Summary of the 3rd workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

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)

Jordan Research and Training Reactor (JRTR) Utilization Facilities

International Nuclear Information System (INIS)

Xoubi, N.

2013-01-01

Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U 3 Si 2 ) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×10 14 cm -2 s -1 , and is controlled by a Hafnium control absorber rod and B 4 C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

Press kit. EPR (European pressurized water reactor). The advanced nuclear reactor

International Nuclear Information System (INIS)

2004-10-01

Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21 century, which puts the emphasis on sustainable development. In this framework, this document presents the advantages of the EPR (European Pressurized water Reactor). The EPR is the only third generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. (A.L.B.)

A `big-mac` high converting water reactor

Energy Technology Data Exchange (ETDEWEB)

Ronen, Y; Dali, Y [Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Nuclear Engineering

1996-12-01

Currently an effort is being made to get rid of plutonium. Therefore, at this time, a scientific study of a high converting reactor seems to be out of place. However , it is our opinion that the future of nuclear energy lies, among other things in the clever utilization of plutonium. It is also our opinion that one of the best ways to utilize plutonium is in high converting water reactors (authors).

Current status of nuclear research reactor management and utilization program in Thailand

International Nuclear Information System (INIS)

Aramrattana, M.; Busamongkol, Y.

1999-01-01

The TRR1/M1 is the first research reactor and has been in operational for more than 20 years. During the three decades of research reactor operation in Thailand the utilization of research reactor have been broadened in different fields such as agriculture, medicine and industry. Limitation on utilization of the existing reactor in various fields has led to establishing of a new nuclear research center, Ongkharak Nuclear Research Center (ONRC). The ONRC comprises three major facilities, namely Reactor Island, Isotope Production Facility and Waste Processing and Storage Facility. The reactor itself is a 10 MW TRIGA-type fuels, moderated and cooled by light water with beryllium and heavy water as the reflectors. It is a multi-purpose reactor consisting of different facilities inside and around the core for radioisotope production, medical and industrial uses; and for beam experiments such as High Resolution Powder Diffractometry (HRPD), Neutron Radiography (NR), Prompt Gamma Neutron Activation Analysis (PGNAA), and Boron Neutron Capture Therapy (BNCT). The center is expected to be operational by year 2001. (author)

Improvement in fuel utilization in pressurized heavy water reactors due to increased heavy water purity

International Nuclear Information System (INIS)

Balakrishnan, M.R.

1991-01-01

This paper reports that in a pressurized heavy water reactor (PHWR), the reactivity of the reactor and, consequently, the discharge burnup of the fuel depend on the isotopic purity of the heavy water used in the reactor. The optimal purity of heavy water used in PHWRs, in turn, depends on the cost of fabricated uranium fuel and on the incremental cost incurred in improving the heavy water purity. The physics and economics aspects of the desirability of increasing the heavy water purity in PHWRs in India were first examined in 1978. With the cost data available at that time, it was found that improving the heavy water purity from 99.80% to 99.95% was economically attractive. The same problem is reinvestigated with current cost data. Even now, there is sufficient incentive to improve the isotopic purity of heavy water used in PHWRs. Admittedly, the economic advantage that can be derived depends on the cost of the fabricated fuel. Nevertheless, irrespective of the economics, there is also a fairly substantial saving in natural uranium. That the increase in the heavy water purity is to be maintained only in the low-pressure moderator system, and not in the high-pressure coolant system, makes the option of achieving higher fuel burnup with higher heavy water purity feasible

Development of alternative fuel for pressurized water reactors

International Nuclear Information System (INIS)

Cardoso, P.E.; Ferreira, R.A.N.; Ferraz, W.B.; Lameiras, F.S.; Santos, A.; Assis, G. de; Doerr, W.O.; Wehner, E.L.

1984-01-01

The utilization of alternative fuel cycles in Pressurized Water Reactors (PWR) such as Th/U and Th/Pu cycles can permit a better utilization of uranium reserves without the necessity of developing new power reactor concepts. The development of the technology of alternative fuels for PWR is one of the objectives of the 'Program on Thorium Utilization in Pressurized Water Reactors' carried out jointly by Empresas Nucleares Brasileiras S.A. (NUCLEBRAS), through its Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) and by German institutions, the Julich Nuclear Research Center (KFA), the Kraftwerk Union A.G. (KWU) and NUKEM GmbH. This paper summarizes the results so far obtained in the fuel technology. The development of a fabrication process for PWR fuel pellets from gel-microspheres is reported as well as the design, the specification, and the fabrication of prototype fuel rods for irradiation tests. (Author) [pt

Method of measuring reactor water level

International Nuclear Information System (INIS)

Shinohara, Kaoru.

1979-01-01

Purpose: To provide a water level measuring system so that a reactor water level detecting signal can be corrected in correspondence to a recirculation flow, thereby to carry out a correct water level detection in a wide range of the reactor. Method: According to the operation record of a precursor reactor, the ratio Δh of the lowering of the water level due to the recirculation flow is lowered in proportion to the ratiowith respect to the rated differential pressure of the recirculation flow. Accordingly, the flow of recirculation pump is measured by an elbow differential pressure generator utilizing an elbow of a pipe, and the measured value is multiplied by a gain by a ratio setter, and therefter, an addition computation is carried out by an adder for correcting the signal from a water level detector. When the signal from the water level detector is corrected in this manner, the influence of the lowering of the water level due to the recirculation flow can be removed, and an interlocker predetermined in the defined water level can be actuated, thus the influence of the dynamic pressure due to the recirculation flow acting on the instrumental pipe line detecting the reactor water level can be removed effectively. (Yoshino, Y.)

Feasibility study and economic analysis on thorium utilization in heavy water reactors

International Nuclear Information System (INIS)

1978-07-01

Even though natural uranium is a more easily usable fuel in heavy water reactors, thorium fuel cycles have also been considered owing to certain attractive features of the thorium fuel cycle in heavy water reactors. The relatively higher fission neutron yield per thermal neutron absorption in 233 U combined with the very low neutron absorption cross section of heavy water make it possible to achieve breeding in a heavy water reactor operating on Th- 233 U fuel cycle. Even if the breeding ratio is very low, once a self-sustaining cycle is achieved, thereafter dependence on uranium can be completely eliminated. Thus, with a self-sustaining Th- 233 U fuel cycle in heavy water reactors, a given quantity of natural uranium will be capable of supporting a much larger installed generating capacity to significantly longer period of time. However, since thorium does not contain any fissile isotope, fissile material has to be added at the beginning. Concentrated fissile material is considerably more expensive than the 235 U contained in natural uranium. This makes the fuel cycle cost higher with thorium fuel cycle, at least during the initial stages. The situation is made worse by the fact that, because of its higher thermal neutron absorption cross section, thorium requires a higher concentration of fissile material than 238 U. Nevertheless, because of the superior nuclear characteristics of 233 U, once uranium becomes more expensive, thorium fuel cycle in heavy water reactors may become economically acceptable. Furthermore, the energy that can be made available from a given quantity of uranium is considerably increased with a self-sustaining thorium fuel cycle

Sea water desalination utilizing waste heat by low temperature evaporation

International Nuclear Information System (INIS)

Raha, A.; Srivastava, A.; Rao, I.S.; Majumdar, M.; Srivastava, V.K.; Tewari, P.K.

