.A.

2006-01-01

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

A control system for industrial plant (e.g. a pressurized water reactor)

International Nuclear Information System (INIS)

Spiller, C.R.L.

1990-01-01

A control system for an industrial plant eg. a pressurised water nuclear reactor, comprises a plurality of instrument sets and a plurality of logic sets. The instrument sets have a number of sensors which detect parameters (temperature, pressure vibration) of the industrial plant, and have two serial link controllers which supply the output signals from each sensor in the instrument set sequentially to the logic sets via conductors. The logic sets have a number of auto select logic circuits, each of which selects data from the sensors from one of the instrument sets, and a synchroniser ensures that the output signals from the sensors detecting the same parameter are supplied to a voting logic circuit at the same time. The voting logic circuit performs a voting function on the output signals to produce a series of high reliability signals which are converted to parallel high reliability signals by a series to a parallel converter. The high reliability signals are supplied to a fault logic shutdown circuit which controls the operation of shutdown mechanisms for the industrial plant. (author)

A Robust Multivariable Feedforward/Feedback Controller Design for Integrated Power Control of Boiling Water Reactor Power Plants

International Nuclear Information System (INIS)

Shyu, S.-S.; Edwards, Robert M.

2002-01-01

In this paper, a methodology for synthesizing a robust multivariable feedforward/feedback control (FF/FBC) strategy is proposed for an integrated control of turbine power, throttle pressure, and reactor water level in a nuclear power plant. In the proposed method, the FBC is synthesized by the robust control approach. The feedforward control, which is generated via nonlinear programming, is added to the robust FBC system to further improve the control performance. The plant uncertainties, including unmodeled dynamics, linearization, and model reduction, are characterized and estimated. The comparisons of simulation responses based on a nonlinear reactor model demonstrate the achievement of the proposed controller with specified performance and endurance under uncertainty. It is also important to note that all input variables are manipulated in an orchestrated manner in response to a single output's setpoint change

Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

International Nuclear Information System (INIS)

Goldberg, A.; Streit, R.D.

1981-05-01

Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads

Development of light water reactors and subjects for hereafter

International Nuclear Information System (INIS)

Murao, Yoshio

1995-01-01

As for light water reactors, the structure is relatively simple, and the power plants of large capacity can be realized easily, therefore, they have been used for long period as main nuclear reactors. During that period, the accumulation of experiences on the design, manufacture, operation, maintenance and regulation of light water has become enormous, and in Japan, the social base for maintaining and developing light water reactor technologies has been prepared sufficiently. If the nuclear power generation using seawater uranium is considered, the utilization of uranium for light water reactor technologies can become the method of producing the own energy for Japan. As the factors that threaten the social base of light water reactor technologies, there are a the lowering of the desire to promote light water reactors, the effect of secular deterioration, the price rise of uranium resources, the effect of plutonium accumulation, the effect of the circumstances in developing countries and the sure recruiting of engineers. The construction and the principle of working of light water reactors and the development of light water reactors hereafter, for example, the improvement on small scale and the addition of new technology resulting in cost reduction and the lowering of the quality requirement for engineers, the improvement of core design, the countermeasures by design to serious accidents and others are described. (K.I.)

Economic competitiveness requirements for evolutionary water cooled reactors

International Nuclear Information System (INIS)

Hudson, C.R.; Bertel, E.; Paik, K.H.; Roh, J.H.; Tort, V.

1999-01-01

This paper analyses the necessary economic conditions for evolutionary water cooled reactors to be competitive. Utilising recent national cost data for fossil-fired base load plants expected to be commissioned by 2005 -2010, target costs for nuclear power plants are discussed. Factors that could contribute to the achievement of those targets by evolutionary water cooled reactors are addressed. The feed-back from experience acquired in implementing nuclear programmes is illustrated by some examples from France and the Republic of Korea. The paper discusses the impacts on nuclear power competitiveness of globalisation and deregulation of the electricity market and privatisation of the electricity sector. In addition, issues related to external cost internalisation are considered. (author)

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors

International Nuclear Information System (INIS)

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.; Trapp, T.J.; Spinrad, B.I.

1981-09-01

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are too costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized

A decision support system for maintenance management of a boiling-water reactor power plant

International Nuclear Information System (INIS)

Shen, J.H.; Ray, A.; Levin, S.

1996-01-01

This article reports the concept and development of a prototype expert system to serve as a decision support tool for maintenance of boiling-water reactor (BWR) nuclear power plants. The code of the expert system makes use of the database derived from the two BWR units operated by the Pennsylvania Power and Light Company in Berwick, Pennsylvania. The operations and maintenance information from a large number of plant equipment and sub-systems that must be available for emergency conditions and in the event of an accident is stored in the database of the expert system. The ultimate goal of this decision support tool is to identify the relevant Technical Specifications and management rules for shutting down any one of the plant sub-systems or removing a component from service to support maintenance. 6 refs., 7 figs

Analysis of scrams and forced outages at boiling water reactors

International Nuclear Information System (INIS)

Earle, R.T.; Sullivan, W.P.; Miller, K.R.; Schwegman, W.J.

1980-07-01

This report documents the results of a study of scrams and forced outages at General Electric Boiling Water Reactors (BWRs) operating in the United States. This study was conducted for Sandia Laboratories under a Light Water Reactor Safety Program which it manages for the United States Department of Energy. Operating plant data were used to identify the causes of scrams and forced outages. Causes of scrams and forced outages have been summarized as a function of operating plant and plant age and also ranked according to the number of events per year, outage time per year, and outage time per event. From this ranking, identified potential improvement opportunities were evaluated to determine the associated benefits and impact on plant availability

Supercritical Water Reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

2014-01-01

The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

Water quality maintaining device of power plant

International Nuclear Information System (INIS)

Kobayashi, Minoru; Inami, Ichiro.

1994-01-01

The device of the present invention reduces the amount of leaching materials of ion exchange resins from a water processing system of a BWR tyep plant, improves the water quality of reactor water to maintain the water at high purity. That is, steams used for power generation are condensated in a condensate system. A condensate filter and a condensate desalter for cleaning the condensates are disposed. A resin storage hopper is disposed for supplying the ion exchange resins to the water processing system. A device for supplying a nitrogen gas or an inert gas is disposed in the hopper. With such a constitution, the ion exchange resins in the water processing system are maintained in a nitrogen gas or inert gas atmosphere or at a low dissolved oxygen level in an operation stage in the power plant. Accordingly, degradation of the ion exchange resins in the water processing system is suppressed and the amount of the leaching material from the resins is reduced. As a result, the amount of the resins leached into the reactor is reduced, so that the reactor water quality can be maintained at high purity. (I.S.)

Integrated plant safety assessment. Systematic Evaluation Program. La Crosse Boiling Water Reactor. Dairyland Power Cooperative, Docket No. 50-409. Final report

International Nuclear Information System (INIS)

1983-06-01

The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the La Crosse Boiling Water Reactor, operated by Dairyland Power Cooperative. The La Crosse plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review

Research and development of the supercritical-pressure light water cooled reactor

International Nuclear Information System (INIS)

Oka, Yoshiaki

2003-01-01

The concept of high temperature reactor cooled by light water (SCR) has been developed at the University of Tokyo since 1989. Major elements of reactor conceptual design and safety were studied. It includes fuel rod design, core design of thermal and fast reactors, plant heat balance, safety design, accident and transient analysis, LOCA, PSA, plant control, start-up and stability. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical FPP in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil Fired Power Plants (FPP) will be fully utilized for SCR. Although the concept was developed at the University of Tokyo mostly with our own funds and resources, four funding was/is provided for the research in Japan so far. Those are TEPCO studies with Japanese vendors in 1994 and 1995. JSPS (Monbusho) funding of pulse radiolysis of supercritical water to the University of Tokyo, Japanese-NERI program of METI to Toshiba team on thermal hydraulics, corrosion and plant system and Japanese-NERI program of MEXT on water chemistry to the University of Tokyo. The concept was taken as the reference of HPLWR study in Europe with funding of EU in 2000 and 2001. The concept was evaluated in the Generation 4 reactor program in USA. It was selected as only one water-cooled Generation 4 reactor. This paper describes the overview of the conceptual design at the University of Tokyo and R and D in the world

Towards intrinsically safe light-water reactors

International Nuclear Information System (INIS)

Hannerz, K.

1983-02-01

The reactor-safety issue is one of the principal problems threatening the future of the nuclear option, at least in participatory democracies. It has contributed to widespread public distrust and is the direct cause of the escalation in design complexity and quality assurance requirements that are rapidly eroding the competitive advantage of nuclear power. Redesign of the light-water reactor can eliminate those features that leave it open to public distrust and obstructive intervention. This redesign appears feasible within the realm of proven technology in those fields (fuels, materials, water chemistry, waste technology, etc.) in which extended operating experience is essential for confidence in system performance. A pressurized water reactor outline design developed to achieve the above goal is presented. The key feature is the design of the primary system extracting heat from the core so that the latter is protected from damage caused by any credible system failure or any destructive intervention from the outside by either violent means (up to and including nonnuclear warfare) or by mistaken or malicious use of the plant control systems. Such a design objective can be achieved by placing the entire primary circulation system in a large pressurized pool of cold water with a high boric acid content. Enough water is provided in the pool to allow core-decay-heat removal by evaporation for at least one week following any incident with no cooling systems operating. Subsequently it is assumed that a supply of further water (a few cubic meters per hour) from the outside can be arranged, even without the presence of the plant operating personnel

Reactor plant for Belene NPP completion

International Nuclear Information System (INIS)

Dragunov, Yu. G.; Ryzhov, S. B.; Ermakov, D. N.; Repin, A. I.

2004-01-01

Construction of 'Belene' NPP was started at the end of 80-ties using project U-87 with V-320 reactor plant, general designer of this plant is OKB 'Gidropress'. At the beginning of 90-ties, on completing the considerable number of deliveries and performance of civil engineering work at the site the NPP construction was suspended. Nowadays, considering the state of affairs at the site and the work performed by Bulgarian Party on preservation of the equipment delivered, the most perspective is supposed to be implementation of the following versions in completing 'Belene' NPP: for completion of Unit 1 - reactor plant VVER-1000 on the basis of V-320 reactor with the maximum use of the delivered equipment (V-320M) having the extended service life and safety improvement; for Unit 2 - advanced reactor plant VVER-1000. For the upgraded reactor plant V-230M the basic solutions and characteristics are presented, as well as the calculated justification of strength and safety analyses, design of the reactor core and fuel cycle, instrumentation and control systems, application of the 'leak-before break' in the project and implementation of safety measures. For the modernised reactor plant V-392M the main characteristics and basic changes are presented, concerning reactor pressure vessel, steam generator, reactor coolant pump set. Design of NPP with the modernized reactor plant V-320M meets the up-to-date requirements and can be licensed for completion and operation. In the design of NPP with the advanced reactor plant the basic solutions and the equipment are used that are similar to those used in standard reactor plant V-320 and new one with VVER-1000 under construction and completion in Russia, and abroad. Compliance of reactor design with the up-to-date international requirements, considering the extended service life of the main equipment, shows its rather high potential for implementation during completion of 'Belene' NPP

Light water reactor safeguards system evaluation

International Nuclear Information System (INIS)

Varnado, G.B.; Ericson, D.M. Jr.; Bennett, H.A.; Hulme, B.L.; Daniel, S.L.

1978-01-01

A methodology for assessing the effectiveness of safeguards systems was developed in this study and was applied to a typical light water reactor plant. The relative importance of detection systems, barriers, response forces and other safeguards system components was examined in extensive parameter variation studies. (author)

Development of Next-Generation LWR (Light Water Reactor) in Japan

International Nuclear Information System (INIS)

Yamamoto, T.; Kasai, S.

2011-01-01

The Next-Generation Light Water Reactor development program was launched in Japan in April 2008. The primary objective of the program is to cope with the need to replace existing nuclear power plants in Japan after 2030. The reactors to be developed are also expected to be a global standard design. Several innovative features are envisioned, including a reactor core system with uranium enrichment above 5%, a seismic isolation system, the use of long-life materials and innovative water chemistry, innovative construction techniques, safety systems with the best mix of passive and active concepts, and innovative digital technologies to further enhance reactor safety, reliability, economics, etc. In the first 3 years, a plant design concept with these innovative features is established and the effectiveness of the program is reevaluated. The major part of the program will be completed in 2015. (author)

Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

International Nuclear Information System (INIS)

Boldyrev, V.M.; Mikhaj, V.I.

1985-01-01

Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

Experiences on reduction of reactor water silica and fresh resin leaching organics for Kuo-Sheng Nuclear Power Plant

International Nuclear Information System (INIS)

Wen, T-J.; Wang, C-H.

2010-01-01

The silica level in reactor water of Kuo-Sheng nuclear power plants has been slowly increased from 200 ppb to the high level above 500 ppb in recent years. The results obtained from steam/liquid mass balance calculation indicated that an increase of reactor water silica was mainly caused by continuing equilibrium leakage from deep bed condensate demineralizers, where the ion exchange zone was periodically disturbed by resin backwashing - scrubbing operation. The fastest and the most effective way to reduce the silica inventory in reactor system is to operate by continuously precoating of two sets of the reactor water clean up filter demineralizers to a lower effluent silica end point, and perhaps as frequently as three or four days. Leaching organic contaminants into feed water from the ion exchange resin becomes a key greater problem of current concern for the stable water quality promotion of condensate demineralizer. The presence of those impurities have practically been difficult to analyze by simple quality testing of the resin, and may result in as much as a hundred fold increase in chloride and sulfate in reactor water. As resin displacement with high leachable TOC, a repeated continuous soaking and effectively rinsing is required so that steady state TOC content less than 150 ppb should be achieved in an acceptably short period of time before put in-service. It is clear that cation resin containing high leachables generates high level of sulfates and sometimes also gives unexpected level of chlorides. The current TOC limits in condensate demineralizer effluent with 0.1 ppb become a significant experience to maintain reactor water soluble impurity in low levels. New resin should be subjected to TOC quality control testing prior to acceptance especially when first placed into service. TOC and organic chloride leachables for as-received virgin cation resin that are to be used in condensate polisher should be limited to be less than 100 mg-TOC and 0.5 mg-Cl per

Nuclear fuel performance in boiling water reactors

International Nuclear Information System (INIS)

Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

1981-01-01

A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

International Nuclear Information System (INIS)

Noh, Hyung Gyun; Kang, Hie Chan

2016-01-01

In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins

Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Noh, Hyung Gyun [Pohang University, Pohang (Korea, Republic of); Kang, Hie Chan [Kunsan University, Gunsan (Korea, Republic of)

2016-05-15

In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins.

The pressurized water reactor

International Nuclear Information System (INIS)

Gallagher, J.L.

1987-01-01

Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

International Nuclear Information System (INIS)

Schene, R.

2009-01-01

Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

Predicted effect of power uprating on the water chemistry of commercial boiling water reactors

International Nuclear Information System (INIS)

Yeh, Tsung-Kuang; Wang, Mei-Ya; Chu, Charles F.; Chang Ching

2009-01-01

The approach of power uprating has been adopted by operators of light water reactors in the past few decades in order to increase the power generation efficiency of nuclear reactors. The power uprate strategy is apparently applicable to the three nuclear reactors in Taiwan as well. When choosing among the three types of power uprating, measurement uncertainty, stretch power uprating, and extended power uprating, a deliberate and thorough evaluation is required before a final decision and an optimal selection can be made. One practical way of increasing the reactor power is to deliberately adjust the fuel loading pattern and the control rod pattern and thus to avoid replacing the primary coolant pump with a new one of larger capacity. The power density of the reactor will increase with increasing power, but the mass flow rate in the primary coolant circuit (PCC) of a light water reactor will slightly increase (usually by less than 5 %) or even remain unchanged. Accordingly, an uprated power would induce higher neutron and gamma photon dose rates in the reactor coolant but have a minor or no effect on the mass flow rate of the primary coolant. The radiolysis product concentrations and the electrochemical corrosion potential (ECP) values differ largely in the PCC of a boiling water reactor (BWR). It is very difficult to measure the water chemistry data directly at various locations of an actual reactor. Thus the impact of power uprating on the water chemistry of a BWR operating under hydrogen water chemistry (HWC) can only be theoretically evaluated through computer modelling. In this study, the DEMACE computer code was modified to investigate the impact of power uprating on the water chemistry under a fixed mass flow rate in the primary coolant circuit of a BWR/6 type plant. Simulations were carried out for hydrogen concentrations in feedwater ranging from 0.0 to 2.0 mg . kg -1 and for power levels ranging from 100 % to 120 %. The responses of water chemistry and ECP

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

Water cooled type nuclear power plant

International Nuclear Information System (INIS)

Arai, Shigeki.

1981-01-01

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

Overview of the US Department of Energy Light Water Reactor Sustainability Program

International Nuclear Information System (INIS)

McCarthy, K.A.; Williams, D.L.; Reister, R.

2012-01-01

The US Department of Energy Light Water Reactor Sustainability (LWRS) Program is focused on enabling the long-term operation of US commercial power plants. Decisions on life extension will be made by commercial power plant owners - the information provided by the research and development activities in the LWRS Program will reduce the uncertainty (and therefore the risk) associated with making those decisions. The LWRS Program encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper provides an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables. (author)

Safety analysis of an expert reactor protection system in nuclear power plants

International Nuclear Information System (INIS)

EL-Kafas, A.E.A.E.