2007-01-01

Economics of a process is controlled by management of energy and resources. Fresh water has become most valued resource in industries. Desalination is a process by which fresh water resource is generated from sea water or brackish water, but it is an energy intensive process. The energy cost contributes around 25-40% to the total cost of the desalted water. Utilization of waste heat from industrial streams is one of the ecofriendly ways to produce low cost desalted water. Keeping this in mind Low Temperature Evaporation (LTE) desalination technology utilizing low quality waste heat in the form of hot water (as low as 50 deg C) or low pressure steam (0.13 bar) has been developed for offshore and land based applications to produce high purity water (conductivity < 2μS/cm) from sea water. The probability of the scale formation is practically eliminated by operating it at low temperature and controlling the brine concentration. It also does not require elaborate chemical pretreatment of sea water except chlorination, so it has no environmental impact. LTE technology has found major applications in nuclear reactors where large quantity of low quality waste heat is available to produce high quality desalted water for make up water requirement replacing conventional ion exchange process. Successful continuous operation of 30 Te/day LTE desalination plant utilizing waste heat from nuclear research reactor has demonstrated the safety, reliability, extreme plant availability and economics of nuclear desalination by LTE technology. It is also proposed to utilize waste heat from Main Heat Transport (MHT) purification circuit of Advanced Heavy Water Reactor (AHWR) to produce about 250 Te/ day high quality desalinated water by Low Temperature Evaporation (LTE) process for the reactor make up and plant utilization. Recently we have commissioned a 50 Te/day 2-effect low temperature desalination plant with cooling tower where the specific energy and cooling water requirement are

Reactor utilization, Annex A

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1984-01-01

Reactor was operated until August 1984 due to prohibition issued by the Ministry since the reactor does not have the emergency cooling system nor special filters in the ventilation system yet. This means that the operation plan was fulfilled by 69%. This annex includes detailed tables containing data about utilization of reactor experimental channels, irradiated samples, as well as interruptions of operation. Detailed data about reactor power during this period are shown as well

Method of controlling the water quality in nuclear reactors

International Nuclear Information System (INIS)

Ibe, Hidefumi.

1985-01-01

Purpose: To obtain a simple and reliable water quality calculation system and water quality control method based thereon for the entire primary coolant circuits in BWR type reactors. Method: In a method of controlling the water quality of the reactor water by injecting hydrogen into the primary coolant circuits of a nuclear reactor, by utilizing a first linear relationship established between the concentration of oxygen and hydrogen in the main steam system and the concentration of radiolysis products in the reactor core and separators and mixing plenum portions, each of the above-mentioned concentrations is calculated from the concentrations for hydrogen or oxygen. Further, by utilizing the first linear relationship established between the concentrations for the oxygen and hydrogen in the recycling system and the concentration of the radiolysis products in the system from the downcomer to the lower plenum portion, the above-mentioned concentration is calculated from the concentration for oxygen and hydrogen. Then, the hydrogen injection rate into the primary coolant system is determined such that the calculated value takes an aimed value. (Ikeda, J.)

Status of research and development on reduced-moderation water reactors

International Nuclear Information System (INIS)

Iwamura, Takamichi

2002-01-01

To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

Status of research and development on reduced-moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

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

Utilization of research reactors - A global perspective

International Nuclear Information System (INIS)

Muranaka, R.G.

1988-01-01

This paper presents 1) a worldwide picture of research reactors, operable, shutdown, under construction and planned, 2) statistics on utilization of research reactors including TRIGA reactors, and 3) some results of a survey conducted during 1988 on the utilization of research reactors in developing Member States in the Asia-Pacific Region

Summary of the 4th workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

2001-09-01

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

International Nuclear Information System (INIS)

Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

2008-01-01

The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

The Nuclear option for U.S. electrical generating capacity additions utilizing boiling water reactor technology

International Nuclear Information System (INIS)

Garrity, T.F.; Wilkins, D.R.

1993-01-01

The technology status of the Advanced Boiling Water (ABWR) and Simplified Boiling Water (SBWR) reactors are presented along with an analysis of the economic potential of advanced nuclear power generation systems based on BWR technology to meet the projected domestic electrical generating capacity need through 2005. The forecasted capacity needs are determined for each domestic North American Electric Reliability Council (NERC) region. Extensive data sets detailing each NERC region's specific generation and load characteristics, and capital and fuel cost parameters are utilized in the economic analysis of the optimal generation additions to meet this need by use of an expansion planning model. In addition to a reference case, several sensitivity cases are performed with regard to capital costs and fuel price escalation

Summary of the 4th workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-09-01

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

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.)

EPR (European Pressurized water Reactor) The advanced nuclear reactor

International Nuclear Information System (INIS)

2004-01-01

Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

Effective utilization and management of research reactors

Energy Technology Data Exchange (ETDEWEB)

Muranaka, R [International Atomic Energy Agency, Vienna (Austria). Div. of Research and Isotopes

1984-06-01

The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors.

Effective utilization and management of research reactors

International Nuclear Information System (INIS)

Muranaka, R.

1984-01-01

The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors

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.)

Pressurized water reactors: the EPR project

International Nuclear Information System (INIS)

Py, J.P.; Yvon, M.

2007-01-01

EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

Thorium in heavy water reactors

International Nuclear Information System (INIS)

Andersson, G.

1984-12-01

Advanced heavy water reactors can provide energy on a global scale beyond the foreseeable future. Their economic and safety features are promising: 1. The theoretical feasibility of the Self Sufficient Equilibrium Thorium (SSET) concept is confirmed by new calculations. Calculations show that the adjuster rod geometry used in natural uranium CANDU reactors is adequate also for SSET if the absorption in the rods is graded. 2. New fuel bundle designs can permit substantially higher power output from a CANDU reactor. The capital cost for fuel, heavy water and mechanical equipment can thereby be greatly reduced. Progress is possible with the traditional fuel material oxide, but the use of thorium metal gives much larger effects. 3. A promising long range possibility is to use pressure tanks instead of pressure tubes. Heat removal from the core is facilitated. Negative temperature and void coefficients provide inherent safety features. Refuelling under power is no longer needed if control by moderator displacement is used. Reduced quality demand on the fuel permits lower fuel costs. The neutron economy is improved by the absence of pressure and clandria tubes and also by the use of radial and axial blankets. A modular seed blanket design can reduce the Pa losses. The experience from construction of tank designs is good e.g. AAgesta, Attucha. It is now also possible to utilize technology from LWR reactors and the implementation of advanced heavy water reactors would thus be easier than HTR or LMFBR systems. (Author)

Utilization of research reactors

International Nuclear Information System (INIS)

1962-01-01

About 200 research reactors are now in operation in different parts of the world, and at least 70 such facilities, which are in advanced stages of planning and construction, should be critical within the next two or three years. In the process of this development a multitude of problems are being encountered in formulating and carrying out programs for the proper utilization of these facilities, especially in countries which have just begun or are starting their atomic energy work. An opportunity for scientific personnel from different Member States to discuss research reactor problems was given at an international symposium on the Programing and Utilization of Research Reactors organized by the Agency almost immediately after the General Conference session. Two hundred scientists from 35 countries, as well as from the European Nuclear Energy Agency and EURATOM, attended the meeting which was held in Vienna from 16 to 21 October 1961

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

Science.gov (United States)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

International Nuclear Information System (INIS)

Permana, Sidik

2009-01-01

Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

The plutonium utilization in thermal and fast reactor in Japan

International Nuclear Information System (INIS)

Amanuma, T.; Uematsu, K.

1977-01-01

The nuclear power development in Japan is rather extensive one, and the installed nuclear power capacity is expected to reach 49,000 MWe by 1985 and possibly to reach 170,000 MWe by 2000 according to a prediction. Currently istalled nuclear power is mainly based on Light Water type Reactor, and this trend is expected to persist for the time-being. The plutonium produced by LWR will be accumulated to 20 tons by 1985 and to more than 200 tons by 2000. If the produced plutonium will simply be stored, it will raise the economic pressure to utilities and the management and physical protection problems associated with plutonium storing. Therefore, it is not too wise simply to store plutonium in a locked vault. In Japan, there are three ways of solving these problems which are currently worked out. There is no doubt that the best solution is to use plutonium in fast reactors. To reach this goal, an Experimental Fast Reactor ''JOYO''has been constructed and it is waiting for criticality in very near future. A prototype fast breeder reactor ''MONJU'', which is designed for about 300 MWe, is nearing to the last stage of the design work. The start of its construction will take place in a few yesars. The domonstration fast breeder reactor will come next to ''MONJU'' and the large scale commercial use of fast breeder reactor is expected to start around 1995. To anwer the near-term need for plutonium utilization, two technologies, which are equally important to Japan, are currently developed. One is the recycle use of plutonium into LWR. This technology has long been jointly developed by research organizations and utilities. Some of fuel irradiation data are already obtained and the physics study has also been extensive. The application of this technology is expected to start about 1987. The other is to burn plutonium in an Advanced Thermal Reactor (D 2 O moderated, Boiling Water Cooled) which shows better characteristics of using plutonium. The 160 MWe ''Fugen'' is a prototype

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

Utilization of nuclear research reactors

International Nuclear Information System (INIS)

1980-01-01

Full text: Report on an IAEA interregional training course, Budapest, Hungary, 5-30 November 1979. The course was attended by 19 participants from 16 Member States. Among the 28 training courses which the International Atomic Energy Agency organized within its 1979 programme of technical assistance was the Interregional Training Course on the Utilization of Nuclear Research Reactors. This course was held at the Nuclear Training Reactor (a low-power pool-type reactor) of the Technical University, Budapest, Hungary, from 5 to 30 November 1979 and it was complemented by a one-week Study Tour to the Nuclear Research Centre in Rossendorf near Dresden, German Democratic Republic. The training course was very successful, with 19 participants attending from 16 Member States - Bangladesh, Bolivia, Czechoslovakia, Ecuador, Egypt, India, Iraq, Korean Democratic People's Republic, Morocco, Peru, Philippines, Spain, Thailand, Turkey, Vietnam and Yugoslavia. Selected invited lecturers were recruited from the USA and Finland, as well as local scientists from Hungarian institutions. During the past two decades or so, many research reactors have been put into operation around the world, and the demand for well qualified personnel to run and fully utilize these facilities has increased accordingly. Several developing countries have already acquired small- and medium-size research reactors mainly for isotope production, research in various fields, and training, while others are presently at different stages of planning and installation. Through different sources of information, such as requests to the IAEA for fellowship awards and experts, it became apparent that many research reactors and their associated facilities are not being utilized to their full potential in many of the developing countries. One reason for this is the lack of a sufficient number of trained professionals who are well acquainted with all the capabilities that a research reactor can offer, both in research and

Reactor utilization, Part 1

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1981-01-01

The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

Aspects on optimization of natural uranium fuel utilization in heavy water reactors

International Nuclear Information System (INIS)

1978-08-01

This paper is dealing with a possibility to decrease the natural uranium consumption of CANDU PHWR using the once-through cycle. This possibility is based on the utilization of slightly enriched uranium. The optimal two-zone structure of a reactor using natural uranium is found out. The optimal criterium is the maximization of the burnup (equivalent to minimization of uranium requirements) with a constraint on power density radial uniformity factor. As regards the enriched uranium, the optimal enrichment and the two-zone structure of a reactor which minimizes the natural uranium requirement with constraints on uniformity factor and maximum burnup are established. Corresponding to a maximum burnup of 16,000 MWd/t and 1% enrichment, the natural uranium requirement is found to be 10% less than that of the natural uranium reactor

Operation and utilizations of Dalat nuclear research reactor

International Nuclear Information System (INIS)

Hien, P.Z.

1988-01-01

The reconstructed Dalat nuclear research reactor was commissioned in March 1984 and up to September 1988 more than 6200 hours of operation at nominal power have been recorded. The major utilizations of the reactor include radioisotope production, activation analysis, nuclear data research and training. A brief review of the utilizations of the reactor is presented. Some aspects of reactor safety are also discussed. (author)

Operation and utilization of low power research reactor critical facility for Advanced Heavy Water Reactor (AHWR)

International Nuclear Information System (INIS)

De, S.K.; Karhadkar, C.G.

2017-01-01

An Advanced Heavy Water Reactor (AHWR) has been designed and developed for maximum power generation from thorium considering large reserves of thorium. The design envisages using 54 pin MOX cluster with different enrichment of "2"3"3U and Pu in Thoria fuel pins. Theoretical models developed to neutron transport and the geometrical details of the reactor including all reactivity devices involve approximations in modelling, resulting in uncertainties. With a view to minimize these uncertainties, a low power research reactor Critical Facility was built in which cold clean fuel can be arranged in a desired and precise geometry. Different experiments conducted in this facility greatly contribute to understand and validate the physics design parameters

The European pressurized water reactor

International Nuclear Information System (INIS)

Leny, J.C.

1993-01-01

The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

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.)

Enhancement of research reactor utilization in the developing countries

International Nuclear Information System (INIS)

Bashir, J.; Butt, N.M.

1994-06-01

As the research reactor represents a significant capital investment on the part of any institution and in addition there are recurring annual operating costs, therefore, the subject of its effective utilization has always been of interest. World wide there are about three hundred research reactors. Of these, 92 are located in the developing countries. Together, these reactors represent quite significant research potential. In the present paper, reasons of under utilization, procedures necessary to measure the productivity, ways and means of enhancing the utilization of research reactors are described. In the end, use of two research reactors at PINSTECH are described to illustrate some of the ways in which a successful utilization of a research reactor can made in the developing country. (author) 9 figs

Annual report on operation, utilization and technical development of research reactors and hot laboratory

International Nuclear Information System (INIS)

1990-09-01

This report describes the activities of the Department of Research Reactor Operation in fiscal year of 1989. It also presents some technical topics on the reactor operation and utilization in details. The Department is responsible for operation of the research reactors, JRR-2 and JRR-4, and the Hot Laboratory. The research reactor JRR-3 was reconstructed to enhance the performance for utilization. The first criticality was achieved on March 22, 1989, and it subsequently went into operation. In connection with the reactor operation, the various research and development activities in the area of fuel management, water chemistry, radiation monitoring and material irradiation have been made. In the Hot Laboratory, post-irradiation examinations of fuels and materials have been carried out along with the development of related techniques. (author)

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

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.)