1996-01-01

the purpose of the dissertation is to develop a real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification of plant transients (with and without scram). for this ERPS. probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents . the real- time information during transients and accidents can be obtained to asses the operator in his decision - making . Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. The system model consists of the dynamic differential equations for reactor core, pressurizer, steam generator, turbine and generator, piping and plenums. The system of equations can be solved by appropriate codes also displayed directly from sensors of the plant. All scenarios of transients, accidents and fault tress for plant systems are learned to ERPS

Safety analysis of an expert reactor protection system in nuclear power plants

International Nuclear Information System (INIS)

El-Kafas, A.A.

1997-01-01

The purpose of the dissertation is to develop real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification plant transients (with and without scram). For this erps, probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents. The real - time information during transients and accidents can be obtained to assess the operator in his decision - making. Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. 5-15 figs., 42 refs

Basic safety principles of KLT-40C reactor plants

International Nuclear Information System (INIS)

Beliaev, V.; Polunichev, V.

2000-01-01

The KLT-40 NSSS has been developed for a floating power block of a nuclear heat and power station on the basis of ice-breaker-type NSSS (Nuclear Steam Supply System) with application of shipbuilding technologies. Basic reactor plant components are pressurised water reactor, once-through coil-type steam generator, primary coolant pump, emergency protection rod drive mechanisms of compensate group-electromechanical type. Basic RP components are incorporated in a compact steam generating block which is arranged within metal-water shielding tank's caissons. Domestic regulatory documents on safety were used for the NSSS design. IAEA recommendations were also taken into account. Implementation of basic safety principles adopted presently for nuclear power allowed application of the KLT-40C plant for a floating power unit of a nuclear co-generation station. (author)

Dual pressurized light water reactor producing 2000 M We

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-10-15

The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

Comparative evaluation of recent water hammer events in light water reactors

International Nuclear Information System (INIS)

House, R.K.; Sursock, J.P.; Kim, J.H.

1987-01-01

Water hammer events that occurred in commercial U.S. light water reactors in the five-year period from 1981 to 1985 were surveyed, and a preliminary evaluation of the events was conducted. The information developed supplements a previous study which evaluated water hammer events in the twelve-year period from 1969 to 1981. The current study of water hammer events in the 1980's confirms that the rate of events remains relatively constant (less than 0.25 events per plant year) in both PWRs and BWRs. Although water hammer events are not normally considered a safety issue, the economic impact of the events on plant operations can be significant. One particular severe water hammer event is estimated to have cost the plant owner $10 million for repair and evaluation alone. A variety of key characteristics of the recent water hammer events are summarized to establish a basis for further study of preventative methods

Water Chemistry Control in Reducing Corrosion and Radiation Exposure at PWR Reactor

International Nuclear Information System (INIS)

Febrianto

2006-01-01

Water chemistry control plays an important role in relation to plant availability, reliability and occupational radiation exposures. Radiation exposures of nuclear plant workers are determined by the radiation rate dose and by the amount maintenance and repair work time Water chemistry has always been, from beginning of operation of power Pressurized Water Reactor, an important factor in determining the integrity of reactor components, fuel cladding integrity and minimize out of core radiation exposures. For primary system, the parameters to control the quality of water chemistry have been subject to change in time. Reactor water coolant pH need to be optimally controlled and be operated in range pH 6.9 to 7.4. At pH lower than 6.9, cause increasing the radiation exposure level and increasing coolant water pH higher than 7.4 will decrease radiation exposure level but increasing risk to fuel cladding and steam generator tube. Since beginning 90 decade, PWR water coolant pH tend to be operated at pH 7.4. This paper will discuss concerning water chemistry development in reducing corrosion and radiation exposure dose in PWR reactor. (author)

United States Department of Energy's reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

International Nuclear Information System (INIS)

1997-06-01

This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage

Sea water take-up facility for cooling reactor auxiliary

International Nuclear Information System (INIS)

Numata, Noriko; Mizutani, Akira; Hirako, Shizuka; Uchiyama, Yuichi; Oda, Atsushi.

1997-01-01

The present invention provides an improvement of a cooling sea water take-up facility for cooling auxiliary equipments of nuclear power plant. Namely, an existent sea water take-up facility for cooling reactor auxiliary equipments has at least two circulation water systems and three independent sea water systems for cooling reactor auxiliary equipments. In this case, a communication water channel is disposed, which connects the three independent sea water systems for cooling reactor auxiliary equipments mutually by an opening/closing operation of a flow channel partitioning device. With such a constitution, even when any combination of two systems among the three circulation water systems is in inspection at the same time, one system for cooling the reactor auxiliary equipments can be kept operated, and one system is kept in a stand-by state by the communication water channel upon periodical inspection of water take-up facility for cooling the auxiliary equipments. As a result, the sea water take-up facility for cooling auxiliary equipments of the present invention have operation efficiency higher than that of a conventional case while keeping the function and safety at the same level as in the conventional case. (I.S.)

Present status and recent improvements of water chemistry at Russian VVER plants

International Nuclear Information System (INIS)

Mamet, V.; Yurmanov, V.

2001-01-01

Water chemistry is an important contributor to reliable plant operation, safety barrier integrity, plant component lifetime, radiation safety, environmental impact. Primary and secondary water chemistry guidelines of Russian VVER plants have been modified to meet the new safety standards. At present 14 VVER units of different generation are in operation at 5 Russian NPPs. There are eight 4-loop pressurised water reactors VVER-1000 (1000 MWe) and six 6-loop pressurised water reactors VVER-440 (440 MWe). Generally, water chemistry at East European VVER plants (about 40 VVER-440 and VVER-1000 units in Ukraine, Bulgaria, Slovakia, Czech Republic, Hungary, Finland and Armenia) is similar to water chemistry at Russian VVER plants. Due to similar design and structural materials some water chemistry improvements were introduced at East European plants after they has been successfully implemented at Russian plants and vice versa. Some water chemistry improvements will be implemented at modern VVER plants under construction in Ukraine, Slovakia, Czech Republic, Iran, China, India. (R.P.)

Balancing passive and active systems for evolutionary water cooled reactors

International Nuclear Information System (INIS)

Fil, N.S.; Allen, P.J.; Kirmse, R.E.; Kurihara, M.; Oh, S.J.; Sinha, R.K.

1999-01-01

Advanced concepts of the water-cooled reactors are intended to improve safety, economics and public perception of nuclear power. The potential inclusion of new passive means in addition or instead of traditional active systems is being considered by nuclear plant designers to reach these goals. With respect to plant safety, application of the passive means is mainly intended to simplify the safety systems and to improve their reliability, to mitigate the effect of human errors and equipment malfunction. However, some clear drawbacks and the limited experience and testing of passive systems may raise additional questions that have to be addressed in the design process for each advanced reactor. Therefore the plant designer should find a reasonable balance of active and passive means to effectively use their advantages and compensate their drawbacks. Some considerations that have to be taken into account when balancing active/passive means in advanced water-cooled reactors are discussed in this paper. (author)

Development of a new model to evaluate the probability of automatic plant trips for pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Shimada, Yoshio [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan); Kawai, Katsunori; Suzuki, Hiroshi [Mitsubishi Heavy Industries Ltd., Tokyo (Japan)

2001-09-01

In order to improve the reliability of plant operations for pressurized water reactors, a new fault tree model was developed to evaluate the probability of automatic plant trips. This model consists of fault trees for sixteen systems. It has the following features: (1) human errors and transmission line incidents are modeled by the existing data, (2) the repair of failed components is considered to calculate the failure probability of components, (3) uncertainty analysis is performed by an exact method. From the present results, it is confirmed that the obtained upper and lower bound values of the automatic plant trip probability are within the existing data bound in Japan. Thereby this model can be applicable to the prediction of plant performance and reliability. (author)

Medium-sized water reactors for undeveloped regions

International Nuclear Information System (INIS)

Osmachkin, V. S.

2004-01-01

In the new century the growth of population and an increasing of energy demands together with the difficulties of fossil fuel supply are expected. It is important to find optimal ways in solving such problems without the climate warming. The nuclear power having many advantages in comparison with fossil fuel technologies could play the great role in near future. The Medium-Sized Nuclear Reactors for production of electricity, heat and fresh water are considered as a main direction of nuclear power applications in the developing world It is important to discuss the requirements to such nuclear plants for using in the Countries with Small and Medium Electricity Grids. Particularly, cost-benefit analysis of construction NPP has to include assessment of all type risks and effectiveness of plant. In the paper an attention is paid on Water Reactors designed on the basis of navy technology. Such compact PWR built on special mills and placed on special floating vessel could be used in undeveloped regions. Total plant can be transported to any point of World Ocean and return back to mill for repair or decommissioning after exhaustion of lifetime. It is expected that such reactors with innovative design approach, provision of high safety and proper economic efficiency, based on leasing procedures, could be very attractive for medium-sized and developing countries.(author)

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Hydrogen production by high temperature electrolysis of water vapour and nuclear reactors

International Nuclear Information System (INIS)

Jean-Pierre Py; Alain Capitaine

2006-01-01

This paper presents hydrogen production by a nuclear reactor (High Temperature Reactor, HTR or Pressurized Water Reactor, PWR) coupled to a High Temperature Electrolyser (HTE) plant. With respect to the coupling of a HTR with a HTE plant, EDF and AREVA NP had previously selected a combined cycle HTR scheme to convert the reactor heat into electricity. In that case, the steam required for the electrolyser plant is provided either directly from the steam turbine cycle or from a heat exchanger connected with such cycle. Hydrogen efficiency production is valued using high temperature electrolysis. Electrolysis production of hydrogen can be performed with significantly higher thermal efficiencies by operating in the steam phase than in the water phase. The electrolysis performance is assessed with solid oxide and solid proton electrolysis cells. The efficiency from the three operating conditions (endo-thermal, auto-thermal and thermo-neutral) of a high temperature electrolysis process is evaluated. The technical difficulties to use the gases enthalpy to heat the water are analyzed, taking into account efficiency and technological challenges. EDF and AREVA NP have performed an analysis to select an optimized process giving consideration to plant efficiency, plant operation, investment and production costs. The paper provides pathways and identifies R and D actions to reach hydrogen production costs competitive with those of other hydrogen production processes. (authors)

Outline of advanced boiling water reactor

International Nuclear Information System (INIS)

Yoshio Matsuo

1987-01-01

The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

Operation management of the prototype heavy water reactor 'Fugen'

Energy Technology Data Exchange (ETDEWEB)

Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan))

1983-09-01

The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported.

Light water reactor safety

CERN Document Server

Pershagen, B

2013-01-01

This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

Reactor water level control device

International Nuclear Information System (INIS)

Hiramatsu, Yohei.

1980-01-01

Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

Power-up of Fugen reactor and development of demonstration plant

International Nuclear Information System (INIS)

Sawai, Sadamu; Akebi, Michio; Yazaki, Akira.

1979-06-01

The Fugen Nuclear Power Station is the 165 MWe prototype plant characterized by heavy water-moderated, boiling light water-cooled, pressure tube type, and was developed by the Power Reactor and Nuclear Fuel Development Corporation, Japan. The plant went into commercial operation on March 20, 1979, in Tsuruga, Fukui Prefecture. Some delay in the overall schedule occurred due to the shortage of cement caused by the oil crisis, more stringent regulations as the result of stress corrosion cracking experienced in BWRs, and design modifications. All functional tests, the final check-up of the whole plant, and remaining modifying works had been completed by March 10, 1978. The minimum criticality was achieved with 22 mixed oxide fuel assemblies on March 20, 1978. Thereafter, the tests on reactor physics, plant dynamics, the performances of components and systems, and radiation and water chemistry have been carried out. 5 MWe was sent to grid system for the first time on July 29, 1978. The commercial operation of the plant was licenced by the Government on March 30, 1979. The conceptual design of the 600 MWe demonstration plant was finished in early 1979, and the detailed design is to be carried out in 1979 and 1980. The main design principle was incorporated in the conceptual design, but some modifications are to be made to reduce the power cost and to facilitate the easy maintenance. (Kako, I.)

Status of advanced technology and design for water cooled reactors: Light water reactors

International Nuclear Information System (INIS)

1988-10-01

Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

Feasibility of a single-purpose reactor plant for district heating in Finland

International Nuclear Information System (INIS)

Tarjanne, R.; Vuori, S.; Eerikaeinen, L.; Saukkoriipi, L.

A feasibility study of a single-purpose reactor for district heating is presented. The reactor chosen is of an ordinary pressurized water reactor type with a thermal output of 100 to 200 MW. Primary circuit steam generators employed in ordinary PWR's are replaced by water-water heat exchangers. For safety reasons an intermediate circuit separates the primary from the network water. The conditions of the district heating systems in Finland were taken into account, which led to the choice of an average temperature of 160 0 C for the reactor coolant and a pressure of 13.5 bar. This, coupled with minimal control requirements helped design a considerably simple reactor plant. On condition, the reactor satisfies the basic heat demand in a district heating system, the effective annual full-power operation time of the heating reactor is from 5000 h to 7000 h. Economic comparisons indicated that the heating reactor may be competitive if the operation period is of this order. As the reactor has to be sited near the heat consumption area for reasons of economy, the safety aspects are of major importance and may in themselves preclude the realization of the heating idea. (author)

Tritium concentration in the heavy water upgrading plants

International Nuclear Information System (INIS)

Croitoru, C.; Pop, F.; Titescu, Gh.; Dumitrescu, M.; Ciortea, C.; Stefanescu, I.; Peculea, M.; Pitigoi, Gh.; Trancota, D. . E-mail of corresponding author: croitoru@icsi.ro; Croitoru, C.)

2005-01-01

In the course of time heavy water used in CANDU nuclear power plants, as moderator or coolant, degrades, as a result of its impurification with light water and tritium. Concentration diminution below 99.8% mol for moderator and 99.75% mol for coolant causes an inefficient functioning of CANDU reactor. By isotopic distillation, light water is removed. Simultaneously tritium concentration takes place. The heavy water upgrading plant from Cernavoda is an isotopic separation cascade with two stages. The paper presents, for this plant, a theoretical study of the tritium concentration. (author)

Water hammer in nuclear power plants

International Nuclear Information System (INIS)

1979-07-01

In the past few years, licensees of operating reactors have reported a large number of water hammer events during commercial operation. Most of these events resulted in damage to piping supports and restraints. A few cases involved small cracks or ruptures in feedwater systems. As a result, in 1977 the NRC staff initiated a review of reported water hammer events and of the potential for occurrence of water hammer in all fluid systems that could have an impact on plant safety. The objectives of the review were to identify the causes of water hammer events that could affect reactor safety and to recommend further staff actions needed to reduce the likelihood of such events

Bacterial Colonization of Pellet Softening Reactors Used during Drinking Water Treatment

NARCIS (Netherlands)

Hammes, F.; Boon, N.; Vital, M.; Ross, P.; Magic-Knezev, A.; Dignum, M.

2010-01-01

Pellet softening reactors are used in centralized and decentralized drinking water treatment plants for the removal of calcium (hardness) through chemically induced precipitation of calcite. This is accomplished in fluidized pellet reactors, where a strong base is added to the influent to increase

Advanced light water reactor plant

International Nuclear Information System (INIS)

Giedraityte, Zivile

2008-01-01

For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

Reactor water sampling device

International Nuclear Information System (INIS)

Sakamaki, Kazuo.

1992-01-01

The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

Development of plant dynamic analysis code for integrated self-pressurized water reactor (ISPDYN), and comparative study of pressure control methods

International Nuclear Information System (INIS)

Kusunoki, Tsuyoshi; Yokomura, Takeyoshi; Nabeshima, Kunihiko; Shimazaki, Junya; Shinohara, Yoshikuni.

1988-01-01

This report describes the development of plant dynamic analysis code (ISPDYN) for integrated self-pressurized water reactor, and comparative study of pressure control methods with this code. ISPDYN is developed for integrated self-pressurized water reactor, one of the trial design by JAERI. In the transient responses, the calculated results by ISPDYN are in good agreement with the DRUCK calculations. In addition, this report presents some sensitivity studies for selected cases. Computing time of this code is very short so as about one fifth of real time. The comparative study of self-pressurized system with forced-pressurized system by this code, for rapid load decrease and increase cases, has provided useful informations. (author)

Power control device for heavy water moderated reactor

International Nuclear Information System (INIS)

Matsushima, Hidesuke; Masuda, Hiroyuki.

1978-01-01

Purpose: To improve self controllability of a nuclear power plant, as well as enable continuous power level control by a controlled flow of moderators in void pipes provided in a reactor core. Constitution: Hollow void pipes are provided in a reactor core to which a heavy water recycle loop for power control, a heavy water recycle pump for power control, a heavy water temperature regulator and a heavy water flow rate control valve for power control are connected in series to constitute a heavy water recycle loop for flowing heavy water moderators. The void ratio in each of the void pipes are calculated by a process computer to determine the flow rate and the temperature for the recycled heavy water. Based on the above calculation result, the heavy water temperature regulator is actuated by way of a temperature setter at the heavy water inlet and the heavy water flow rate is controlled by the actuation of the heavy water flow rate control valve. (Kawakami, Y.)