Experience in utilizing research reactors in Yugoslavia

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J.; Raisic, N. [Boris Kidric Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia); Copic, M.; Gabrovsek, Z. [Jozef Stefan Institute Ljubljana (Yugoslavia)

1972-07-01

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

Experience in utilizing research reactors in Yugoslavia

International Nuclear Information System (INIS)

Pop-Jordanov, J.; Raisic, N.; Copic, M.; Gabrovsek, Z.

1972-01-01

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

User oriented design features of Korea Multi-purpose Research Reactor and its utilization plan

International Nuclear Information System (INIS)

Kim, Byungkoo; Jun, Byungjin

1994-01-01

Construction of a 30 MW class Korea Multi-purpose Research Reactor (KMRR) is near its completion and expected to reach initial criticality by the end of 1994 at KAERI Taejon site. As Korea will become one of developed countries during the lifetime of this reactor and many worldwide high performance research reactors of the first generation are reaching end of life, it is believed that KMRR will meet the increasing domestic needs to utilize high performance research reactor and its worldwide role will be important as well. In reactor design, effort has been focused on optimization which can satisfy various future utilization demands as much as possible with enhanced safety as a basic requirement. Light water cooled and heavy water reflected compact core using low enriched and high uranium loaded fuel, computer controlled operation, etc. are expected to provide truly multi-purpose user environments with stable high quality neutron flux. High level experimental facilities and equipment for reactor fuel and material test, various studies using neutron beam, radioisotope production, semiconductor doping, neutron activation analysis, etc., will be completed in parallel with the reactor or gradually depending on users' needs. When KMRR becomes fully operational, it will not only serve the domestic users but also be a valuable tool for a worldwide research community using a research reactor

Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

International Nuclear Information System (INIS)

Shiba, T.; Fallot, M.

2015-01-01

To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

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

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.

Use of reactor plants of enhanced safety for sea water desalination, industrial and district heating

International Nuclear Information System (INIS)

Panov, Yu.; Polunichev, V.; Zverev, K.

1997-01-01

Russian designers have developed and can deliver nuclear complexes to provide sea water desalination, industrial and district heating. This paper provides an overview of these designs utilizing the ABV, KLT-40 and ATETS-80 reactor plants of enhanced safety. The most advanced nuclear powered water desalination project is the APVS-80. This design consists of a special ship equipped with the distillation desalination plant powered at a level of 160 MW(th) utilizing the type KLT-40 reactor plant. More than 20 years of experience with water desalination and reactor plants has been achieved in Aktau and Russian nuclear ships without radioactive contamination of desalinated water. Design is also proceeding on a two structure complex consisting of a floating nuclear power station and a reverse osmosis desalination plant. This new technology for sea water desalination provides the opportunity to considerably reduce the specific consumption of power for the desalination of sea water. The ABV reactor is utilized in the ''Volnolom'' type floating nuclear power stations. This design also features a desalinator ship which provides sea water desalination by the reverse osmosis process. The ATETS-80 is a nuclear two-reactor cogeneration complex which incorporates the integral vessel-type PWR which can be used in the production of electricity, steam, hot and desalinated water. (author). 9 figs

Design-development and operation of the Experimental Boiling-Water Reactor (EBWR) facility, 1955--1967

International Nuclear Information System (INIS)

Boing, L.E.; Wimunc, E.A.; Whittington, G.A.

1990-11-01

The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission -- demonstrating the practicality of the direct-boiling concept -- and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light- water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program. 13 refs., 12 figs

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

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.)

Country report: utilization of MINT's research reactor

International Nuclear Information System (INIS)

Ab Khalic b Hj Wood; Adnan b Bukhari; Wee Boon Siong

2004-01-01

MINT has only one research reactor, i.e. TRIGA MKII reactor, equipped with various neutron irradiation facilities such as rotary rack and rabbit system. Apart from counting facilities for NAA work, other facilities available for the respective studies include facilities for neutron radiography and SANS. At Present most of reactor operation time has been utilized for samples irradiation related to the NAA application. Majority of the samples are from MINT analytical chemistry laboratory where the present authors work, and the rest of the samples are from local universities. They provide analytical chemistry services for other government departments as well as private companies. In order to improve the reactor utilization, the management of MINT has formed Reactor Interest Group (RIG) at the national level in 2002, which embraces members from various institutions in this country. To support the RIG activities, MINT provides seed funding to finance various activities for the reactor utilization, which include financing project to make use of SANS, neutron radiography and radioisotopes production (mainly for tracer studies carried out by MINT's tracer group) facilities, and funding for basic study in BNCT. (author)

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

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.)

RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1984

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1984-12-01

During the 1984 the reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981. Operation of the primary cooling system was changed in order to avoid appearance of the previously noticed aluminium oxyhydrate on the surface of the fuel element claddings. The new cooling regime enabled more efficient heavy water purification. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks are planned: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. Financing of the planned activities will be partly covered by the IAEA. this Part I of the report includes 8 Annexes describing in detail the reactor operation, and 6 special papers dealing with the problems of reactor operation and utilization

Research on Reduced-Moderation Water Reactor (RMWR)

International Nuclear Information System (INIS)

Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro

1999-11-01

The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from 238 U to 239 Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPCO) in 1998. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR with high void fraction and super-flat core, a long operation cycle BWR using void channels, a high conversion BWR without blankets, a high conversion PWR using heavy water as a coolant, and a PWR for plutonium multi-recycle using seed-blanket type fuel assemblies. The present report summarizes the objectives, domestic and international trends, principles and characteristics, core conceptual designs and future R and D plans of the RMWR. (J.P.N.)

Research on Reduced-Moderation Water Reactor (RMWR)

Energy Technology Data Exchange (ETDEWEB)

Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

1999-11-01

The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from {sup 238}U to {sup 239}Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPCO) in 1998. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR with high void fraction and super-flat core, a long operation cycle BWR using void channels, a high conversion BWR without blankets, a high conversion PWR using heavy water as a coolant, and a PWR for plutonium multi-recycle using seed-blanket type fuel assemblies. The present report summarizes the objectives, domestic and international trends, principles and characteristics, core conceptual designs and future R and D plans of the RMWR. (J.P.N.)

Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

International Nuclear Information System (INIS)

Snead, Lance Lewis; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Robb, Kevin R.; Snead, Mary A.

2015-01-01

This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

Energy Technology Data Exchange (ETDEWEB)

Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-04-01

This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

International Nuclear Information System (INIS)

Johansen, R.

2011-01-01

Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

The thorium fuel cycle in water-moderated reactor systems

International Nuclear Information System (INIS)

Critoph, E.

1977-05-01

Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An 'inventory' of uranium of between 1 and 2 Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium) is some two decades

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

LIGHT WATER MODERATED NEUTRONIC REACTOR

Science.gov (United States)

Christy, R.F.; Weinberg, A.M.