Improving economics and safety of water cooled reactors. Proven means and new approaches

International Nuclear Information System (INIS)

2002-05-01

Nuclear power plants (NPPs) with water cooled reactors [either light water reactors (LWRs) or heavy water reactors (HWRs)] constitute the large majority of the currently operating plants. Water cooled reactors can make a significant contribution to meeting future energy needs, to reducing greenhouse gas emissions, and to energy security if they can compete economically with fossil alternatives, while continuing to achieve a very high level of safety. It is generally agreed that the largest commercial barrier to the addition of new nuclear power capacity is the high capital cost of nuclear plants relative to other electricity generating alternatives. If nuclear plants are to form part of the future generating mix in competitive electricity markets, capital cost reduction through simplified designs must be an important focus. Reductions in operating, maintenance and fuel costs should also be pursued. The Department of Nuclear Energy of the IAEA is examining the competitiveness of nuclear power and the means for improving its economics. The objective of this TECDOC is to emphasize the need, and to identify approaches, for new nuclear plants with water cooled reactors to achieve competitiveness while maintaining high levels of safety. The cost reduction methods discussed herein can be implemented into plant designs that are currently under development as well as into designs that may be developed in the longer term. Many of the approaches discussed also generally apply to other reactor types (e.g. gas cooled and liquid metal cooled reactors). To achieve the largest possible cost reductions, proven means for reducing costs must be fully implemented, and new approaches described in this document should be developed and implemented. These new approaches include development of advanced technologies, increased use of risk-informed methods for evaluating the safety benefit of design features, and international consensus regarding commonly acceptable safety requirements that

Improving economics and safety of water cooled reactors. Proven means and new approaches

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-05-01

Nuclear power plants (NPPs) with water cooled reactors [either light water reactors (LWRs) or heavy water reactors (HWRs)] constitute the large majority of the currently operating plants. Water cooled reactors can make a significant contribution to meeting future energy needs, to reducing greenhouse gas emissions, and to energy security if they can compete economically with fossil alternatives, while continuing to achieve a very high level of safety. It is generally agreed that the largest commercial barrier to the addition of new nuclear power capacity is the high capital cost of nuclear plants relative to other electricity generating alternatives. If nuclear plants are to form part of the future generating mix in competitive electricity markets, capital cost reduction through simplified designs must be an important focus. Reductions in operating, maintenance and fuel costs should also be pursued. The Department of Nuclear Energy of the IAEA is examining the competitiveness of nuclear power and the means for improving its economics. The objective of this TECDOC is to emphasize the need, and to identify approaches, for new nuclear plants with water cooled reactors to achieve competitiveness while maintaining high levels of safety. The cost reduction methods discussed herein can be implemented into plant designs that are currently under development as well as into designs that may be developed in the longer term. Many of the approaches discussed also generally apply to other reactor types (e.g. gas cooled and liquid metal cooled reactors). To achieve the largest possible cost reductions, proven means for reducing costs must be fully implemented, and new approaches described in this document should be developed and implemented. These new approaches include development of advanced technologies, increased use of risk-informed methods for evaluating the safety benefit of design features, and international consensus regarding commonly acceptable safety requirements that

Analysis of Chemical Species Along the Process Stages of Demineralized Water Production at Reactor G.A. Siwabessy

International Nuclear Information System (INIS)

Nurul Huda; Setyono; Sumijanto; Diah E L; Ihsan, M.

2003-01-01

The tank water of multipurpose reactor G.A. Siwabessy is supplied from a water demineralization plant which works based on ion exchange processes. Controlling the quality of the water produced by this plant is one of many factor which effects the quality of the reactor tank water. This experiment resulted a characteristic pattern data of water and its chemical species content along process stages of demineralized water production at the reactor. The experiment results showed that the pH (degree of acidity), electric conductivity and dissolved cation (Ca 2+ , Mg 2+ ) lied at the permissible range. The value fluctuation of these variables showed a right pattern. It can be concluded that the water produced by this plant met the requirements to be used as primary cooling water of the reactor. However, the value of pH is still too low although it lied in the tolerance limit. Beside that, it isn't all of water impurities concentration can be predicted by the value of pH and conductivity. Therefore, the determination of water quality for the need of reactor tank water quiet to be done continually to keep the water condition in order to meet the quality required, and to evaluate and developed its production technology. (author)

An overview of reactor vessel internals segmentation for nuclear plant decommissioning

International Nuclear Information System (INIS)

Litka, T.J.

1994-01-01

Several nuclear plants have undergone reactor vessel (RV) internals segmentation as part of or in preparation for decommissioning the plant. In addition, several other nuclear facilities are planning for similar work efforts. The primary technology used for segmentation of RV internals, whether in-air or underwater is Plasma Arc Cutting (PAC). Metal Disintegration Machining (MDM) is also used for difficult to make cuts. PAC and MDM are deployed by various means including Long Handled Tools (LHTs), fixtures, tracks, and multi-axis manipulators. These enable remote cutting due to the radiation and/or underwater environment. A Boiling Water Reactor (BWR), a Pressurized Water Reactor (PWR), and a High Temperature Gas Reactor (HTGR) have had their internals removed and segmented using PAC and MDM. The cutting technology used for each component, location of cut, cut geometry and environment had to be determined well before the actual cutting operations. This allowed for the design, fabrication, and testing of the delivery systems. The technologies, selection process, and methodology for RV internals segmentation will be discussed in this paper

A dual pressurized water reactor producing 2000 MWe

International Nuclear Information System (INIS)

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

2010-01-01

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

Reactor water clean-up device

International Nuclear Information System (INIS)

Tanaka, Koji; Egashira, Yasuo; Shimada, Fumie; Igarashi, Noboru.

1983-01-01

Purpose: To save a low temperature reactor water clean-up system indispensable so far and significantly simplify the system by carrying out the reactor water clean-up solely in a high temperature reactor water clean-up system. Constitution: The reactor water clean-up device comprises a high temperature clean-up pump and a high temperature adsorption device for inorganic adsorbents. The high temperature adsorption device is filled with amphoteric ion adsorbing inorganic adsorbents, or amphoteric ion adsorbing inorganic adsorbents and anionic adsorbing inorganic adsorbents. The reactor water clean-up device introduces reactor water by the high temperature clean-up pump through a recycling system to the high temperature adsorption device for inorganic adsorbents. Since cations such as cobalt ions and anions such as chlorine ions in the reactor water are simultaneously removed in the device, a low temperature reactor water clean-up system which has been indispensable so far can be saved to realize the significant simplification for the entire system. (Seki, T.)

Two Pilot Plant Reactors Designed for the In Situ Bioremediation of Chlorobenzene-contaminated Ground Water: Hydrogeological and Chemical Characteristics and Bacterial Consortia

Energy Technology Data Exchange (ETDEWEB)

Vogt, Carsten, E-mail: vogt@umb.ufz.de; Alfreider, Albin [UFZ Centre for Environmental Research, Department of Environmental Microbiology (Germany); Lorbeer, Helmut [University of Technology Dresden, Institute of Waste Management and Contaminated Site Treatment (Germany); Ahlheim, Joerg; Feist, Bernd [UFZ Centre for Environmental Research, Department of Industrial and Mining Landscapes (Germany); Boehme, Olaf [GFE GmbH Halle (Germany); Weiss, Holger [UFZ Centre for Environmental Research, Department of Industrial and Mining Landscapes (Germany); Babel, Wolfgang; Wuensche, Lothar [UFZ Centre for Environmental Research, Department of Environmental Microbiology (Germany)

2002-05-15

The SAFIRA in situ pilot plant in Bitterfeld, Saxonia-Anhalt, Germany, currently serves as the test site for eight different in situ approaches to remediate anoxic chlorobenzene (CB)-contaminated ground water. Two reactors, both filled with original lignite-containing aquifer material, are designed for the microbiological in situ remediation of the ground water by the indigenous microbial consortia. In this study, the hydrogeological, chemical and microbiological conditions of the in flowing ground water and reactor filling material are presented,in order to establish the scientific basis for the start of the bioremediation process itself. The reactors were put into operation in June 1999. In the following, inflow CB concentrations in the ground water varied between 22 and 33 mg L{sup -1}; a chemical steady state for CB in both reactors was reached after 210 till 260 days operation time. The sediments were colonized by high numbers of aerobic, iron-reducing and denitrifying bacteria, as determined after 244 and 285 days of operation time. Furthermore, aerobic CB-degrading bacteria were detected in all reactor zones. Comparative sequence analysis of16S rDNA gene clone libraries suggest the dominance of Proteobacteria (Comamonadaceae, Alcaligenaceae, Gallionella group, Acidithiobacillus) and members of the class of low G+C gram-positive bacteria in the reactor sediments. In the inflowing ground water, sequences with phylogenetic affiliation to sulfate-reducing bacteria and sequences not affiliated with the known phyla of Bacteria, were found.

Two Pilot Plant Reactors Designed for the In Situ Bioremediation of Chlorobenzene-contaminated Ground Water: Hydrogeological and Chemical Characteristics and Bacterial Consortia

International Nuclear Information System (INIS)

Vogt, Carsten; Alfreider, Albin; Lorbeer, Helmut; Ahlheim, Joerg; Feist, Bernd; Boehme, Olaf; Weiss, Holger; Babel, Wolfgang; Wuensche, Lothar

2002-01-01

The SAFIRA in situ pilot plant in Bitterfeld, Saxonia-Anhalt, Germany, currently serves as the test site for eight different in situ approaches to remediate anoxic chlorobenzene (CB)-contaminated ground water. Two reactors, both filled with original lignite-containing aquifer material, are designed for the microbiological in situ remediation of the ground water by the indigenous microbial consortia. In this study, the hydrogeological, chemical and microbiological conditions of the in flowing ground water and reactor filling material are presented,in order to establish the scientific basis for the start of the bioremediation process itself. The reactors were put into operation in June 1999. In the following, inflow CB concentrations in the ground water varied between 22 and 33 mg L -1 ; a chemical steady state for CB in both reactors was reached after 210 till 260 days operation time. The sediments were colonized by high numbers of aerobic, iron-reducing and denitrifying bacteria, as determined after 244 and 285 days of operation time. Furthermore, aerobic CB-degrading bacteria were detected in all reactor zones. Comparative sequence analysis of16S rDNA gene clone libraries suggest the dominance of Proteobacteria (Comamonadaceae, Alcaligenaceae, Gallionella group, Acidithiobacillus) and members of the class of low G+C gram-positive bacteria in the reactor sediments. In the inflowing ground water, sequences with phylogenetic affiliation to sulfate-reducing bacteria and sequences not affiliated with the known phyla of Bacteria, were found

Elements of Design Consideration of Once-Through Cycle, Supercritical-Pressure Light Water Cooled Reactor

International Nuclear Information System (INIS)

Yoshiaki Oka; Sei-ichi Koshizuka; Yuki Ishiwatari; Akifumi Yamaji

2002-01-01

The paper describes elements of design consideration of supercritical-pressure, light water cooled reactors as well as the status and prospects of the research and development. It summarizes the results of the conceptual design study at the University of Tokyo from 1989. The research and development started in Japan, Europe and USA. The major advantages of the reactors are 1. Compact reactor and turbines due to high specific enthalpy of supercritical water 2.Simple plant system because of the once-through coolant cycle 3.Use of the experience of LWR and fossil-fired power plants. The temperatures of the major components such as reactor pressure vessel, coolant pipes, pumps and turbines are within the experience, in spite of the high outlet coolant temperature. 4.Similarity to LWR safety design and criteria, but no burnout phenomenon 5.Potential cost reduction due to smaller material expenditure and short construction period 6.The smallest reactor not in power rating, but in plant sizes. 7.High-thermal efficiency and low coolant flow rate because of high enthalpy rise. 8.Water cooled reactors potentially free from SCC (stress corrosion cracking) problems. 9.Compatibility of tight-fuel-lattice fast reactor core due to small coolant flow rate, potentially easy shift to fast breeder reactor without changing coolant technology. 10.Potential of producing energy products such as hydrogen and high quality hydro carbons. (authors)

Design of an integral missile shield in integrated head assembly for pressurized water reactor at commercial nuclear plants

International Nuclear Information System (INIS)

Baliga, Ravi; Watts, Tom Neal; Kamath, Harish

2015-01-01

In ICONE22, the authors presented the Integrated Head Assembly (IHA) design concept implemented at Callaway Nuclear Power Plant in Missouri, USA. The IHA concept is implemented to reduce the outage duration and the associated radiation exposure to the workers by reducing critical path time during Plant Refueling Outage. One of the head area components in the IHA is a steel missile shield designed to protect the Control Rod Drive Mechanism (CRDM) assembly from damaging other safety-related components in the vicinity in the Containment. Per Federally implemented General Design Criteria for commercial nuclear plants in the USA, the design of Nuclear Steam Supply System (NSSS) must provide protection from the damages caused by a postulated event of CRDM housing units and their associated parts disengaging from the reactor vessel assembly. This event is considered as a Loss of Coolant Accident (LOCA) and assumes that once the CRDM housing unit and their associated parts disengage from the reactor vessel internals assembly, they travel upward by the water jet with the following sequence of events: Per Reference 1, the drive shaft and control rod cluster are forced out of the reactor core by the differential pressure across the drive shaft with the assumption that the drive shaft and control rod cluster, latched together, are fully inserted when the accident occurs. After the travel, the rod cluster control spider will impact the lower side of the upper support plate inside the reactor vessel fracturing the flexure arms in the joint freeing the drive shaft from the control rod cluster. The control rod cluster is stopped by the upper support plate and will remain below the upper support plate during this accident. However, the drive shaft will continue to accelerate in the upward direction until it is stopped by a safety feature in the IHA. The integral missile shield as a safety feature in the IHA is designed to stop the CRDM drive shaft from moving further up in the

Radioactivity, radiation protection and monitoring during dismantling of light-water reactors

International Nuclear Information System (INIS)

Hummel, L.; Zech, J.B.

2005-01-01

Based on the radioactivity inventory in the systems and components of light-water reactors observed during operation, the impact of actions during plant emptying after the conclusion of power operation and possible subsequent long-term safe enclosure concerning the composition of the nuclide inventory of the plant to be dismantled will be described. Derived from this will be the effects on radioactivity monitoring in the plant, physical radiation protection monitoring, and the measured characterization of the residual materials resulting from the dismantling. The impact of long-term interim storage will also be addressed in the discussion. The talk should provide an overview of the interrelationships between source terms, decay times and the radioactivity monitoring requirements of the various dismantling concepts for commercial light-water reactors. (orig.)

Proposal for Dual Pressurized Light Water Reactor Unit Producing 2000 MWe

International Nuclear Information System (INIS)

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

2009-01-01

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

Instrumentation and control strategies for an integral pressurized water reactor

Directory of Open Access Journals (Sweden)

Belle R. Upadhyaya

2015-03-01

Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

Major outage trends in light water reactors. Interim report

International Nuclear Information System (INIS)

Burns, E.T.

1978-04-01

The report is a summary of the major outages which occurred in light water reactor plants during the period January 1971 through June 1977. Only those outages greater than 100 hours duration (exclusive of refueling outages) are included in the report. The trends in outages related to various reactor systems and components are presented as a function of plant age, and alternatively, calendar year. The principal contributors to major outages are ranked by their effect on the overall outage time for PWRs and BWRs. In addition, the outage history of each operating nuclear plant greater than 150 MWe is presented, along with a brief summary of those outages greater than two months duration

Water chemistry management of research reactor in JAERI

Energy Technology Data Exchange (ETDEWEB)

Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

Fuel handling system of nuclear reactor plants

International Nuclear Information System (INIS)

Faulstich, D.L.

1991-01-01

This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

Development of hot water supply system for a small district heating reactor

International Nuclear Information System (INIS)

Murase, Toshihiko; Narabayashi, Tadashi; Shimazu, Yoichiro

2007-01-01

On the earth, there are many environmental problems. For example, rapid increase of world population causes the enormous consumption of fossil fuel and emission of CO 2 into the global air. Now, mankaind faced to deal with these serious problems. One solution for these problems is utilization of nuclear reactors. Currently, about 65% of thermal output of a nuclear reactor is thrown away to the sea or the atmosphere through a turbine condenser. When a hot-water pipeline from a nuclear plant will be constructed, the exhaust heat from nuclear reactor will able to be utilized. Therefore, authors began to study nuclear power plant system for district heating. This reactor is based on a PWR plant. Its thermal output is 10 MWth and its electrical output is 3.4 MW. The nuclear plant supply electricity and heat for 2000 to 3000 houses. The plant aim to supply all the energy for the adjacent pepole's life, for example, heat, electricity and hydrogen for fuel battery car. This total-energy supply system assumed to be built in Northern area such as Hokkaido in Japan. In order to develop an optimum thermal design method for the system, heat transport experiments and thermal-hydraulic calculations were carried out. Using a metal pipe covered with foam-polyurethane thermal insulator, feed-water temperature and return-water temperature was measured to evaluate heat loss. As the result, the heat loss from the hot-water temperature was very little. The thermal-hydraulic calculation method was verified and applied to actual pipeline size calculation. The result of heat loss calculation will be 0.2degC/5 km. considering these results, the best pipe specification was obtained. (author)

U.S. experience with hydrogen water chemistry in boiling water reactors

International Nuclear Information System (INIS)

Cowan, R.L.; Head, R.A.; Indig, M.E.; Ruiz, C.P.; Simpson, J.L.

1988-01-01

Hydrogen water chemistry in boiling water reactors is currently being adopted by many utilities in the U.S., with eleven units having completed preimplementation test programs, four units operating permanently with hydrogen water chemistry, and six other units in the process of installing permanent equipment. Intergranular stress corrosion cracking protection is required for the recirculation piping system and other regions of the BWR systems. The present paper explores progress in predicting and monitoring hydrogen water chemistry response in these areas. Testing has shown that impurities can play an important role in hydrogen water chemistry. Evaluation of their effects are also performed. Both computer modeling and in plant measurements show that each plant will respond uniquely to feedwater hydrogen addition. Thus, each plant has its own unique hydrogen requirement for recirculation system protecion. Furthermore, the modeling, and plant measurements show that different regions of the BWR respond differently to hydrogen injection. Thus, to insure protection of components other than the recirculation systems may require more (or less) hydrogen demand than indicated by the recirculation system measurements. In addition, impurities such as copper can play a significant role in establishing hydrogen demand. (Nogami, K.)

Measuring device for water quality at reactor bottom

Energy Technology Data Exchange (ETDEWEB)

Urata, Hidehiro; Takagi, Jun-ichi

1995-10-27

The present invention concerns measurement for water quality at the bottom of a reactor of a BWR type plant, in which reactor water is sampled and analyzed in a state approximate to conditions in a pressure vessel. Based on the result, hydrogen injection amount is controlled during hydrogen injection operation. Namely, a monitor for water quality is disposed to a sampling line in communication with the bottom of a pressure vessel. A water quality monitor is disposed to a drain sampling line in communication with the bottom of the pressure vessel. A corrosion potentiometer is disposed to the pressure sampling line or the drain sampling line. A dissolved oxygen measuring device is disposed to the pressure vessel sampling line or the drain sampling line. With such a constitution, the reactor water can be sampled and analyzed in a state approximate to the conditions in the pressure vessel. In addition, signals from the water quality monitor are inputted to a hydrogen injection amount control device. As a result, the amount of hydrogen injected to primary coolants can be controlled in a state approximate to the conditions in the pressure vessel. (I.S.).

Measuring device for water quality at reactor bottom

International Nuclear Information System (INIS)

Urata, Hidehiro; Takagi, Jun-ichi.

1995-01-01

The present invention concerns measurement for water quality at the bottom of a reactor of a BWR type plant, in which reactor water is sampled and analyzed in a state approximate to conditions in a pressure vessel. Based on the result, hydrogen injection amount is controlled during hydrogen injection operation. Namely, a monitor for water quality is disposed to a sampling line in communication with the bottom of a pressure vessel. A water quality monitor is disposed to a drain sampling line in communication with the bottom of the pressure vessel. A corrosion potentiometer is disposed to the pressure sampling line or the drain sampling line. A dissolved oxygen measuring device is disposed to the pressure vessel sampling line or the drain sampling line. With such a constitution, the reactor water can be sampled and analyzed in a state approximate to the conditions in the pressure vessel. In addition, signals from the water quality monitor are inputted to a hydrogen injection amount control device. As a result, the amount of hydrogen injected to primary coolants can be controlled in a state approximate to the conditions in the pressure vessel. (I.S.)

General description of advanced heavy water reactor

International Nuclear Information System (INIS)

Kakodkar, A.; Sinha, R.K.; Dhawan, M.L.

1999-01-01

Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U 233 )O 2 and (Th-Pu)O 2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m 3 capacity directly into the core. Gravity driven water pool of capacity 6000 m 3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

Light-water-reactor hydrogen manual

International Nuclear Information System (INIS)

Camp, A.L.; Cummings, J.C.; Sherman, M.P.; Kupiec, C.F.; Healy, R.J.; Caplan, J.S.; Sandhop, J.R.; Saunders, J.H.

1983-06-01

A manual concerning the behavior of hydrogen in light water reactors has been prepared. Both normal operations and accident situations are addressed. Topics considered include hydrogen generation, transport and mixing, detection, and combustion, and mitigation. Basic physical and chemical phenomena are described, and plant-specific examples are provided where appropriate. A wide variety of readers, including operators, designers, and NRC staff, will find parts of this manual useful. Different sections are written at different levels, according to the most likely audience. The manual is not intended to provide specific plant procedures, but rather, to provide general guidance that may assist in the development of such procedures

The Consortium for Advanced Simulation of Light Water Reactors

International Nuclear Information System (INIS)

Szilard, Ronaldo; Zhang, Hongbin; Kothe, Douglas; Turinsky, Paul

2011-01-01

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

Water vapor as a perspective coolant for fast reactors

International Nuclear Information System (INIS)

Kalafati, D.D.; Petrov, S.I.

1978-01-01

Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

Advanced light-water reactors

International Nuclear Information System (INIS)

Golay, M.W.; Todreas, N.E.

1990-01-01

Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

Heavy water moderated gas-cooled reactors

International Nuclear Information System (INIS)

Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

1964-01-01

France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

Passive systems for light water reactors

International Nuclear Information System (INIS)

Adinolfi, R.; Noviello, L.

1990-01-01

The paper reviews the most original concepts that have been considered in Italy for the back-fitting of the nuclear power plants in order to reduce the probability and the importance of the release to the environment in case of a core melt. With reference either to BWR or PWR, passive concepts have been considered for back-fitting in the following areas: pump seals damage prevention and ECCS passive operation; reactor passive depressurization; molten reactor core passive cooling; metal containment passive water cooling through a water tank located at high level; containment isolation improvement through a sealing system; containment leaks control and limitation of environmental release. In addition some considerations will be made on the protection against external events introduced from the beginning on the PUN design either on building and equipment lay-out either on structure design. (author). 5 figs

Pumps for German pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Dernedde, R.

1984-01-01

The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

Chinese development of water-cooled reactors for non-electric applications

International Nuclear Information System (INIS)

Sun Yuliang; Duo Dong

1997-01-01

China is very densely inhabited land where approximately 75% of the primary energy consumption is contributed by coal. The strong dependence on coal results in two significant problems, the burden on transportation and the emission of environmental pollutants. Distances between coal production and consumption result in a burden on China's railway, road and water transport systems of approximately 40%, 25%, and 20% of their respective capacities. Environmentally, although the per capita annual CO2 emission is well under the world average, China ranks third after the USA and Russia in CO2 emission. Both of these problems can be alleviated through the increase use of nuclear energy. A dominant consumer of China's primary energy is in the form of heat application, of which district heating is a significant portion. The State is supporting the development of nuclear heating reactors for district heating purposes. The Institute of Nuclear Energy Technology (INET), with the support of the State, completed the construction of a 5MW test nuclear heating reactor in 1989. Since then, this reactor has been successfully operated for heating purposes, safety demonstration experiments and for tests on other applications. Subsequently, a 200MW commercial nuclear heating demonstration plant was approved by the State Council and design and licensing work on this plant is currently in progress at INET. This paper provides a review of the design parameters for these two nuclear heating plants. Other applications of the nuclear heating reactor, including seawater desalination, air conditioning and as an industrial process steam supply are currently under consideration. INET has considered two designs of a nuclear desalination plant (steam only and co-generation) coupled with the 200MW nuclear heating reactor. Also, INET is investigating use of this reactor for air conditioning and process chilled water production. The current status of these efforts are described in this paper. (author

Water chemistry in nuclear power stations with high-temperature reactors with particular reference to the AVR

International Nuclear Information System (INIS)

Nieder, R.; Resch, G.

1976-01-01

The water-steam cycle of a nuclear power plant with a helium-cooled high-temperature reactor differs in design data significantly and extensively from the corresponding cycles of light-water-cooled nuclear reactors and resembles to a great extent the water-steamcycle of a modern conventional power plant. The radioactive constituents of the water-steam cycle can be satisfactorily removed apart from Tritium by means of a pre-coat filter with powder-resisn, as comprehensive experiments have demonstrated. (orig.) [de

Revised reactor accident source terms in the U.S. and implementation for light water reactors

International Nuclear Information System (INIS)

Soffer, L.; Lee, J.Y.

1992-01-01

Current NRC reactor accident source terms used for licensing are contained in Regulatory Guides 1.3 and 1.4 and specify that 100 % of the core inventory of noble gases and 25 % of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental (I 2 ) iodine. These assumptions have strongly affected present nuclear plant designs. Severe accident research results have confirmed that although the current source term is very substantial and has resulted in a very high level of plant capability, the present source term is no longer compatible with a realistic understanding of severe accidents. The NRC has issued a proposed revision of the reactor accident source terms as part of several regulatory activities to incorporate severe accident insights for future plants. A revision to 10 CFR 100 is also being proposed to specify site criteria directly and to eliminate source terms and doses for site evaluation. Reactor source terms will continue to be important in evaluating plant designs. Although intended primarily for future plants, existing and evolutionary power plants may voluntarily apply revised accident source term insights as well in licensing. The proposed revised accident source terms are presented in terms of fission product composition, magnitude, timing and iodine chemical form. Some implications for light water reactors are discussed. (author)

Integrated gasification gas combined cycle plant with membrane reactors: Technological and economical analysis

International Nuclear Information System (INIS)

Amelio, Mario; Morrone, Pietropaolo; Gallucci, Fausto; Basile, Angelo

2007-01-01

In the present work, the capture and storage of carbon dioxide from the fossil fuel power plant have been considered. The main objective was to analyze the thermodynamic performances and the technological aspects of two integrated gasification gas combined cycle plants (IGCC), as well as to give a forecast of the investment costs for the plants and the resulting energy consumptions. The first plant considered is an IGCC* plant (integrated gasification gas combined cycle plant with traditional shift reactors) characterized by the traditional water gas shift reactors and a CO 2 physical adsorption system followed by the power section. The second one is an IGCC M plant (integrated gasification gas combined cycle plant with membrane reactor) where the coal thermal input is the same as the first one, but the traditional shift reactors and the physical adsorption unit are replaced by catalytic palladium membrane reactors (CMR). In the present work, a mono-dimensional computational model of the membrane reactor was proposed to simulate and evaluate the capability of the IGCC M plant to capture carbon dioxide. The energetic performances, efficiency and net power of the IGCC* and IGCC M plants were, thus, compared, assuming as standard a traditional IGCC plant without carbon dioxide capture. The economical aspects of the three plants were compared through an economical analysis. Since the IGCC* and IGCC M plants have additional costs related to the capture and disposal of the carbon dioxide, a Carbon Tax (adopted in some countries like Sweden) proportional to the number of kilograms of carbon dioxide released in the environment was assumed. According to the economical analysis, the IGCC M plant proved to be more convenient than the IGCC* one

Water regime of steam power plants

International Nuclear Information System (INIS)

Oesz, Janos

2011-01-01

The water regime of water-steam thermal power plants (secondary side of pressurized water reactors (PWR); fossil-fired thermal power plants - referred to as steam power plants) has changed in the past 30 years, due to a shift from water chemistry to water regime approach. The article summarizes measures (that have been realised by chemists of NPP Paks) on which the secondary side of NPP Paks has become a high purity water-steam power plant and by which the water chemistry stress corrosion risk of heat transfer tubes in the VVER-440 steam generators was minimized. The measures can also be applied to the water regime of fossil-fired thermal power plants with super- and subcritical steam pressure. Based on the reliability analogue of PWR steam generators, water regime can be defined as the harmony of construction, material(s) and water chemistry, which needs to be provided in not only the steam generators (boiler) but in each heat exchanger of steam power plant: - Construction determines the processes of flow, heat and mass transfer and their local inequalities; - Material(s) determines the minimal rate of general corrosion and the sensitivity for local corrosion damage; - Water chemistry influences the general corrosion of material(s) and the corrosion products transport, as well as the formation of local corrosion environment. (orig.)

Reactor water level control device

International Nuclear Information System (INIS)

Utagawa, Kazuyuki.

1993-01-01

A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

Studies on improvements in the control methods of boiling water reactor plant

International Nuclear Information System (INIS)

Mankin, Shuichi

1982-08-01

In order to improve the performance of regulation and load following control of boiling water reactor plant, optimal control theory is applied and new types of control method are developed. Case-α controller is first formulated on the basis of the optimal linear regulator theory applied to the linealized model of the system; it is then modified by adding a integration-type action in a feed back loop and by the use of variable gain and reference for adapting to the power level requested. Case-#betta# controller consists of a hierarchical control scheme which has classical P.I. type sub-loop controllers at the first level and a linear optimal regulator at the second level. The controller is designed on the basis of the optimal regulator theory applied to the multivariate autoregressive system model which is obtained from the identification experiments, where the system model is determined with the conventional sub-loop controllers included. The results of the simulation experiments show these control methods proposed have performed fairly well and will be useful for the improvement of the performance of nuclear power plant control. In addition, it is suggested that these control methods will be also attractive for the control of other production plants because these were developed in the attempt to solve the problems deviated from so called 'The gap between the optimal contro theory and actual systems.' (author)

Development of reactor water level sensor for extreme conditions

Energy Technology Data Exchange (ETDEWEB)

Miura, K; Ogasawara, T [Sukegawa Electric Co., Ltd., Hitachi, Ibaraki (Japan); Shibata, Akira; Nakamura, Jinichi; Saito, Takashi; Tsuchiya, Kunihiko [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

In the Fukushima accident, measurement failure of water level was one of the most important factors which caused serious situation. The differential pressure type water level indicators are widely used in various place of nuclear power plant but after the accident of TMI-2, the need of other reliable method has been required. The BICOTH type and the TRICOTH type water level indicator for light water power reactors had been developed for in-pile water level indicator but currently those are not adopted to nuclear power plant. In this study, the development of new type water level indicator composed of thermocouple and heater is described. Demonstration test and characteristic evaluation of the water level indicator were performed and we had obtained satisfactory results. (author)

An organic profile of a pressurised water reactor secondary plant

International Nuclear Information System (INIS)

Eeden, Nestor van; Stwayi, Mandisibuntu; Gericke, Gerhard

2012-01-01

Make-up water addition to the steam/water cycle at Koeberg Nuclear Power Station usually results in a corresponding increase of the chloride concentration in the steam generator blowdown system. During plant transients, when higher than normal make-up is required to the secondary plant, the concentration of chloride occasionally exceeds the limiting value for the station chemistry performance indicator. Irrespective of this, the demineralised water make-up supply tanks, which are routinely analysed for chloride, are within all recognised acceptable standards for secondary water make-up and therefore these tanks do not initially appear to be the source of chloride contamination. Water treatment at the plant relies essentially on ion exchange, which has been proven to be very effective in removing inorganic ionic species such as chloride. Organic compounds are less effectively removed by ion exchange and may pass through the treatment system, and these organics can reside undetected in the make-up water tanks. Historically, the elevated chloride concentration following high system make-up has been attributed to chlorinated organic compounds known as trihalomethanes being present in the make-up water tanks, but no rigorous study had been undertaken. As it has been assumed that the majority of chloride in the secondary system originates from the make-up water organic impurities, it was considered important to confirm this by compiling an organic profile of the secondary plant. The use of organic additives was also taken into account in the profile. This work has confirmed the contribution from trihalomethanes and has also found that other organochlorides contribute even more significantly to the overall chloride inventory of the secondary plant. (orig.)

The design features of integrated modular water reactor (IMR)

International Nuclear Information System (INIS)

Kanagawa, T.; Goto, M.; Usui, S.; Suzuta, T.; Serizawa, A.; Kunugi, T.; Yamauchi, T.; Itoh, G.; Matsumura, T.

2004-01-01

Small-to-medium-sized (300-600 MWe) reactors are required for the electric power market in the near future (2010-2030). The main theme in the development of small-to-medium-sized reactor is how to realize competitive cost against other energy sources. As measures to this disadvantage, greatly simplified and small-scale design is needed. From such point of view, Integrated Modular Water Reactor (IMR), whose electric output power is 350 MWe, adopts integrated and high temperature two-phase natural circulation system for the primary system. In this design, main coolant pipes, a pressurizer, and reactor coolant pumps are not needed, and the sizes of the reactor vessel and steam generators are minimized. Additionally, to enhance the economy of the whole plant, fluid systems, and Instrumentation and Control systems of IMR have also been reviewed to make them simplest and smallest taking the advantage of the IMR concept and the state of the art technologies. For example, the integrated primary system and the stand-alone direct heat removal system make the safety system very simple, i.e., no injection, no containment spray, no emergency AC power, etc. The chemical and volume control system is also simplified by eliminating the boron control system and the seal water system of reactor coolant pumps. In this paper, the status of the IMR development and the outline of the IMR design efforts to achieve the simplest and smallest plant are presented. (authors)

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)

Manning designs for nuclear district-heating plant (NDHP) with RUTA-type reactor

International Nuclear Information System (INIS)

Gerasimova, V.S.; Mikhan, V.I.; Romenkov, A.A.

2001-01-01

RUTA-type reactor is a water cooled water-moderated pool-type reactor with an atmospheric pressure air medium. The reactor has been designed for heating and hot water supply. Nuclear district heating plant (NDHP) with RUTA-type reactor facility has been designed with a three circuit layout. Primary circuit components are arranged integrally in the reactor vessel. Natural coolant circulation mode is used in the primary circuit. A peculiarity of RUTA-based NDHP as engineered system is a smooth nature of its running slow variation of the parameters at transients. Necessary automation with application of computer equipment will be provided for control and monitoring of heat production process at NDHP. Under developing RUTA-based NDHP it is foreseen that operating staff performs control and monitoring of heat generation process and heat output to consumers as well as current maintenance of NDHP components. All other works associated with NDHP operation should be fulfilled by extraneous personnel. In so doing the participation of operating staff is also possible. (author)

Advances in commercial heavy water reactor power stations

International Nuclear Information System (INIS)

Brooks, G.L.

1987-01-01

Generating stations employing heavy water reactors have now firmly established an enviable record for reliable, economic electricity generation. Their designers recognize, however, that further improvements are both possible and necessary to ensure that this reactor type remains attractively competitive with alternative nuclear power systems and with fossil-fuelled generation plants. This paper outlines planned development thrusts in a number of important areas, viz., capital cost reduction, advanced fuel cycles, safety, capacity factor, life extension, load following, operator aida, and personnel radiation exposure. (author)

A stability identification system for boiling water nuclear reactors

International Nuclear Information System (INIS)

Belblidia, L.A.; Chevrier, A.

1994-01-01

Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

The Swr 1000: a nuclear power plant concept with boiling water reactor for maximum safety and economy of operation

International Nuclear Information System (INIS)

Brettschuh, W.