1957-09-17

A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

Estabilishing requirements for the next generation of pressurized water reactors--reducing the uncertainty

International Nuclear Information System (INIS)

Chernock, W.P.; Corcoran, W.R.; Rasin, W.H.; Stahlkopf, K.E.

1987-01-01

The Electric Power Research Institute is managing a major effort to establish requirements for the next generation of U.S. light water reactors. This effort is the vital first step in preserving the viability of the nuclear option to contribute to meet U.S. national electric power capacity needs in the next century. Combustion Engineering, Inc. and Duke Power Company formed a team to participate in the EPRI program which is guided by a Utility Steering committee consisting of experienced utility technical executives. A major thrust of the program is to reduce the uncertainties which would be faced by the utility executives in choosing the nuclear option. The uncertainties to be reduced include those related to safety, economic, operational, and regulatory aspects of advanced light water reactors. This paper overviews the Requirements Document program as it relates to the U.S. Advanced Light Water Reactor (ALWR) effort in reducing these uncertainties and reports the status of efforts to establish requirements for the next generation of pressurized water reactors. It concentrates on progress made in reducing the uncertainties which would deter selection of the nuclear option for contributing to U.S. national electric power capacity needs in the next century and updates previous reports in the same area. (author)

General outline of the operation and utilization of the BR2 reactor

International Nuclear Information System (INIS)

Baugnet, J.M.; Leonard, F.; Gandolfo, J.M.; Lenders, H.

1978-01-01

The BR2 reactor is a high-flux material testing reactor of the thermal heterogeneous type. The fuel is 93% 235 U enriched uranium in the form of plates clad in aluminium. The moderator consists of beryllium and light water, the water being pressurized (12.5kg/cm 2 )and acting also as coolant. The pressure vessel is of aluminium, and is placed in a pool of demineralized water. One should stress the following main features of the design: the experimental channels are skew, the tube bundle presenting the form of a hyperboloid of revolution (see figure 1)-this gives easy access at the top and bottom reactor covers allowing complex instrumented devices, while maintaining a very high neutron flux at the core; great flexibilty of utilization, due to the fact that it is possible to adapt the core configuration to the experimental loading as the fissile charge can be centred on different experimental channels; although BR2 is a thermal reactor, it is possible to achieve neutron spectra very similar to those obtained in a fast reactor, either by the use of absorbing screens or by the use of fissile material within the experimental device; five 200mm diameter channels are available for loading large experimental irradiation devices, as in-pile sodium, gas or water loops. (author)

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

Modified fuel assembly design for pressurized water reactors with improved fuel utilization

International Nuclear Information System (INIS)

Galperin, A.; Ronen, Y.

1983-01-01

A method for reactivity control through variation of the moderator content in the reactor core was proposed. The main idea is to adjust the amount of water in the core from a low value at beginning of cycle to a high value at end of cycle, so as to compensate for fissile material burnup and buildup of fission products. The possible implementation of this idea may be carried out by introducing a number of hollow tubes into the fuel assembly between the fuel rods. Then variation of the moderator content in the core may be managed through a change of the water level in these tubes. cated a potential savings in the fuel cycle requirements and costs. Preliminary steady-state thermal-hydraulic calculations indicate the possibility of implementing the proposed method in the existing pressurized water reactor plants. Feasibility of the proposed design may be finally established after rigorous thermal hydraulics as well as safety analysis calculations. Furthermore, there is need to elaborate the mechanical design of the pressure vessel internals together with cost benefit analysis

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.)

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.)

Feasibility study of high temperature reactor utilization in Czech Republic after 2025

Energy Technology Data Exchange (ETDEWEB)

Losa, Evžen, E-mail: evzen.losa@fjfi.cvut.cz [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (Czech Republic); Heřmanský, Bedřich; Kobylka, Dušan; Rataj, Jan; Sklenka, Ľubomír [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (Czech Republic); Souček, Václav; Kohout, Petr [AZIN CZ, s.r.o., Hanusova 3, 140 00 Praha 4 (Czech Republic)

2014-05-01

High temperature reactors (HTRs) were examined as an option to intended future broadening of the nuclear energy production in Czech Republic. The known qualities as the inherent safety, high thermal utilization and non-electrical applications have been assessed in years 2009–2011 during the survey funded by Czech Ministry of Industry and Trade. The survey of high temperature reactors with spherical fuel was initiated by reason of mature state of the art of this technology type in South Africa and in China, where in both countries pilot plants were planned. Unfortunately, the global financial crisis caused the decision of stopping the governmental support in South African programme was made. In China, however, the development still continues. Czech Republic has almost 60 years nuclear research history and the knowledge of operation of gas cooled and heavy water moderated reactor has been gained in the past. Nevertheless, the design of light water reactors was more developed in former Soviet Union, which provided Czech scientists by initial knowledge base; hence the research has been reoriented to this technology. But, the demands on future nuclear reactors application are still growing and the same or even higher living standard of next generations have to be taken into consideration. Therefore the systems, which can produce more energy and less waste, are getting into foreground of interest of Czech decision makers. The high temperature reactor technology seems to be the successful representative of the GEN IV reactor types, which will be operated commercially in the near future. The broad spectrum of utilization enables this system to be an option after 2030, when the electricity demand is planned to be covered from about 50% by nuclear in our country.

Feasibility study of high temperature reactor utilization in Czech Republic after 2025

International Nuclear Information System (INIS)

Losa, Evžen; Heřmanský, Bedřich; Kobylka, Dušan; Rataj, Jan; Sklenka, Ľubomír; Souček, Václav; Kohout, Petr

2014-01-01

High temperature reactors (HTRs) were examined as an option to intended future broadening of the nuclear energy production in Czech Republic. The known qualities as the inherent safety, high thermal utilization and non-electrical applications have been assessed in years 2009–2011 during the survey funded by Czech Ministry of Industry and Trade. The survey of high temperature reactors with spherical fuel was initiated by reason of mature state of the art of this technology type in South Africa and in China, where in both countries pilot plants were planned. Unfortunately, the global financial crisis caused the decision of stopping the governmental support in South African programme was made. In China, however, the development still continues. Czech Republic has almost 60 years nuclear research history and the knowledge of operation of gas cooled and heavy water moderated reactor has been gained in the past. Nevertheless, the design of light water reactors was more developed in former Soviet Union, which provided Czech scientists by initial knowledge base; hence the research has been reoriented to this technology. But, the demands on future nuclear reactors application are still growing and the same or even higher living standard of next generations have to be taken into consideration. Therefore the systems, which can produce more energy and less waste, are getting into foreground of interest of Czech decision makers. The high temperature reactor technology seems to be the successful representative of the GEN IV reactor types, which will be operated commercially in the near future. The broad spectrum of utilization enables this system to be an option after 2030, when the electricity demand is planned to be covered from about 50% by nuclear in our country

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)

Factors affecting nuclear research reactor utilization across countries

International Nuclear Information System (INIS)

Hien, P.D.