2001-01-01

The SWR 1000 is a design concept for a light water reactor nuclear power plant that meets all requirements regarding plant safety, economic efficiency and environ-mental friendliness. As a result of the plant's safety concept, the occurrence of core damage can, for all practical intents and purposes, be ruled out. If a core melt accident should nevertheless occur, the molten core can be retained inside the RPV, thus ensuring that all consequences of such an accident remain restricted to the plant itself. The power generating costs of the SWR 1000 are lower than with those of coal-fired and combined-cycle power plants. Power generation using nuclear energy does not release carbon dioxide to the environment, thus meeting the need for sustainable protection of our global climate. (author)

Contemporary pressurized water reactor technology in the world

International Nuclear Information System (INIS)

Komarek, A.

1991-01-01

The recent political events enabled Czechoslovak industrial companies to come into direct contact with leading western companies involved in pressurized water ractor technology. A survey is presented of the present situation at the world market of PWR type nuclear power plant suppliers and suppliers of fuel cycle services. Information is given on the potential bids for the next Czechoslovak nuclear power plants with PWR reactors. Economic aspects of the potential bids are presented including some considerations about the participation of the Czechoslovak nuclear industry as a supplier of the reactor for the future power plants. Main technical parameters are listed of PWR units with an output about 1000 MW supplied by Westinghouse EC, ABB -Combustion Engineering and Siemens AG. At present, the bids for new Czechoslovak nuclear power plants are being evaluated. No information on terms of the bids actually coming from foreign companies is used in the article. (Z.S.). 9 figs., 5 tabs

Dual-purpose light water reactor supplying heat for desalination

International Nuclear Information System (INIS)

Waplington, G.; Fichtner, H.

1978-01-01

The technical as well as the economic aspects of using a large commercial light water reactor for the production of both electricity and potable water have been examined. For the basis of the study, the multistage flash distillation process was selected, in conjunction with a reactor rated at not less than 2100 MW (thermal). Combined use of a condensing and a back-pressure turbine (the latter matched to distillation plant steam requirements) represents a convenient method for supplying process heat. Overall costs can be fairly allocated to the two products using the ''power credit'' method. A sample economic evaluation indicates highly favorable water costs as compared with more conventional distillation schemes based on fossil fuel

NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

International Nuclear Information System (INIS)

1994-08-01

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Super critical water reactors

International Nuclear Information System (INIS)

Dumaz, P.; Antoni, O; Arnoux, P.; Bergeron, A; Renault, C.; Rimpault, G.

2005-01-01

Water is used as a calori-porter and moderator in the most major nuclear centers which are actually in function. In the pressurized water reactor (PWR) and boiling water reactor (BWR), water is maintained under critical point of water (21 bar, 374 Centigrade) which limits the efficiency of thermodynamic cycle of energy conversion (yield gain of about 33%) Crossing the critical point, one can then use s upercritical water , the obtained pressure and temperature allow a significant yield gains. In addition, the supercritical water offers important properties. Particularly there is no more possible coexistence between vapor and liquid. Therefore, we don't have more boiling problem, one of the phenomena which limits the specific power of PWR and BWR. Since 1950s, the reactor of supercritical water was the subject of studies more or less detailed but neglected. From the early 1990s, this type of conception benefits of some additional interests. Therefore, in the international term G eneration IV , the supercritical water reactors had been considered as one of the big options for study as Generation IV reactors. In the CEA, an active city has engaged from 1930 with the participation to a European program: The HPWR (High Performance Light Water Reactor). In this contest, the R and D studies are focused on the fields of neutrons, thermodynamic and materials. The CEA intends to pursue a limited effort of R and D in this field, in the framework of international cooperation, preferring the study of versions of rapid spectrum. (author)

Water chemistry: protecting the industry's investment. Making or breaking plant operations

International Nuclear Information System (INIS)

Anon.

1983-01-01

Good water chemistry is a good way to preserve the life of steam generators and other plant components. Pipe cracks in boiling-water reactors, tube pitting, denting and cracking in pressurized-water reactors are all problems that are surfacing due to poor water chemistry, i.e., the lack of water purity. Water is essential to power generation and is corrosive under the best of conditions. But to a metal system filled with water and subject to high pressure, high temperature, and impurities such as chlorides, the potential for rapid and permanent damage rises to serious proportions. In addition, radiation levels increase from corrosive products circulated through the reactor vessel

Two-loop feed water control system in BWR plants

International Nuclear Information System (INIS)

Omori, Takashi; Watanabe, Takao; Hirose, Masao.

1982-01-01

In the process of the start-up and shutdown of BWR plants, the operation of changing over feed pumps corresponding to plant output is performed. Therefore, it is necessary to develop the automatic changeover system for feed pumps, which minimizes the variation of water level in reactors and is easy to operate. The three-element control system with the water level in reactors, the flow rate of main steam and the flow rate of feed water as the input is mainly applied, but long time is required for the changeover of feed pumps. The two-loop feed control system can control simultaneously two pumps being changed over, therefore it is suitable to the automatic changeover control system for feed pumps. Also it is excellent for the control of the recirculating valves of feed pumps. The control characteristics of the two-loop feed water control system against the external disturbance which causes the variation of water level in reactors were examined. The results of analysis by simulation are reported. The features of the two-loop feed water control system, the method of simulation and the evaluation of the two-loop feed water control system are described. Its connection with a digital feed water recirculation control system is expected. (Kako, I.)

Control of water chemistry in operating reactors

International Nuclear Information System (INIS)

Riess, R.

1997-01-01

Water chemistry plays a major role in fuel cladding corrosion and hydriding. Although a full understanding of all mechanisms involved in cladding corrosion does not exist, controlling the water chemistry has achieved quite some progress in recent years. As an example, in PWRs the activity transport is controlled by operating the coolant under higher pH-values (i.e. the ''modified'' B/Li-Chemistry). On the other hand, the lithium concentration is limited to a maximum value of 2 ppm in order to avoid an acceleration of the fuel cladding corrosion. In BWR plants, for example, the industry has learned on how to limit the copper concentration in the feedwater in order to limit CILC (Copper Induced Localized Corrosion) on the fuel cladding. However, economic pressures are leading to more rigorous operating conditions in power reactors. Fuel burnups are to be increased, higher efficiencies are to be achieved, by running at higher temperatures, plant lifetimes are to be extended. In summary, this paper will describe the state of the art in controlling water chemistry in operating reactors and it will give an outlook on potential problems that will arise when going to more severe operating conditions. (author). 3 figs, 6 tabs

Control of water chemistry in operating reactors

Energy Technology Data Exchange (ETDEWEB)

Riess, R [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

1997-02-01

Water chemistry plays a major role in fuel cladding corrosion and hydriding. Although a full understanding of all mechanisms involved in cladding corrosion does not exist, controlling the water chemistry has achieved quite some progress in recent years. As an example, in PWRs the activity transport is controlled by operating the coolant under higher pH-values (i.e. the ``modified`` B/Li-Chemistry). On the other hand, the lithium concentration is limited to a maximum value of 2 ppm in order to avoid an acceleration of the fuel cladding corrosion. In BWR plants, for example, the industry has learned on how to limit the copper concentration in the feedwater in order to limit CILC (Copper Induced Localized Corrosion) on the fuel cladding. However, economic pressures are leading to more rigorous operating conditions in power reactors. Fuel burnups are to be increased, higher efficiencies are to be achieved, by running at higher temperatures, plant lifetimes are to be extended. In summary, this paper will describe the state of the art in controlling water chemistry in operating reactors and it will give an outlook on potential problems that will arise when going to more severe operating conditions. (author). 3 figs, 6 tabs.

Thermal and stability considerations for a supercritical water-cooled fast reactor during power-raising phase of plant startup

International Nuclear Information System (INIS)

Cai, Jiejin; Ishiwatari, Yuki; Oka, Yoshiaki; Ikejiri, Satoshi

2009-01-01

This paper describes thermal analyses and linear stability analyses of the Supercritical Water-cooled Fast Reactor with 'two-path' flow scheme during the power-raising phase of plant startup. For thermal consideration, the same criterion of the maximum cladding surface temperature (MCST) as applied to the normal operating condition is used. For thermal-hydraulic stability consideration, the decay ratio of 0.5 is applied, which is taken from BWRs. Firstly, we calculated the flow rate distribution among the parallel flow paths from the reactor vessel inlet nozzles to the mixing plenum below the core using a system analysis code. The parallel flow paths consist of the seed fuel assemblies cooled by downward flow, the blanket fuel assemblies cooled by downward flow and the downcomer. Then, the MCSTs are estimated for various reactor powers and feedwater flow rates with system analyses. The decay ratios are estimated with linear stability analyses. The available range of the reactor power and feedwater flow rate to satisfy the thermal and stability criteria is obtained. (author)

A small floating seawater desalination plant using a nuclear heating reactor coupled with the MED process

International Nuclear Information System (INIS)

Dong Duo; Wu Shaorong; Zhang Dafang; Wu Zongxin

1997-01-01

A small floating seawater desalination plant using a nuclear heating reactor coupled with a multi-effect distillation (MED) process was designed by the Institute of Nuclear Energy Technology, Tsinghua University of China. It was intended to supply potable water to remove coastal areas or islands where both fresh water and energy are severely lacking, and also to serve as a demonstration and training facility. The design of a small floating plant coupled two proven technologies in the cogeneration mode: a nuclear heating reactor (NHR-10), with inherent, passive safety features based on NHR-5 experience, and a low temperature MED process. The secondary loop was designed as a safety barrier between the primary loop and the steam loop. With a 10 MW(th) heating reactor, the floating plant could provide 4,000 m 3 /d of potable water and 750 kW of electricity. The design concept and parameters, safety features, coupling scheme and floating plant layout are presented in the paper. (author). 3 refs, 4 figs, 3 tabs

Steam plant for pressurized water reactors

International Nuclear Information System (INIS)

Anon.

1983-01-01

This book discusses the research and development organisations and users to highlight those aspects of the steam turbine and associated plant which are particularly related to the PWR system. The contents include: Characteristics of the steam system (including feed train, dump system and safety aspects); overall design aspects of high and half speed turbines; design aspects of the steam generator and seismic considerations; moisture separators and reheaters; feed pumps and their drives; water treatment; safety related valves; operational experience; availability and performance

Implications for accident management of adding water to a degrading reactor core

International Nuclear Information System (INIS)

Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J.

1994-02-01

This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents

Implications for accident management of adding water to a degrading reactor core

Energy Technology Data Exchange (ETDEWEB)

Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1994-02-01

This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents.

Standard Guide for Benchmark Testing of Light Water Reactor Calculations

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

Evaluation of responses to IE Bulletin 82-02: degradation of threaded fasteners in reactor coolant pressure boundary of pressurized-water-reactor plants

International Nuclear Information System (INIS)

Anderson, W.; Sterner, P.

1985-05-01

IE Bulletin 82-02 was issued by the NRC on June 2, 1982, to notify licensees about incidents of severe degradation of threaded fasteners. The bulletin required appropriate action including submittal of information from pressurized water reactors having an operating license. Responses from 41 licensees included their recent experience with degradation of threaded fasteners in primary system components. Data from recent regular inspections of reactor coolant pressure boundary component connections of 6-in. size and larger are compiled for technical evaluation. Statistical analysis is used to determine significant factors related to frequency of leakage incidents in connections, occurrence of degradation of bolts and studs, and the need for bolt replacement. Factors examined include the age of the plant, types of components, use of lubricants and sealants, and differences between plants. The compiled data indicate that, on the average, 10% of the bolted connections show evidence of leaking during an 18-month period. Also, 80% of the connections that show evidence of leakage undergo some degradation of the bolting. Results of the analysis show a significant decrease in the occurrence of bolting degradation events as the age of the plant increases. The data also show that valves are less subject to bolting corrosion. A group of 5 of the 41 plants accounted for about one-half of the reported leakage and corrosion events. The common characteristic found for four of these five plants was the lubricant used. The use of nickel-graphite based lubricants appears to offer a significantly reduced incidence of leakage and corrosion, based on late corrections to the reported data. The data also permit the conclusion that the use of molybdenum-disulfide-based lubricants and graphite-based lubricants results in a significantly increased incidence of leakage and corrosion. Reporting of data on lubricants was of poor quality and detracted from the value of the bulletin responses

Nuclear power desalinating complex with IRIS reactor plant and Russian distillation desalinating unit

International Nuclear Information System (INIS)

Kostin, V. I.; Panov, Yu.K.; Polunichev, V. I.; Fateev, S. A.; Gureeva, L. V.

2004-01-01

This paper has been prepared as a result of Russian activities on the development of nuclear power desalinating complex (NPDC) with the IRIS reactor plant (RP). The purpose of the activities was to develop the conceptual design of power desalinating complex (PDC) and to evaluate technical and economical indices, commercial attractiveness and economical efficiency of PDC based on an IRIS RP with distillation desalinating plants. The paper presents the main results of studies as applied to dual-purpose PDC based on IRIS RP with different types of desalinating plants, namely: characteristics of nuclear power desalinating complex based on IRIS reactor plant using Russian distillation desalinating technologies; prospective options of interface circuits of the IRIS RP with desalinating plants; evaluations of NPDC with IRIS RP output based on selected desalinating technologies for water and electric power supplied to the grid; cost of water generated by NPDC for selected interface circuits made by the IAEA DEEP code as well as by the Russian TEO-INVEST code; cost evaluation results for desalinated water of PDC operating on fossil fuel and conditions for competitiveness of the nuclear PDC based on IRIS RP compared with analog desalinating complexes operating on fossil fuel.(author)

Database structure and file layout of Nuclear Power Plant Database. Database for design information on Light Water Reactors in Japan

International Nuclear Information System (INIS)

Yamamoto, Nobuo; Izumi, Fumio.

1995-12-01

The Nuclear Power Plant Database (PPD) has been developed at the Japan Atomic Energy Research Institute (JAERI) to provide plant design information on domestic Light Water Reactors (LWRs) to be used for nuclear safety research and so forth. This database can run on the main frame computer in the JAERI Tokai Establishment. The PPD contains the information on the plant design concepts, the numbers, capacities, materials, structures and types of equipment and components, etc, based on the safety analysis reports of the domestic LWRs. This report describes the details of the PPD focusing on the database structure and layout of data files so that the users can utilize it efficiently. (author)

Seawater desalination using small and medium light water reactors

International Nuclear Information System (INIS)

Shimamura, Kazuo

2000-01-01

Water is an essential substance for sustaining human life. As Japan is an island country, surrounded by the sea and having abundant rainfall, there is no scarcity of water in daily life except during abnormally dry summers or after disasters such as earthquakes. Consequently, there is hardly any demand for seawater desalination plants except on remote islands, Okinawa and a part of Kyushu. However, the IAEA has forecast a scarcity of drinking water in developing countries at the beginning of the 21st century. Further, much more irrigation water will be required every year to prevent cultivated areas from being lost by desertification. If developing countries were to produce such water by seawater desalination using current fossil fuel energy technology, it would cause increased air pollution and global warming. This paper explains the concept of seawater desalination plants using small and medium water reactors (hereinafter called 'nuclear desalination'), as well as important matters regarding the export nuclear desalination plants to developing countries. (author)

Light-water reactor research and development

International Nuclear Information System (INIS)

1985-05-01

This report on the national program of research and development on light water reactors is the second of two reports requested in 1982 by W. Kenneth Davis, Deputy Secretary of the Department of Energy. A first report, published in September 1983, treated the needs for safety-related R and D. In this second report, the Energy Research Advisory Board finds that, although many light water reactors are providing reliable and economic electricity, it appears unlikely that U.S. utilities will order additional reactors until the currently unacceptable economic risk, created by the regulatory climate and uncertain demand, is reduced. Thus it is unlikely that the private sector alone will fund major LWR design improvements. However, nuclear power will continue on its current course of expansion overseas. DOE participation is vitally needed to support the national interest in LWR technology. The report outlines R and D needs for a program to improve the safety, reliability, and economics of the present generation of plants; to develop evolutionary improved designs to be ready when needed; and to explore innovative longer-term concepts for deployment after the year 2000. The respective roles of government and the private sector are discussed

The status of improved pressurized heavy water reactor development - A new option for PHWR -

International Nuclear Information System (INIS)

Park, Tae Keun; Yeo, Ji Won

1996-03-01

Currently, the 900 MWe class Improved Pressurized Heavy Water Reactor (PHWR), which is a type of CANDU reactor based on the systems and components of operating CANDU plants, is under development. The improved PHWR has a 480 fuel channel calandria, uses 37 element natural uranium fuel bundles and has a single unit containment. Adaptation of a steel-lined containment structure and improved containment isolation systems permit a reduced exclusion area boundary (EAB) compared to the existing larger capacity CANDU reactors (Darlington, Bruce B). The improved PHWR buildings are arranged to achieve minimum spacing between reactor units. Plant safety and economy are increased through various design changes based on the operating experience of existing CANDU plants. 4 refs. (Author)

Stability monitoring of a natural-circulation-cooled boiling water reactor

International Nuclear Information System (INIS)

Hagen, T.H.J.J. van der.

1989-01-01

Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

Aging management strategy for reactor internals of Korean nuclear power plants

International Nuclear Information System (INIS)

Hwang, Seong Sik; Kim, Soung Woo; Lee, Sam Lai; Hong, Seung Mo; Kim, Hong Pyo; Kim, Dong Jin; Lim, Yun Soo; Kim, Joung Soo; Jung, Man Kyo; Park, Jang Yul

2010-01-01

This report describes various factors on the IASCC of reactor internals in terms of fluence, stress, water chemistries and materials. Materials of each components of Korean nuclear power plants have been surveyed. A technical report for a management of reactor internals issued by EPRI was reviewed for a selection of most susceptible area among many components. Baffle former bolts are considered to be the most susceptible parts due to high irradiation level(fluence) and high tensile stress. Neutron fluence of Kori-1 and Kori-2 was calculated based on fuel exchange history, fuel performance and plant operation history. This report will be used for more advanced inspection and maintenance guidelines development by supplying inspection intervals and components (most susceptible regions) for the long term operation plants

Qualification issues for advanced light-water reactor protection systems

International Nuclear Information System (INIS)

Korsah, K.; Clark, R.L.; Antonescu, C.

1993-01-01

The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and fiber optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying the I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light-water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light-water reactor. The template was then used to suggest a methodology for the qualification of microprocessor-based protection systems. The methodology identifies standards/regulatory guides (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems. This approach addresses in part issues raised in NRC policy document SECY-91-292, which recognizes that advanced I ampersand C systems for the nuclear industry are ''being developed without consensus standards. as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.''

77 FR 15812 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

Science.gov (United States)

2012-03-16

... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... Feedwater Systems for Light- Water Reactors.'' DG-1265 is proposed revision 2 of Regulatory Guide 1.68.1... Plants,'' dated January 1977. This regulatory guide is being revised to: (1) expand the scope of the...

Water feeding/condensating device and operation method in nuclear power plant

International Nuclear Information System (INIS)

Shibayama, Takashi.

1989-01-01

The present invention overcomes a problem in reactor water level control occurring upon operation of a water feeding/condensating system in a nuclear power plant. That is, the water feed system to a nuclear reactor is constituted with parallel circuit comprising a reactor feedwater pump driven by a steam turbine and a serial circuit composed of a reactor feedwater pump driven by an electrical motor and a pump adjusting valve for controlling the amount of feedwater at the exit of the motor driven feedwater pump. Further, a reactor feedwater control valve having a function of controlling the feedwater to the reactor is disposed to the bypass pipeway for bypassing the parallel circuit of feedwater pumps. In this constitution, water can be fed to the nuclear reactor by way of the reactor feedwater pump bypass control valve upon starting and stopping of a nuclear feedwater pump driven by electric motor upon starting and shutdown of the nuclear reactor. Accordingly, stable water level control can be conducted for the reactor core with no effect of rapid pressure fluctuation due to the starting and the stopping of the reactor feedwater pump driven by electric motor. (I.S.)

Cogeneration of Electricity and Potable Water Using The International Reactor Innovative And Secure (IRIS) Design

International Nuclear Information System (INIS)

Ingersoll, D.T.; Binder, J.L.; Kostin, V.I.; Panov, Y.K.; Polunichev, V.; Ricotti, M.E.; Conti, D.; Alonso, G.

2004-01-01

The worldwide demand for potable water has been steadily growing and is projected to accelerate, driven by a continued population growth and industrialization of emerging countries. This growth is reflected in a recent market survey by the World Resources Institute, which shows a doubling in the installed capacity of seawater desalination plants every ten years. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh/m3 of produced desalted water. At current U.S. water use rates, a dedicated 1000 MW power plant for every one million people would be required to meet our water needs with desalted water. Nuclear energy plants are attractive for large scale desalination application. The thermal energy produced in a nuclear plant can provide both electricity and desalted water without the production of greenhouse gases. A particularly attractive option for nuclear desalination is to couple a desalination plant with an advanced, modular, passively safe reactor design. The use of small-to-medium sized nuclear power plants allows for countries with smaller electrical grid needs and infrastructure to add new electrical and water capacity in more appropriate increments and allows countries to consider siting plants at a broader number of distributed locations. To meet these needs, a modified version of the International Reactor Innovative and Secure (IRIS) nuclear power plant design has been developed for the cogeneration of electricity and desalted water. The modular, passively safe features of IRIS make it especially well adapted for this application. Furthermore, several design features of the IRIS reactor will ensure a safe and reliable source of energy and water even for countries with limited nuclear power experience and infrastructure. The IRIS-D design utilizes low-quality steam extracted from the low-pressure turbine to boil seawater in a multi-effect distillation desalination plant. The desalination plant is based on the horizontal

Startup and commissioning of pressurized water reactors

International Nuclear Information System (INIS)

Albert, L.J.; Gilbert, C.F.

1983-05-01

A critical phase of plant development is the test, startup, and commissioning period. The effort expended prior to commissioning has a definite effect on the reliability and continuing availability of the plant during its life. This paper describes a test, startup, and commissioning program for a pressurized water reactor (PWR) plant. This program commences with the completion of construction and continues through the turnover of equipment/systems to the owner's startup/ commissioning group. The paper addresses the organization of the test/startup group, planning and scheduling, test procedures and initial testing, staffing and certification of the test group, training of operators, and turnover to the owner

Review of irradiation experiments for water reactor safety research

International Nuclear Information System (INIS)

Tobioka, Toshiaki

1977-02-01

A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

High performance light water reactor

International Nuclear Information System (INIS)

Squarer, D.; Schulenberg, T.; Struwe, D.; Oka, Y.; Bittermann, D.; Aksan, N.; Maraczy, C.; Kyrki-Rajamaeki, R.; Souyri, A.; Dumaz, P.

2003-01-01

The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:-A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.-Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a 'reference design', developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the 'reference design' was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to 'calibrate' the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly. Preliminary selection was made for the HPLWR scale

Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

International Nuclear Information System (INIS)

Park, T-K.; Jung, S-H.

1996-01-01

During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D 2 O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

Nuclear power plant control room task analysis. Pilot study for pressurized water reactors

International Nuclear Information System (INIS)

Barks, D.B.; Kozinsky, E.J.; Eckel, S.

1982-05-01

The purposes of this nuclear plant task analysis pilot study: to demonstrate the use of task analysis techniques on selected abnormal or emergency operation events in a nuclear power plant; to evaluate the use of simulator data obtained from an automated Performance Measurement System to supplement and validate data obtained by traditional task analysis methods; and to demonstrate sample applications of task analysis data to address questions pertinent to nuclear power plant operational safety: control room layout, staffing and training requirements, operating procedures, interpersonal communications, and job performance aids. Five data sources were investigated to provide information for a task analysis. These sources were (1) written operating procedures (event-based); (2) interviews with subject matter experts (the control room operators); (3) videotapes of the control room operators (senior reactor operators and reactor operators) while responding to each event in a simulator; (4) walk-/talk-throughs conducted by control room operators for each event; and (5) simulator data from the PMS

Real-time stability monitoring method for boiling water reactor nuclear power plants

International Nuclear Information System (INIS)

Fukunishi, K.; Suzuki, S.

1987-01-01

A method for real-time stability monitoring is developed for supervising the steady-state operation of a boiling water reactor core. The decay ratio of the reactor power fluctuation is determined by measuring only the output neutron noise. The concept of an inverse system is introduced to identify the dynamic characteristics of the reactor core. The adoption of an adaptive digital filter is useful in real-time identification. A feasibility test that used measured output noise as an indication of reactor power suggests that this method is useful in a real-time stability monitoring system. Using this method, the tedious and difficult work for modeling reactor core dynamics can be reduced. The method employs a simple algorithm that eliminates the need for stochastic computation, thus making the method suitable for real-time computation with a simple microprocessor. In addition, there is no need to disturb the reactor core during operation. Real-time stability monitoring using the proposed algorithm may allow operation under less stable margins

Fuel handling grapple for nuclear reactor plants

International Nuclear Information System (INIS)

Rousar, D.L.

1992-01-01

This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2000-01-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Analysis of water hammer in control rod drive systems of boiling water reactor nuclear power plants

International Nuclear Information System (INIS)

Safwat, H.H.; Arastu, A.H.; Lau, S.

1983-01-01

The method of characteristics is applied to analyze water hammer in BWR (Boiling Water Reactor) Control Rod Drive (CRD) Systems following fast opening of scram valves. The modelling of the CRD mechanism is presented. Numerical predictions are compared to experimental data. (author)

NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

International Nuclear Information System (INIS)

1994-08-01

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Water quality control method and device for nuclear power plant and nuclear power plant

International Nuclear Information System (INIS)

Nagase, Makoto; Asakura, Yamato; Uetake, Naoto; Sawa, Toshio; Uchida, Shunsuke; Takeda, Renzo; Osumi, Katsumi.

1993-01-01

In a BWR type nuclear power plant, water quality of coolants is controlled so as to lower deposition rate of Co ions in reactor water on a fuel cladding tube. The water quality control method includes (1) decreasing an iron concentration in feedwater to less than 0.1ppb, (2) adjusting coolants weakly acidic and (3) controlling dissolved oxygen concentration in reactor water to 20ppb. This can decrease 60 Co ion concentration even if 60 Co ion concentration is increased by the change of environment for the operation in future, such as an operation with hydrogen injection and extention of fuel burnup degree. (T.M.)

Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

International Nuclear Information System (INIS)

2012-03-01

The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

Corrosion problems in boiling water reactors and their remedies

International Nuclear Information System (INIS)

Rosborg, B.

1989-01-01

This article briefly presents current corrosion problems in boiling water reactors and their remedies. The problems are different forms of environmentally assisted cracking, and the remedies are divided into material-, environment-, and stress-related remedies. The list of problems comprises: intergranular stress corrosion cracking (IGSCC) in weld-sensitized stainless steel piping; IGSCC in cold-bent stainless steel piping; irradiation-assisted stress corrosion cracking (IASCC) in stainless alloys; IGSCC in high-strength stainless alloys. A prospective corrosion problem, as judged from literature references, and one which relates to plant life, is corrosion fatigue in pressure vessel steel, since the reactor pressure vessel is the most critical component in the BWR pressure boundary as regards plant safety. (author)

Project of SVBR-75/100 reactor plant with improved safety for nuclear sources of small and medium power

International Nuclear Information System (INIS)

Dragunov, Yu. G.; Stepanov, V. S.; Klimov, N. N.; Dedul, A. V.; Bolvanchikov, S. N.; Zrodnikov, A. V.; Tolhinsky, G. I.; Komlev, O. G.

2004-01-01

As a result of the joint work performed recently by FSUE OKB Gidropress, SNC RF-IPPE and other organizations the technical feasibility is shown for creation and usage in nuclear power engineering of the unified reactor plant (RP) SVBR-75/100 with fast neutron reactor core and lead-bismuth coolant (LBC) in the primary circuit. Technical design of SVBR-75/100 reactor plant is based on the following: 50-year operation experience in development and operation of RP with LBC for nuclear submarines; experience in development and operation of fast reactor with sodium coolant; experience in optimization of LBC technology at nuclear submarines and ground-based test benches; conceptual design of SVBR-75 reactor plant (for renovation of Units 2, 3 and 4 of Novovoronezh NPP). Technical solutions laid down in the basis of SVBR-75/100 reactor plant design are oriented towards the industrial basis, structural materials existing in Russia, as well as the unique LBC technology with experimental and practical support. The concept of SVBR-75/100 reactor plant safety assurance is based on the following provisions: maximum usage of inherent safety supported by physical features of fast neutron reactor, chemically inert LBC in the primary circuit, integral layout and special design solutions; maximum possible combination of normal operation and safety functions in RP systems. Small power of SVBR-75/100 RP makes it possible to manufacture the complete set of RP main equipment at the factory and delivery it to NPP site as-finished practically using any transport including railway. Possible fields of application of SVBR-75/100 reactor plants: modular NPPs of different power; renovation of NPPs with light water reactors exhausted their service life; independent nuclear power sources for different applications (ground-based nuclear water-desalinating plants, etc. )(author)

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)

Characteristics and economy of the European reactor of pressurized water (EPR)

International Nuclear Information System (INIS)

Ortiz V, J.; Ramirez S, J.R.; Palacios H, J.C.

2005-01-01

The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

Long term review of research on light water reactor types

International Nuclear Information System (INIS)

Sumiya, Yutaka

1982-01-01

In Japan, 24 nuclear power plants of 17.18 million kWe capacity are in operation, and their rate of operation has shown the good result of more than 60% since 1980. One of the research on the development of light water reactors is the electric power common research, which was started in 1976, and 272 researches were carried out till 1982. It contributed to the counter-measures to stress corrosion cracking, thermal fatigue and the thinning of steam generator tubes, to the reduction of crud generation and the remote control and automation of inspection and maintenance, and to the verification of safety. The important items for the future are the cost down of nuclear power plant construction, the development of robots for nuclear power plants, the improvement of the ability to follow load variation, and the development of light water reactors of new types. It is necessary to diversify the types of reactors to avoid the effect of a serious trouble which may occur in one type of reactors. Tokyo Electric Power Co., Inc., thinks that the Japanese type PWRs having the technical features of KWU type PWRs are desirable for the future development. The compatibility with the condition of installation permission in Japan, the required design change and the economy of the standard design PWRs of KWU (1.3 million kW) have been studied since October, 1981, by KWU and three Japanese manufacturers. (Kako, I.)

Extended fuel cycle operation for pressurized water reactor plants

International Nuclear Information System (INIS)

Silvestri, G.J. Jr.

1978-01-01

A nuclear steam turbine power plant system having an arrangement therein for extended fuel cycle operation is described. The power plant includes a turbine connected at its inlet to a source of motive fluid having a predetermined pressure associated therewith. The turbine has also connected thereto an extraction conduit which extracts steam from a predetermined location therein for use in an associated apparatus. A bypass conduit is provided between a point upstream of the inlet and the extraction conduit. A flow control device is provided within the bypass conduit and opens when the pressure of the motive steam supply drops beneath the predetermined pressure as a result of reactivity loss within the nuclear reactor. Opening of the bypass conduit provides flow to the associated apparatus and at the same time provides an increased flow orifice to maintain fluid flow rate at a predetermined level

Plant maintenance and advanced reactors, 2006

Energy Technology Data Exchange (ETDEWEB)

Agnihotri, Newal (ed.)

2006-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

Dynamic model for a boiling water reactor

International Nuclear Information System (INIS)

Muscettola, M.

1963-07-01

A theoretical formulation is derived for the dynamics of a boiling water reactor of the pressure tube and forced circulation type. Attention is concentrated on neutron kinetics, fuel element heat transfer dynamics, and the primary circuit - that is the boiling channel, riser, steam drum, downcomer and recirculating pump of a conventional La Mont loop. Models for the steam and feedwater plant are not derived. (author)

Domestic manufacturing and reliability improvement of reactor water recirculation equipment

International Nuclear Information System (INIS)

Kobayashi, Hidekazu; Oi, Masao; Shida, Toichi; Yokomori, Takashi

1982-01-01

The reactor coolant recirculation system is one of the important systems to control the reactor output in BWR nuclear power plants. Its components require high reliability and maintainability as well as controllability. For many Japanese nuclear power plants, recirculation pumps, fluid couplings and others have been imported so far. Hitachi Ltd. has established a domestic manufacturing organization through the development and test of these equipment. The fundamental design conditions for these equipment are the improvement of the rate of utilization of plant facility, the capability to follow load, and output power stability. In this paper, the specifications, the investigation of moment of inertia and the design features of recirculation pumps, driving motors and variable frequency power supply systems are described. The paper also reports on the combination test implemented to evaluate the recirculation system. The combination test includes the test using water rheostat for the power source facility and the loading test for a recirculation pump. The application of those system equipment to an actual plant was analyzed and evaluated on a basis of the test data obtained. The result showed that the equipment can achieve the rate of change of reactor power of 30%/min. Those equipment have been employed for No. 2 reactor plant of the Fukushima No. 2 Nuclear Power Station, the Tokyo Electric Power Co., Inc. (Wakatsuki, Y.)

The European Pressurized Water Reactor (EPR). State of the art after the preliminary design phase

International Nuclear Information System (INIS)

Bouteille, F.; Schneider, D.

2002-01-01

The European Pressurized Water Reactor (EPR) is an evolutionary development of the pressurized water reactor product lines built by Framatome and Siemens in France and Germany. Under the technical leadership of both nuclear power plant suppliers (now merged in Framatome ANP, a joint venture of AREVA and Siemens) the future-oriented plant concepts was developed in close cooperation with German and French utilities and in compliance with the European Utility Requirements. The EPR has safety features with which even extremely improbable, beyond design-basis events can be controlled and their effects can be limited to such an extent that no emergency response actions need be taken outside of the immediate plant site. This also means that safety systems prevent containment failure even in the improbable case of a core melt. This was confirmed by the French and German reactor safety authorities. The selected high thermal output also insures the economic viability of the innovative reactor concept, so that the power generation costs which can be achieved with the EPR will be absolutely competitive with those of fossil energy carriers. Framatome ANP has thus developed a pressurized water reactor ready for offer at the right time, which can completely fulfill the most rigorous requirements in terms of nuclear safety and economy. (Author)

Safety system in a heavy water detritiation plant

International Nuclear Information System (INIS)

Balteanu, O.; Stefan, I.; Retevoi, C.

2003-01-01

In a CANDU 6 type reactor a quantity of 55·10 15 Bq/year of tritium is generated, 95% being in the D 2 O moderator which can achieve a radioactivity of 2.5-3.5·10 12 Bq/kg. Tritium in heavy water contributes with 30-50% to the doses received by operation personnel and up to 20% to the radioactivity released in the environment. The large quantity of heavy water used in this type of reactors (500 tones) make storage very difficult, especially for environment. The extraction of tritium from tritiated heavy water of CANDU reactors solve the following problems: the radiation level in the operation area, the costs of maintenance and repair reduction due to reduction of personnel protection measures, the increase of NPP utilisation factor by shutdown time reduction for maintenance and repair, use the extracted tritium for fusion reactors and not for the last, lower costs and risk for storage heavy water waste. Heavy water detritiation methods, which currently are used in the industrial or experimental plant, are based on catalytic isotope exchange or electrolysis followed cryogenic distillation or permeation. The technology developed at Institute of Cryogenics and Isotope Separation is based upon catalytic exchange between tritiated water and deuterium, followed by cryogenic distillation of hydrogen isotopes. The nature of the fluids that are processed in detritiation requires the operation of the plant in safety conditions. The paper presents the safety system solution chose in order to solve this task, as well as a simulation of an incident and safety system response. The application software is using LabView platform that is specialised on control and factory automation applications. (author)

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

The installation welding of pressure water reactor coolant piping

International Nuclear Information System (INIS)

Deng Feng

2010-01-01

Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

On the reliability of steam generator performance at nuclear power plants with WWER type reactors

International Nuclear Information System (INIS)

Styrikovich, M.A.; Margulova, T.Kh.