2000-01-01

In view of the worldwide declining trend of research reactor utilization and the fact that many reactors in developing countries are under-utilised, a question naturally arises as to whether the investment in a research reactor is justifiable. Statistical analyses were applied to reveal relationships between the status of reactor utilization and socio-economic conditions among countries, that may provide a guidance for reactor planning and cost benefit assessment. The reactor power has significant regression relationships with size indicators such as GNP, electricity consumption and R and D expenditure. Concerning the effectiveness of investment in research reactors, the number of reactor operation days per year only weakly correlates with electricity consumption and R and D expenditure, implying that there are controlling factors specific of each group of countries. In the case of less developed countries, the low customer demands on reactor operation may be associated with the failure in achieving quality assurance for the reactor products and services, inadequate investment in the infrastructure for reactor exploitation, the shortage of R and D funding and well trained manpower and the lack of measures to get the scientific community involved in the application of nuclear techniques. (author)

Decommissioning and re-utilization of the Musashi Reactor

International Nuclear Information System (INIS)

Tomio Tanzawa; Nobukazu Iijima; Norikazu Horiuchi; Tadashi Yoshida; Tetsuo Matsumoto; Naoto Hagura; Ryouhei Kamiya

2008-01-01

The Musashi Institute of Technology Research Reactor (the Musashi Reactor) is a TRIGA-? with maximum thermal power of 100 kW. The decommissioning was decided in May, 2003. The reactor facility is now under decommissioning. The phased decommissioning was selected. Phase 1 consists of permanent shutdown of the reactor and stopping the operational functions, and transportation of the spent nuclear fuels. After completion of the transportation, the reactor facility is characterized as the storage of low level radioactive materials. This is phase 2. Activities of phase 1 were completed and the facility is now under phase 2. Activities of phase 3 consist of dismantling the reactor tank and the shielding, and delivering the radioactive waste to a waste disposal facility. The phase 3 will be started on condition that the undertaking of the waste disposal for research reactors will be established. On the other hand, re-utilization of the facility has being studied, and 'realistic' reactor simulator was turned out by utilizing the reactor installations such as control rod drive and operation console. (authors)

Utility industry evaluation of the Sodium Advanced Fast Reactor

International Nuclear Information System (INIS)

Burstein, S.; DelGeorge, L.O.; Tramm, T.R.; Gibbons, J.P.; High, M.D.; Neils, G.H.; Pilmer, D.F.; Tomonto, J.R.; Wells, J.T.

1990-02-01

A team of utility industry representatives evaluated the Sodium Advanced Fast Reactor plant design, a current liquid metal reactor design created by an industrial team led by Rockwell International under Department of Energy sponsorship. The utility industry team concluded that the plant design offers several attractive characteristics, especially in the safety arena, as well as preserving the traditional attraction of liquid metal reactors, very high fuel utilization. Specific comments and recommendations are provided as a contribution towards improving an already attractive plant design. 18 refs

Utilization of MVP for research on fast reactor

International Nuclear Information System (INIS)

Yokoyama, Kenji

2001-01-01

Utilization of the continuous energy Monte-Carlo code, MVP, for research on fast reactor in Power Reactor and Nuclear Fuel Development Corporation(PNC) is described. In this report, three types of utilization are reviewed; (1) a comparison of the eigenvalues calculated by MVP with the results by the deterministic methods, (2) an improvement of U-238 reaction rate evaluation in JUPITER experimental Analysis and (3) an evaluation of heterogeneity effects for Am reaction rates of the moderated subassemblies. Since the results of MVP can be used as references, MVP is very useful code in research on fast reactor. It is one of indispensable tools in order to verify the models in the deterministic methods. Furthermore, it can be used so as to investigate the new concept reactors, such as a reactor aiming to transmute minor actinides(MA). On the other hand, a problem of the variance reduction remains. Especially, a small reactivity cannot be estimated by MVP because of large variances. The development of a Monte-Carlo method for a small reactivity calculation will promote the utilization of MVP for research on fast reactor. (author)

Survey of thorium utilization in power reactor systems

International Nuclear Information System (INIS)

Schwartz, M.H.; Schleifer, P.; Dahlberg, R.C.

1976-01-01

It is clear that thorium-fueled thermal power reactor systems based on current technology can play a vital role in serving present and long-term energy needs. Advanced thorium converters and thermal breeders can provide an expanded resource base from which the world's growing energy demands can be met. Utilization of a symbiotic system of fast breeders and thorium-fueled thermal reactors can be particularly effective in providing low cost power while conserving uranium resources. Breeder reactors are characterized by high capital costs and very low fuel costs since they produce more fuel than they consume. This excess fuel can be used to fuel thermal converter reactors whose capital costs are low. This symbiosis is optimized when 233 U is bred in the fast breeders and then used to fuel high-conversion-ratio thermal converter reactors operating on the thorium-uranium fuel cycle. The thorium-cycle HTGR, after undergoing more than fifteen years of development in both the United States and Europe, provides for the optimum utilization of our limited uranium resources. Other thermal reactor systems, previously operating on the uranium cycle, also show potential in their capability to utilize the thorium cycle effectively

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.

Progress in design study on reduced-moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Shirakawa, Toshihisa; Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takeda, Renzo [Hitachi Ltd., Tokyo (Japan); Yokoyama, Tsugio [Toshiba Corp., Kawasaki, Kanagawa (Japan); Hibi, Koki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Wada, Shigeyuki [Japan Atomic Power Co., Tokyo (Japan)

2000-06-01

The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from {sup 238}U to {sup 239}Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPC) in 1998, under technical cooperation with three Japanese reactor vendors. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight-lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR type core with high void fraction and super-flat core, a long operation cycle BWR type core using void tube assembly, a high conversion BWR type core without blankets, a high conversion PWR type core using heavy water as a coolant, and a PWR type core for plutonium multi-recycle using seed-blanket type fuel assemblies. Detailed feasibility studies for the RMWR have been continued on core design study. The present report summarizes the recent progress in the design study for the RMWR. (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

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

International Nuclear Information System (INIS)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

2011-01-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Strategies for plutonium recycle in a system of pressurized water reactors

International Nuclear Information System (INIS)

Leaver, D.E.W.

1976-01-01

A methodology is developed to allow a utility fuel manager to determine economic strategies for recycling plutonium in a system of light water reactors. One possible plutonium recycle strategy would be self-generated recycle, in which plutonium discharged from a reactor is recycled back to that same reactor as soon as possible. Another possible strategy is to recycle all the plutonium discharged from several reactors into one reactor. Such a strategy might be advantageous if the reactor receiving the plutonium were of a type that utilized plutonium more effectively than other reactors in the system. There are several considerations which affect the economics of recycling a batch of plutonium to one reactor or cycle vs. another, or which would favor a special recycling strategy. Among these are cycle energy, length of time that plutonium is stored prior to recycle, and isotopes of the plutonium. The methodology developed is used to quantitatively illustrate the effect on recycle strategy of these parameters. The problem of choosing the plutonium recycle strategy which results in the minimum fuel cost is formulated as a mathematical programming problem. The objective function for this problem is the total discounted fuel cost for the reactor system over a specified planning period. The savings of an optimal recycle strategy over self-generated recycle would be typically one million dollars per year for a utility with several large PWRs

Nuclear safety in light water reactors severe accident phenomenology

CERN Document Server

Sehgal, Bal Raj

2011-01-01

This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

Automated ultrasonic examination of light water reactor systems

International Nuclear Information System (INIS)

Walter, J.H.