1974-01-01

The problem of ensuring reliable operation of steam generators in a nuclear power plant with a water-cooled, water-moderated reactor (WWER) was studied. At a nuclear power plant with a vertical steam generator (specifically, a Westinghouse product) the steam generator tubes were found to have been penetrated. Shutdown was due to corrosion disintegration of the austenitic stainless steel, type 18/8, used as pipe material for the heater surface. The corrosion was the result of the action of chlorine ions concentrated in the moisture contained in the iron oxide films deposited in low parts of the tube bundle, directly at the tube plate. Blowing through did not ensure complete removal of the film, and in some cases the construction features of the steam generator made removal of the film practically impossible. Replacement of type 18/8 stainless steel by other construction material, e.g., Inconel, did not give good results. To ensure reliable operation of vertical steam generators in domestic practice, the generators are designed without a low tube plate (a variant diagram of the vertical steam generator of such construction for the water-cooled, water-moderated reactor 1000 is presented). When low tube plates are used the film deposition is intolerable. For organization of a non-film regime a complex treatment of the feed water is used, in which the amount of complexion is calculated from the stoichmetric ratios with the composition of the feed water. It is noted that, if 100% condensate purification is used with complexon processing of the feed water to the generator, we can calculate the surface of the steam-generator heater without considering the outer placement on the tubes. In this the cost of the steam generator and all the nuclear power plants with WWER type reactors is decreased even with installation of a 100% condensate purification. It is concluded that only simultaneous solution of construction and water-regime problems will ensure relaible operation of

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

Energy Technology Data Exchange (ETDEWEB)

Hanson, Duane J. (Idaho National Engineering Laboratory, Idaho Falls, ID 83415 (United States)); Arcieri, William C. (Idaho National Engineering Laboratory, Idaho Falls, ID 83415 (United States)); Ward, Leonard W. (Idaho National Engineering Laboratory, Idaho Falls, ID 83415 (United States))

1994-07-01

A five-step methodology was developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information that personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment. ((orig.))

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

International Nuclear Information System (INIS)

Hanson, Duane J.; Arcieri, William C.; Ward, Leonard W.

1994-01-01

A five-step methodology was developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information that personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment. ((orig.))

Penn State advanced light water reactor concept

International Nuclear Information System (INIS)

Borkowski, J.A.; Smith, K.A.; Edwards, R.M.; Robinson, G.E.; Schultz, M.A.; Klevans, E.H.

1987-01-01

The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

Aging assessment of Residual Heat Removal systems in Boiling Water Reactors

International Nuclear Information System (INIS)

Lofaro, R.J.; Aggarwal, S.

1992-01-01

The effects of aging on Residual Heat Removal systems in Boiling Water Reactors have been studied as part of the Nuclear Plant Aging Research Program. The aging phenomena has been characterized by analyzing operating experience from various national data bases. In addition, actual plant data was obtained to supplement and validate the data base findings

Water treatment process for nuclear reactors

International Nuclear Information System (INIS)

Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

1992-01-01

Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigate water treatment process for nuclear reactor utilization. Analysis of output water chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerates to obtain the optimum quantity of pure water which reached to 15 cubic meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30 %. output water chemistry agree with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined.5 fig., 3 tab

Cs-137 and Co-60 concentrations in water from the Savannah River and water-treatment plants downstream of SRP

International Nuclear Information System (INIS)

1983-01-01

In preparation for restart of L-Reactor, a comprehensive environmental sampling and analysis program was initiated in March 1983 to determine Cs-137 concentrations in off-site water downstream from Savannah River Plant (SRP). Concentrations of Co-60 also are determined in this sampling and analysis program. This report summarizes the first three months of results. Cesium-137 concentrations are reported for finished water from the Beaufort-Jasper, Port Wentworth and North Augusta water treatment plants for weekly continuous samples during April through June 1983. The very low concentrations of cesium-137 in finished water from downstream water treatment plants showed significant changes during this time. The changes in concentration occurred smoothly and correlate with changes in river flow. No changes in concentration during April through June can be attributed to L-Reactor's only cold water test which occurred June 8 and 9. No Co-60 was observed in any samples

Advanced light water reactor program at ABB-Combustion Engineering Nuclear Power

International Nuclear Information System (INIS)

Cahn, H.

1990-01-01

To meet the needs of Electric Utilities ordering nuclear power plants in the 1990s, ABB-Combustion Engineering is developing two designs which will meet EPRI consensus requirements and new licensing issues. The System 80 Plus design is an evolutionary pressurized water reactor plant modelled after the successful System 80 design in operation in Palo Verde and under construction in Korea. System Plus is currently under review by the US Nuclear Regulatory Commission with final design approval expected in 1991 and design certification in 1992. The Safe Integral Reactor (SIR) plant is a smaller facility with passive safety features and modular construction intended for design certification in the late 1990s. (author)

Water chemistry features of advanced heavy water reactor

International Nuclear Information System (INIS)

Sriram, Jayasree; Vijayan, K.; Kain, Vivekanad; Velmurugan, S.

2015-01-01

Advanced Heavy Water Reactor (AHWR) being designed in India proposes to use Plutonium and Thorium as fuel. The objective is to extract energy from the uranium-233 formed from Thorium. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a natural circulation reactor. Thus, it has got several advanced passive safety features built into the system. The various water coolant systems are listed below. i) Main Heat transport System ii) Feed water system iii) Condenser cooling system iv) Process water system and safety systems. As it is a tube type reactor, the radiolysis control differs from the normal boiling water reactor. The coolant enters the bottom of the coolant channel, boiling takes place and then the entire steam water mixture exits the core through the long tail pipes and reaches the moisture separator. Thus, there is a need to devise methods to protect the tail pipes from oxidizing water chemistry condition. Similarly, the moderator heavy water coolant chemistry differs from that of moderator system chemistry of PHWR. The reactivity worth per ppm of gadolinium and boron are low in comparison to PHWR. As a result, much higher concentration of neutron poison has to be added for planned shutdown, start up and for actuating SDS-2. The addition of higher concentration of neutron poison result in higher radiolytic production of deuterium and oxygen. Their recombination back to heavy water has to take into account the higher production of these gases. This paper also discusses the chemistry features of safety systems of AHWR. In addition, the presentation will cover the chemistry monitoring methodology to be implemented in AHWR. (author)

Chilled water optimization at Beek INEOS PVC Plant : ammonia cycle

NARCIS (Netherlands)

Karami Alaghinloo, B.

2012-01-01

In BEEK INEOS PVC plant, polymerization takes place in a suspension process in twenty reactors in five lines. As the reaction is exothermic, a 17MW chilled water unit (CWU) removing heat from reactors which are producing different grades in batch processes. The objective of the project was to

Towards intrinsically safe light-water reactors

Energy Technology Data Exchange (ETDEWEB)

Hannerz, K

1983-07-01

Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification.

Towards intrinsically safe light-water reactors

International Nuclear Information System (INIS)

Hannerz, K.

1983-07-01

Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification

Technical update on pressure suppression type containments in use in U.S. light water reactor nuclear power plants

International Nuclear Information System (INIS)

1978-07-01

In 1972, Dr. S. H. Hanauer (Technical Advisor to the NRC's Executive Director for Operations) wrote a memorandum that raised several questions on the viability of pressure suppression containment concepts. The concerns raised by Dr. Hanauer have recently become the subject of considerable discussion by several members of the U.S. Congress and public. The report provides a response to these expressed concerns and a status summary for various technical matters that relate to the safety of pressure suppression type containments for light water cooled reactor plants

Reactor type choice and characteristics for a small nuclear heat and electricity co-generation plant

International Nuclear Information System (INIS)

Liu Kukui; Li Manchang; Tang Chuanbao

1997-01-01

In China heat supply consumes more than 70 percent of the primary energy resource, which makes for heavy traffic and transportation and produces a lot of polluting materials such as NO x , SO x and CO 2 because of use of the fossil fuel. The utilization of nuclear power into the heat and electricity co-generation plant contributes to the global environmental protection. The basic concept of the nuclear system is an integral type reactor with three circuits. The primary circuit equipment is enclosed in and linked up directly with reactor vessel. The third circuit produces steam for heat and electricity supply. This paper presents basic requirements, reactor type choice, design characteristics, economy for a nuclear co-generation plant and its future application. The choice of the main parameters and the main technological process is the key problem of the nuclear plant design. To make this paper clearer, take for example a double-reactor plant of 450 x 2MW thermal power. There are two sorts of main technological processes. One is a water-water-steam process. Another is water-steam-steam process. Compared the two sorts, the design which adopted the water-water-steam technological process has much more advantage. The system is simplified, the operation reliability is increased, the primary pressure reduces a lot, the temperature difference between the secondary and the third circuits becomes larger, so the size and capacity of the main components will be smaller, the scale and the cost of the building will be cut down. In this design, the secondary circuit pressure is the highest among that of the three circuits. So the primary circuit radioactivity can not leak into the third circuit in case of accidents. (author)

Knowledges and abilities catalog for nuclear power plant operators: pressurized water reactors

International Nuclear Information System (INIS)

1985-07-01

This document catalogs roughly 5300 knowledges and abilities of reactor operators and senior reactor operators. It results from a reanalysis of much larger job-task analysis data base compiled by the Institute of Nuclear Power Operations (INPO). Knowledges and abilities are cataloged for 45 major power plant systems and 38 emergency evolutions, grouped according to 11 fundamental safety functions (e.g., reactivity control and reactor coolant system inventory control). With appropriate sampling from this catalog, operator licensing examinations having content validity can be developed. A structured sampling procedure for this catalog is under development by the Nuclear Regulatory Commission (NRC) and will be published as a companion document, ''Examiners' Handbook for Developing Operator Licensing Examinations'' (NUREG-1121). The examinations developed by using the catalog and handbook will cover those topics listed under Title 10, Code of Federal Regulations, Part 55

Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

International Nuclear Information System (INIS)

1977-06-01

ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

Topical papers on heavy water, fuel fabrication and reactors

International Nuclear Information System (INIS)

1978-01-01

A total of four papers is presented. The first contribution of the Federal Republic of Germany reviews the market situation for reactors and the relations between reactor producers and buyers as reflected in sales agreements. The second West German contribution gives a world-wide survey of fuel element production as well as of fuel and fuel element demand up to the year 2000. The Canadian paper discusses the future prospects of heavy-water production, while the Ecuador contribution deals with small and medium-sized nuclear power plants

Fault tolerant digital control systems for boiling water reactors

International Nuclear Information System (INIS)

Chakraborty, S.; Cash, N.R.

1986-01-01

In a Boiling Water Reactor nuclear power plant, the power generation control function is divided into several systems, each system controlling only a part of the total plant. Presently, each system is controlled by conventional analog or digital logic circuits with little interaction for coordinated control. The advent of microprocessors has allowed the development of distributed fault-tolerant digital controls. The objective is to replace these conventional controls with fault-tolerant digital controls connected together with digital communication links to form a fully integrated nuclear power plant control system

Water treatment process for nuclear reactors

International Nuclear Information System (INIS)

Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

1993-01-01

Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigates water treatment process for nuclear reactor utilization. Analysis of outwater chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerants to obtain the optimum quantity of pure water which reached to 15 cubic-meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30%. Output water chemistry agrees with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined

IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region

Energy Technology Data Exchange (ETDEWEB)

Alonso, G.; Ramirez, R.; Gomez, C.; Viais, J.

2004-10-03

The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region.

IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region

International Nuclear Information System (INIS)

Alonso, G.; Ramirez, R.; Gomez, C.; Viais, J.

2004-01-01

The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region

Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Solomon, K.A.

1979-07-01

In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

[Research of input water ratio's impact on the quality of effluent water from hydrolysis reactor].

Science.gov (United States)

Liang, Kang-Qiang; Xiong, Ya; Qi, Mao-Rong; Lin, Xiu-Jun; Zhu, Min; Song, Ying-Hao

2012-11-01

Based on high SS/BOD and low C/N ratio of waste water of municipal wastewater treatment plant, the structure of currently existing hydrolysis reactor was reformed to improve the influent quality. In order to strengthen the sludge hydrolysis and improve effluent water quality, two layers water distributors were set up so that the sludge hydrolysis zone was formed between the two layers distribution. For the purpose of the hydrolysis reactor not only plays the role of the primary sedimentation tank but also improves the effluent water biodegradability, input water ratios of the upper and lower water distributor in the experiment were changed to get the best input water ratio to guide the large-scale application of this sort hydrolysis reactor. Results show, four kinds of input water ratio have varying degrees COD and SS removal efficiency, however, input water ratio for 1 : 1 can substantially increase SCOD/COD ratio and VFA concentration of effluent water compared with the other three input water ratios. To improve the effluent biodegradability, input water ratio for 1 : 1 was chosen for the best input water ratio. That was the ratio of flow of upper distributor was 50%, and the ratio of the lower one was 50%, at this case it can reduce the processing burden of COD and SS for follow-up treatment, but also improve the biodegradability of the effluent.

Generic safety issues for nuclear power plants with pressurized heavy water reactors and measures for their resolution

International Nuclear Information System (INIS)

2007-06-01

be used in reassessing the safety of individual operating plants. In 1998, the IAEA completed IAEA-TECDOC-1044 entitled Generic Safety Issues for Nuclear Power Plants with Light Water Reactors and Measures Taken for their Resolution and established the associated LWRGSIDB database (Computer Manual Series No. 13). The present compilation, which is based on broad international experience, is an extension of this work to cover pressurized heavy water reactors (PHWRs). As in the case of LWRs, it is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. It addresses generic safety issues identified in nuclear power plants using PHWRs. In most cases, the measures taken or planned to resolve these issues are also identified. The work on this report was initiated by the Senior Regulators of Countries Operating CANDU-Type Nuclear Power Plants at one of their annual meetings. It was carried out within the framework of the IAEA's programme on National Regulatory Infrastructure for Nuclear Installation Safety and serves to enhance regulatory effectiveness through the exchange of safety related information

Investigation and evaluation of stress-corrosion cracking in piping of light water reactor plants

International Nuclear Information System (INIS)

1979-01-01

In 1975, a Pipe Cracking Study Group, established by the United States Nuclear Regulatory Commission (USNRC), reviewed intergranular stress-corrosion cracking (IGSCC) in Bioling Water Reactors (BWRs) and issued a report. During 1978, IGSCC was reported for the first time in large-diameter piping (> 20 in.) in a BWR in Germany. This discovery, together with the reported questions concerning the interpretation of ultrasonic inspections, led to the activation of a new Pipe Crack Study Group (PCSG) by USNRC. The charter of the new PCSG was expanded: (1) to include review of potential for stress-corrosion cracking in Pressurized Water Reactors (PWRs) as well as BWRs, (2) to examine operating experience in foreign reactors relevant to IGSCC, and (3) to study five specific questions. The PCSG limited the scope of the study to BWR and PWR piping runs and safe ends attached to the reactor pressure vessel. Not considered were components such as the reactor pressure vessel, pumps, valves, steam generators, large steam turbines, etc. Throughout this report, as well as in the title, the safe ends are arbitrarily defined as piping

Water quality monitoring device for nuclear power plant

International Nuclear Information System (INIS)

Kubo, Mitsushi.

1995-01-01

The device of the present invention measures quality of feedwater after heated in a regenerative heat exchanger device of a coolant cleanup system in a BWR type reactor, to detect ions generated from organic materials decomposed at high temperature and specify the position where impurities are formed. Namely, in a power plant having a reactor coolant cleanup pipeline connected to a feedwater pipeline, a water quality measuring portion is disposed to the feedwater system at the downstream of the junction to the feedwater system pipeline. A water quality sample is taken to measure the water quality in a state where the feedwater heated by a feedwater heater and flowing to the reactor, and the cleanup coolants heated by the regenerative heat exchanger are mixed. Thus, the impurities formed at the down stream of the feedwater system pipeline, as well as the water quality including impurities decomposed in a high temperature state can be measured. (I.S.)

Risk contribution from low power and shutdown of a pressurized water reactor

International Nuclear Information System (INIS)

Chu, T.L.; Pratt, W.T.

1997-01-01

During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (a pressurized water reactor) and Grand Gulf (a boiling water reactor), were selected for study by Brookhaven National Laboratory and Sandia National Laboratories, respectively. The program objectives included assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing estimated core damage frequencies, important accident sequences, and other qualitative and quantitative results with full power accidents as assessed in NUREG-1150. The scope included a Level 3 PRA for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. 12 refs., 7 tabs

Development status and application prospect of supercritical-pressure light water cooled reactor

International Nuclear Information System (INIS)

Li Manchang; Wang Mingli

2006-01-01

The Supercritical-pressure Light Water Cooled Reactor (SCWR) is selected by the Generation IV International Forum (GIF) as one of the six Generation IV nuclear systems that will be developed in the future, and it is an innovative design based on the existing technologies used in LWR and supercritical coal-fired plants. Technically, SCWR may be based on the design, construction and operation experiences in existing PWR and supercritical coal-fired plants, which means that there is no insolvable technology difficulties. Since PWR technology will be adopted in the near term and medium term projects in China, and considering the sustainable development of the technology, it is an inevitable choice to research and develop the nuclear system of supercritical light water cooled reactor. (authors)

Advanced ceramic cladding for water reactor fuel

International Nuclear Information System (INIS)

Feinroth, H.