1975-01-01

An automated ultrasonic examination system has been developed to meet the pre- and inservice inspection requirements of light water reactors. This system features remotely-controlled travelling instrument carriers, computerized collection and storage or inspection data in a manner providing real time comparison against code standards, and computer control over the positioning of the instrument carriers to provide precise location data. The system is currently being utilized in the field for a variety of reactor inspections. The principal features of the system and the recent inspection experience are discussed. (author)

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.)

NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER provides 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

Reactor core of light water-cooled reactor

International Nuclear Information System (INIS)

Miwa, Jun-ichi; Aoyama, Motoo; Mochida, Takaaki.

1996-01-01

In a reactor core of a light water cooled reactor, the center of the fuel rods or moderating rods situated at the outermost circumference among control rods or moderating rods are connected to divide a lattice region into an inner fuel region and an outer moderator region. In this case, the area ratio of the moderating region to the fuel region is determined to greater than 0.81 for every cross section of the fuel region. The moderating region at the outer side is increased relative to the fuel rod region at the inner side while keeping the lattice pitch of the fuel assembly constant, thereby suppressing the increase of an absolute value of a void reactivity coefficient which tends to be caused when using MOX fuels as a fuel material, by utilizing neutron moderation due to a large quantity of coolants at the outer side of the fuel region. The void reactivity coefficient can be made substantially equal with that of uranium fuel assembly without greatly reducing a plutonium loading amount or without greatly increasing linear power density. (N.H.)

Research Reactors Types and Utilization

International Nuclear Information System (INIS)

Abdelrazek, I.D.

2008-01-01

A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235 U or 239 Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

Energy Technology Data Exchange (ETDEWEB)

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-04-01

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

International Nuclear Information System (INIS)

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-01-01

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

Pu utilization in fast-breeder and in light-water reactors in Italy

International Nuclear Information System (INIS)

Mangiagalli, D.; Cicognani, F.; Pistella, F.; Testa, G.; Villani, A.; Ariemma, A.; Castelli, G.F.; Linari, A.; Paoletti Gualandi, M.; Musso, B.

1977-01-01

The paper illustrates the most important activities carried out in Italy for the development of fast breeder reactors and its fuel as well as for plutonium recycle in light water reactors. The Italian strategy is based, on one hand, on the short-term commercialization of fast breeder reactors, and on the other, on the adoption of the technology of the Phenix prototype whose further development will be ensured by the joint Italian and French efforts as insured by the important agreements signed by CNEN, NIRA (Nucleare Italiana Reattori Avanzati) and Italian manufacturing industries with CEA and the main French industries. The paper also includes the main results of the ENEL Demonstration Program on Pu prototypes introduced in the Garigliano BWR in 1968 and 1970, and of the destructive and non-destructive analyses on said fuel, as well as of the analyses carried out by CNEN on prototypical fuel fabricated by CNEN and irradiated in various reactors. Furthermore, the paper deals with design and licensing aspects of the 46 Pu-island assembly reload introduced in the Garigliano reactors in 1975 and of a batch of 8 all Pu assemblies loaded in the Trino Vercellese PWR in 1976. Subsequently, the experimental activities planned for the near future both on high burn-up prototypes and on industrial fuel after one cycle of operation are examined [fr

Reactor calculations for improving utilization of TRIGA reactor

International Nuclear Information System (INIS)

Ravnik, M.

1986-01-01

A brief review of our work on reactor calculations of 250 kW TRIGA with mixed core (standard + FLIP fuel) will be presented. The following aspects will be treated: - development of computer programs; - optimization of in-core fuel management with respect to fuel costs and irradiation channels utilization. TRIGAP programme package will be presented as an example of computer programs. It is based on 2-group 1-D diffusion approximation and besides calculations offers possibilities for operational data logging and fuel inventory book-keeping as well. It is developed primarily for the research reactor operators as a tool for analysing reactor operation and fuel management. For this reason it is arranged for a small (PC) computer. Second part will be devoted to reactor physics properties of the mixed cores. Results of depletion calculations will be presented together with measured data to confirm some general guidelines for optimal mixed core fuel management. As the results are obtained using TRIGAP program package results can be also considered as an illustration and qualification for its application. (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)

BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

International Nuclear Information System (INIS)

1987-11-01

This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

Research reactor utilization in chemistry programmes

International Nuclear Information System (INIS)

Bautista, E.

1983-01-01

The establishment and roles of the Philippines Atomic Energy Commission in promoting and regulating the use of atomic energy are explained. The research reactor, PRR-1 is being converted to TRIGA to meet the increasing demands of high-flux. The activities of PAEC in chemistry research programs utilizing reactor are discussed in detail. The current and future plans of Research and Development programs are also included. (A.J.)

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

Microwave reactor for utilizing waste materials

Directory of Open Access Journals (Sweden)

M. Pigiel

2010-01-01

Full Text Available The paper presents a designed and manufactured, semi-industrial microwave reactor for thermal utilization of asbestos-bearing wastes. Presented are also semi-industrial tests of utilizing such wastes. It was found that microwave heating can be applied for utilizing asbestos with use of suitable wetting agents. The wetting agents should ensure continuous heating process above 600 °C, as well as uniform heat distribution in the whole volume of the utilized material. Analysis of the neutralization process indicates a possibility of presenting specific, efficient and effective process parameters of utilizing some asbestos-bearing industrial wastes.

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)

TMI-2 in-vessel hydraulic systems utilize high water and high boron content fluids

International Nuclear Information System (INIS)

Baston, V.F.; Hofstetter, K.J.; Hofman, L.A.; Gallagher, R.E.

1987-01-01

Choice of a hydraulic fluid for use in the Three Mile Island Unit 2 (TMI-2) reactor vessel defueling equipment required consideration of the following constraints for the hydraulic fluid given an accidental spill into the reactor coolant system (RCS). The TMI-2 RCS hydraulic fluid utilized in the hydraulic operations utilized a solution composition of 95 wt% water and 5 wt% of the above base fluid. The TMI-2 hydraulic system utilizes pressures up to 3500 psi. The selected hydraulic fluid has been in use since December 1986 with minimal operational difficulties

Passive safety features in current and future water cooled reactors

International Nuclear Information System (INIS)

1990-11-01

Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

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

Transactions of the Twentieth Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1992-10-01

This report contains summaries of papers on reactor safety research to be presented at the 20th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 21--23, 1992. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included

Utilization of thorium in thermal reactors

International Nuclear Information System (INIS)

Srinivasan, K.R.; Nakra, A.N.