2000-01-01

Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

Evolutionary water cooled reactors: Strategic issues, technologies and economic viability. Proceedings of a symposium

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-12-01

Symposium on evolutionary water cooled reactors: Strategic issues, technologies and economic viability was intended for managers in utilities, reactor design organizations and hardware manufacturing companies and for government decision makers who need to understand technological advances and the potential of evolutionary water cooled reactors to contribute to near and medium term energy demands. The topics addressed include: strategic issues (global energy outlook, the role of nuclear power in sustainable energy strategies, power generation costs, financing of nuclear plant projects, socio-political factors and nuclear safety requirements); technological advances (instrumentation and control, means od improving prevention and mitigation of severe accidents, development of passive safety systems); keys to economic viability (simplification, standardization, advances in construction and project management, feedback of experience from utilities into new designs, and effective management of plant operation)

Light-water nuclear reactors

International Nuclear Information System (INIS)

Drevon, G.

1983-01-01

This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work [fr

Solid radioactive waste processing system for light water cooled reactor plants

International Nuclear Information System (INIS)

Anon.

1979-01-01

Design, construction and performance requirements are given for the operation of the solid radioactive waste processing system for light water-cooled reactor plants. All radioactive or contaminated materials, including spent air and liquid filter elements, spent bead resins, filter sludge, spent powdered resins, evaporator and reverse osmosis concentrates, and dry radioactive wastes are to be processed in appropriate portions of the system. Sections of the standard cover: overall system requirements; equipment requirement; controls and instrumentation; physical arrangement; system capacity and redundancy; operation and maintenance; and system construction and testing. Provisions contained in this standard are to take precedence over ANS-51.1-1973(N18.2-1973) and its revision, ANS-51.8-1975(N18.2a-1975), Sections 2.2 and 2.3. The product resulting from the solid radioactive waste processing system must meet criteria imposed by standards and regulations for transportation and burial (Title 10, Code of Federal Regulations, Part 71, Title 49, Code of Federal Regulations, Parts 100 to 199). As a special feature, all statements in this standard which are related to nuclear safety are set off in boxes

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

Science.gov (United States)

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

2014-08-01

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

Light water ultra-safe plant concept

International Nuclear Information System (INIS)

Klevans, E.

1989-01-01

Since the accident at Three Mile Island (TMI), Penn State Nuclear Engineering Department Faculty and Staff have considered various methods to improve already safe reactor designs and public perception of the safety of Nuclear Power. During 1987 and 1988, the Department of Energy provided funds to the Nuclear Engineering Department at Penn State to investigate a plant reconfiguration originated by M.A. Schultz called ''The Light Water Ultra-Safe Plant Concept''. This report presents a final summary of the project with references to several masters' theses and addendum reports for further detail. The two year research effort included design verification with detailed computer simulation of: (a) normal operation characteristics of the unique pressurizing concept, (b) severe transients without loss of coolant, (c) combined primary and secondary system modeling, and (d) small break and large break loss of coolant accidents. Other studies included safety analysis, low power density core design, and control system design to greatly simplify the control room and required operator responses to plant upset conditions. The overall conclusion is that a reconfigured pressurized water reactor can achieve real and perceived safety improvements. Additionally, control system research to produce greatly simplified control rooms and operator requirements should be continued in future projects

Performance of materials in the component cooling water systems of pressurized water reactors

International Nuclear Information System (INIS)

Lee, B.S.

1993-01-01

The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed

Nuclear power/water pumping-up composite power plant

International Nuclear Information System (INIS)

Okamura, Kiyoshi.

1995-01-01

In a nuclear power/water pumping-up composite power plant, a reversible pump for pumping-up power generation connected to a steam turbine is connected to an upper water reservoir and a lower water reservoir. A pumping-up steam turbine for driving the turbine power generator, a hydraulic pump for driving water power generator by water flowing from the upper water reservoir and a steam turbine for driving the pumping-up pump by steams from a nuclear reactor are disposed. When power demand is small during night, the steam turbine is rotated by steams of the reactor, to pump up the water in the lower water reservoir to the upper water reservoir by the reversible pump. Upon peak of power demand during day time, power is generated by the steams of the reactor, as well as the reversible pump is rotated by the flowing water from the upper water reservoir to conduct hydraulic power generation. Alternatively, hydraulic power generation is conducted by flowing water from the upper reservoir. Since the number of energy conversion steps in the combination of nuclear power generation and pumping-up power generation is reduced, energy loss is reduced and utilization efficiency can be improved. (N.H.)

Thermal shield support degradation in pressurized water reactors

International Nuclear Information System (INIS)

Sweeney, F.J.; Fry, D.N.

1986-01-01

Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

Reactivity margins in heavy water moderated production reactors

International Nuclear Information System (INIS)

Benton, F.D.

1981-11-01

The design of the reactor core and components of the heavy water moderated reactors at the Savannah River Plant (SFP) can be varied to produce a number of isotopes. For the past decade, the predominant reactor core design has been the enriched-depleted lattice. In this lattice, fuel assemblies of highly enriched uranium and target assemblies of depleted uranium, which produce plutonium, occupy alternate lattice positions. This heterogeneous lattice arrangement and a nonuniform control rod distribution result in a reactor core that requires sophisticated calculational methods for accurate reactivity margin and power distribution predictions. For maximum accuracy, techniques must exist to provide a base of observed data for the calculations. Frequent enriched-depleted lattice design changes are required as product demands vary. These changes provided incentive for the development of techniques to combine the results of calculations and observed reactivity data to accurately and conveniently monitor reactivity margins during operation

CANDU - Canadian experience and expectations with the heavy-water reactor

International Nuclear Information System (INIS)

Foster, J.S.; Russell, S.H.

1977-05-01

The paper describes the evolution of the CANDU nuclear-power plants with particular reference to the objectives of safety, reliability and economy; the development of industrial capacity for the supply of fuel, components and heavy water; and the prospective development of advanced fuel cycles and the projected results. It provides data on radiation, releases, and exposures, internal and external to the power plants; plant availability, capacity factors and other performance data; heavy water production data with reference to safety, reliability, and economics; projections of the performance of CANDU reactors operating on a thorium-U-233 cycle and the development required to establish this cycle; and intent with respct to spent-fuel management and radioactive-waste storage. (author)

Supercritical-pressure light water cooled reactors

CERN Document Server

Oka, Yoshiaki

2014-01-01

This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

Simplified simulation of an experimental fast reactor plant

International Nuclear Information System (INIS)

Fujii, Masaaki; Fujita, Minoru.

1978-01-01

Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

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

International Nuclear Information System (INIS)

Huber, Horacio

1989-01-01

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

Westinghouse Small Modular Reactor balance of plant and supporting systems design

Energy Technology Data Exchange (ETDEWEB)

Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

Westinghouse Small Modular Reactor balance of plant and supporting systems design

International Nuclear Information System (INIS)

Memmott, M. J.; Stansbury, C.; Taylor, C.

2012-01-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

Decontamination flange film characterization for a boiling water reactor under hydrogen water chemistry

International Nuclear Information System (INIS)

Baston, V.F.; Garbauskas, M.F.; Bozeman, J.

1996-01-01

Stainless steel artifacts removed from a boiling water reactor class 4 plant that operated under hydrogen water chemistry and experienced a difficult decontamination were submitted for oxide film characterization. The results reported for the corrosion film composition and structure are consistent with existing theoretical concepts for stainless steel corrosion, spinel structure site preferences (octahedral or tetrahedral) for transition metal ions, and potential-pH diagrams. The observed zinc effects on film stability and lower cobalt incorporation are also consistent with these theoretical concepts

Thermophysical properties database of materials for light water reactors and heavy water reactors. Final report of a coordinated research project 1999-2005

International Nuclear Information System (INIS)

2006-06-01

The IAEA Coordinated Research Project (CRP) on the Establishment of a Thermo-physical Properties Database for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs) started in 1999. It was included in the IAEA's Nuclear Power Programme following endorsement in 1997 by the IAEA's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and the TWG-HWR). Furthermore, the TWG on Fuel Performance and Technology (TWG-FPT) also expressed its support. This CRP was conducted as a joint task within the IAEA's project on technology development for LWRs and HWRs in its nuclear power programme. Improving the technology for nuclear reactors through better computer codes and more accurate materials property data can contribute to improved economics of future plants by helping to remove the need for large design margins, which are currently used to account for limitations of data and methods. Accurate representations of thermo-physical properties under relevant temperature and neutron fluence conditions are necessary for evaluating reactor performance under normal operation and accident conditions. The objective of this CRP was to collect and systematize a thermo-physical properties database for light and heavy water reactor materials under normal operating, transient and accident conditions and to foster the exchange of non-proprietary information on thermo-physical properties of LWR and HWR materials. An internationally available, peer reviewed database of properties at normal and severe accident conditions has been established on the Internet. This report is intended to serve as a useful source of information on thermo-physical properties data for water cooled reactor analyses. The properties data have been initially stored in the THERSYST data system at the University of Stuttgart, Germany, which was subsequently developed into an internationally available Internet database named THERPRO at Hanyang University, Republic of Korea

Role of water lubricated bearings in Candu reactors

International Nuclear Information System (INIS)

Kumar, Ashok N.

1999-01-01

During the twentieth century a great emphasis was placed in understanding and defining the operating regime of oil and grease lubricated components. Major advances have been made through elastohydrodynamic lubrication theory in the quantifying the design life of heavily loaded components such as rolling element bearings and gears. Detailed guidelines for the design of oil and grease lubricated components are widely available and are being applied to the successful design of these components. However similar guidelines for water lubricated components are either not available or not well documented. It is often forgotten that the water was used as a lubricant in several components as far back as 1884 B.C. During the twentieth century the water lubricated components continued to play a major role in some high technology industries such as in the power generation plants. In CANDU nuclear reactors water lubrication of several critical components always occupied a pride place and in most cases the only practical mode of lubrication of several critical components always occupied a pride place and in most cases the only practical mode of lubrication. This paper presents some examples of the major water lubricated components in a CANDU reactors. Major part of the paper is focused on presenting an example of successful operating history of water lubricated bearings used in the HT pumps are presented. Both types of bearings have been qualified by tests for operation under normal as well as under more severe postulated condition of loss-of-coolant-accident (LOCA). These bearings have been designed to operate for the 30 years in the existing CANDU 6 (600 MW) reactors. However for the next generation of CANDU 6 reactors which go into service in the year 2003, the HT pump bearing life has been extended to 40 years. (author)

Common safety approach for future pressurized water reactors in France and Germany

International Nuclear Information System (INIS)

Queniart, D.; Gros, G.

1994-01-01

In France and Germany all major activities related to future pressurized water reactors are now proceeding in a coordinated way among the two countries. This holds for utilities and industry in the development of a joint PWR project, the ''European Pressurized Water Reactor (EPR)'' by Electricite de France (EDF), German utilities, Nuclear Power International (NPI), Framatome and Siemens as well as for the technical safety objectives for future evolutionary reactors on the basis of a common safety approach adopted by the safety authorities of both countries for plants to operate form the beginning of the next century. The proposed paper covers this common development of a safety approach and particular technical safety objectives. (authors). 5 refs. 1 fig

Basic philosophy of the safety design of the Toshiba boiling water reactor

International Nuclear Information System (INIS)

Sato, T.

1992-01-01

This paper discusses the safety design of the Toshiba Boiling Water Reactor (TOSBWR) which was created ∼8 years ago. The design concept is intermediate between conventional boiling water reactors (BWRs) and the advanced BWR (ABWR). It utilizes internal pumps and fine motion control rod drive, but the emergency core cooling system (ECCS) configuration is different from both conventional BWRs and the ABWR. The plant output is 1350 MW (electric). The design is based on two important philosophies: the positive cost reduction philosophy and the constant risk philosophy

Exergy analysis of a system using a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water splitting cycle

International Nuclear Information System (INIS)

Granovskii, M.; Dincer, I.; Rosen, M. A.; Pioro, I

2007-01-01

The power generation efficiency of nuclear plants is mainly determined by the permissible temperatures and pressures of the nuclear reactor fuel and coolants. These parameters are limited by materials properties and corrosion rates and their effect on nuclear reactor safety. The advanced materials for the next generation of CANDU reactors, which employ steam as a coolant and heat carrier, permit the increased steam parameters (outlet temperature up to 625 degree C and pressure of about 25 MPa). Supercritical water-cooled (SCW) nuclear power plants are expected to increase the power generation efficiency from 35 to 45%. Supercritical water-cooled nuclear reactors can be linked to thermochemical water splitting cycles for hydrogen production. An increased steam temperature from the nuclear reactor makes it also possible to utilize its energy in thermochemical water splitting cycles. These cycles are considered by many as one of the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require a heat supply at the temperatures over 550-600 degree C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump which increases the temperature the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. A high temperature chemical heat pump which employs the reversible catalytic methane conversion reaction is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with a SCW nuclear plant on one side and thermochemical water splitting cycle on the other, increases the temperature level of the 'nuclear' heat and, thus, the intensity of

Development of the heavy-water organic-cooled reactor. Status report from the United States of America

Energy Technology Data Exchange (ETDEWEB)

Trilling, C A [Atomics International, Division of North American Aviation, Inc., Canoga Park, CA (United States)

1967-01-01

In late 1964 the United States Atomic Energy Commission decided to undertake the development of the heavy-water-moderated nuclear power reactor as part of its overall programme for the development of advanced converter reactors. The inclusion of the heavy-water reactor concept was based on its indicated potential for achieving: efficient utilization of available fuel resources; generation of low cost electric power; feasibility of scale-up to very large single unit plant sizes for the dual purpose of generating power and desalting sea water. The excellent neutron economy inherent in heavy-water moderation allows a significant increase in the amount of power which can be generated from a given amount of ore. If one takes into account the amount of uranium required not only for burn-up but also to inventory new reactors in a rapidly expanding nuclear economy, heavy-water reactors show the potential of extracting one and a half to two times more power from the ore mined than light-water reactors. Such an improvement in dynamic fuel utilization will postpone the depletion of low cost uranium ore reserves, providing more time for the discovery of new ore resources and the development of economic fast breeder reactors. The excellent neutron economy of the heavy-water reactor also allows the achievement of appreciable burn-up with low enrichment fuel, with consequent low fuel cycle costs and therefore low energy generation costs. These low fuel cycle costs make the economics of this type of reactor rather insensitive to rising ore costs. They also make the concept well suited for the most economic production of the large quantities of heat required for water desalination. The use of individual pressure tubes for circulating the coolant through the reactor vessel lends itself to the development of a modular type design, which can be scaled up to very large single unit plant sizes by simply increasing the number of identical pressure tube modules and the number of coolant

Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

International Nuclear Information System (INIS)

Budd, W.A.

1986-03-01

This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs

Recent performance experience with US light water reactor self-actuating safety and relief valves

Energy Technology Data Exchange (ETDEWEB)

Hammer, C.G.

1996-12-01

Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

Recent advances in severe accident technology - direct containment heating in advanced light water reactors

International Nuclear Information System (INIS)

Fontana, M.H.

1993-01-01

The issues affecting high-pressure melt ejection (HPME) and the consequential containment pressurization from direct containment heating (DCH), as they affect advanced light water reactors (ALWRs), specifically advanced pressurized water reactors (APWRs), were reviewed by the U.S. Department of Energy Advanced Reactor Severe Accident Program (ARSAP). Recommendations from ARSAP regarding the design of APWRs to minimize DCH are embodied within the Electric Power Research Institute ALWR Utility Requirements Document, which specifies (a) a large, strong containment; (b) an in-containment refueling water storage tank; (c) a reactor cavity configuration that minimizes energy transport to the containment atmosphere; and (d) a reactor coolant system depressurization system. Experimental and analytical efforts, which have focused on current-generation plants, and analyses for APWRs were reviewed. Although DCH is a subject of continuous research and considerable uncertainties remain, it is the judgment of the ARSAP that reactors complying with the recommended design requirements would have a low probability of early containment failure due to HPME and DCH

Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

International Nuclear Information System (INIS)

Wati, Nurokhim

2008-01-01

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

Development and Deployment Strategy for a Small Advanced Light Water Reactor

International Nuclear Information System (INIS)

Modro, S. Michael; Reith, Raymond; Babka, Pierre

2002-01-01

This paper discusses development and deployment strategies for the modular Multi-Application Small Light Water Reactor (MASLWR). Modularity, small size, capability to transport whole modules including containment on road or by rail, simplicity and safety of this reactor allows innovative deployment strategies for a variety of applications. A larger plant may be constructed of many independent power generation units. The multi-module plant is intended to be operated as a base-load plant. Each reactor is to be operated at full load. However, in response to changes in power demand individual units can brought on line or shut down. A larger plant can be built in small increments to match the power demand balancing capital commitments with revenues from sales of electricity. Also, an unplanned shutdown of a reactor only affects a relatively small portion of the total plant capacity. Simplification of MASLWR design and extensive use of modularization coupled with factory fabrication will result in improved productivity of fieldwork and improved quality achieved in a factory environment. The initial MASLWR design concept development has been completed under the U.S. DOE (Department of Energy) Nuclear Energy Research Initiative (NERI) project. This paper discusses a strategy for developing and deploying a MASLWR plant by 2015. This schedule is realistic because the plant design relies on existing industrial experience and manufacturing capabilities. The development strategy consists of the following elements: concept confirmation through testing (under the NERI program a scaled integral test facility has been constructed and initial testing performed), design concept optimization, and design certification based on prototype testing. (authors)

Nuclear reactor design

CERN Document Server

2014-01-01

This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.