1978-01-01

Large deposits of thorium are found in India. 233 U produced by neutron capture in 232 Th is a more valuable fuel for thermal reactors than the plutonium that results from capture in 238 U. These two facts are the main reasons for the interest in utilizing thorium in power reactors. But natural thorium does not contain any fissile material and its capture cross section is nearly two and a half times that of 238 U. These have made the fuelling cost high. However, in certain conditions and certain types of reactors the costs are comparable with those using uranium fuel. The relative cost effectiveness of different fuels is discussed. Apart from long term interest, the short term interest of using thorium fuel in RAPP type reactors is also briefly described. Finally the reactor physics experiments using thorium fuel and their comparison with calculations are presented. (author)

Achievement and development of neutron beam utilization in research reactors

International Nuclear Information System (INIS)

Isshiki, Masahiko

1996-01-01

Especially regarding the neutron beam experiment in Japan, the basic research has been developed by utilizing the JRR-2 of Japan Atomic Energy Research Institute and the KUR of Kyoto University over long years. Now, the JRR-3M of JAERI was revived as a high performance, general purpose reactor, and bears important roles as the neutron beam experiment center in Japan. Thanks to one of the most powerful reactor neutron sources in the world and the cold neutron source, the environment of research was greatly improved, and the excellent results of researches began to be reported. The discovery of neutrons by Chadwick and the history of the related researches are described. As neutron sources, radioisotopes, accelerators and nuclear reactors are properly used corresponding to purposes. As the utilization of research reactors for neutron sources, the utilization for irradiation and neutron beam experiment are carried out. The outline of the research reactor JRR-3M is explained. The state of utilization in neutron scattering experiment, neutron radiography, prompt γ-ray analysis and the medical irradiation of neutrons is reported. (K.I.)

Concept of innovative water reactor for flexible fuel cycle (FLWR)

International Nuclear Information System (INIS)

Iwamura, T.; Uchikawa, S.; Okubo, T.; Kugo, T.; Akie, H.; Nakatsuka, T.

2005-01-01

In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming LWR-Mixed Oxide (MOX) technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the fuel cycle circumstances during the reactor operation period around 60 years. At present, since the fuel cycle for the plutonium multiple recycling with MOX fuel reprocessing has not been realized yet, reprocessed plutonium from the LWR spent fuel is to be utilized in LWR-MOX. After this stage, the first part of FLWR, i.e. the high conversion type, can be introduced as a replacement of LWR or LWR-MOX. Since the plutonium inventory of FLWR is much larger, the number of the reactor with MOX fuel will be significantly reduced compared to the LWR-MOX utilization. The size of the fuel assembly for the first part is the same as in the RMWR concept, i.e. the hexagonal fuel assembly with the inner face-to-face distance of about 200 mm. Fuel rods are arranged in the triangular lattice with a relatively wide gap size around 3 mm between rods, and the effective MOX length is less than 1.5 m without using the blanket. When

Guidebook on destructive examination of water reactor fuel

International Nuclear Information System (INIS)

1997-01-01

As a result of common efforts of fuel vendors, utilities and research institutes the average burnup pf design batch fuels was increased for both PWRs and BWRs and the fuel failure rate has been reduced. The previously published Guidebook on Non-Destructive Examination of Water Reactor Fuel recommended that more detailed destructive techniques are required for complete understanding of fuel performance. On the basis of contributions of the 14 participants in the ED-WARF-II CRP and proceedings of IAEA Technical Committee on Recent Developments in Post-irradiation Examination Techniques for Water Reactor Fuel this guidebook was compiled. It gives a complete survey of destructive techniques available to date worldwide. The following examination techniques are described in detailed including major principles of equipment design: microstructural studies; elemental analysis; isotopic analysis; measurement of physical properties; measurement of mechanical properties. Besides the examination techniques, methods for refabrication of experimental rods from high burnup power reactor rods as well as methods for verification of non-destructive techniques by using destructive techniques is included

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

Operation and utilization of Indonesia Research Reactors

International Nuclear Information System (INIS)

Kuntoro, Iman; Sujalmo, Saiful; Tarigan, Alim

2004-01-01

For supporting the R and D in nuclear science and technology and its application, BATAN own and operate three research reactors namely, TRIGA-2000, KARTINI and RSG-GAS having thermal power of 2 MW, 100 kW and 30 MW respectively. The main features, operation and utilization progress of the reactors are described in this report. (author)

Safety in the utilization and modification of research reactors

International Nuclear Information System (INIS)

1994-01-01

This Safety Guide presents guidelines, approved by international consensus, for the safe utilization and modification of research reactors. While the Guide is most applicable to existing reactors, it is also recommended for use by organizations planning to put a new reactor into operation. 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

Thorium utilization in power reactors

International Nuclear Information System (INIS)

Saraceno; Marcos.

1978-10-01

In this work the recent (prior to Aug, 1976) literature on thorium utilization is reviewed briefly and the available information is updated. After reviewing the nuclear properties relevant to the thorium fuel cycle we describe briefly the reactor systems that have been proposed using thorium as a fertile material. (author) [es

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

Operation, maintenance and utilization of the RA reactor, Annual report 1978

International Nuclear Information System (INIS)

Milosevic, M.

1978-12-01

It has been planned for 1978 that the RA reactor would be operated for 158 dana at nominal power of 6.5 MW meaning production of 24 648 MWh. The plan was fulfilled since 24 652 MWh was produces. Reactor operation for 158 days is relevant to reactor operation for 200 days in the period before 1975. The reason is increased neutron flux achieved due to improved fuel management and the characteristics of the new 80% enriched fuel. At the end of 1978 the reactor core contained 45% of 80% enriched fuel elements. Increase of neutron flux has shortened the typical time needed for irradiation of the most important samples for isotope production. This significant success in reactor operation is at the same time an obligation for increasing its utilization. Some new trends proposed for increasing reactor utilization capacities were presented at the Conference on utilization of research nuclear reactors in Yugoslavia held in May 1978 [sr

The utility of different reactor types for the research

International Nuclear Information System (INIS)

Stiennon, G.

1983-01-01

The report presents a general view of the use of the different belgian research reactor i.e. venus reactor, BR-1 reactor, BR-2 reactor and BR-3 reactor. Particular attention is given to the programmes which is in the interest of international collaboration. In order to reach an efficient utilization of such reactors they require a specialized personnel groups to deal with the irradiation devices and radioactive materials and post irradiation examinations, creating a complete material testing station. (A.J.)

Boiling water reactor modeling capabilities of MMS-02

International Nuclear Information System (INIS)

May, R.S.; Abdollahian, D.A.; Elias, E.; Shak, D.P.

1987-01-01

During the development period for the Modular Modeling System (MMS) library modules, the Boiling Water Reactor (BWR) has been the last major component to be addressed. The BWRX module includes models of the reactor core, reactor vessel, and recirculation loop. A pre-release version was made available for utility use in September 1983. Since that time a number of changes have been incorporated in BWRX to (1) improve running time for most transient events of interest, (2) extend its capability to include certain events of interest in reactor safety analysis, and (3) incorporate a variety of improvements to the module interfaces and user input formats. The purposes of this study were to briefly review the module structure and physical models, to point the differences between the MMS-02 BWRX module and the BWRX version previously available in the TESTREV1 library, to provide guidelines for choosing among the various user options, and to present some representative results

Analysis of neutronics benchmarks for the utilization of mixed oxide fuel in light water reactor using DRAGON code

International Nuclear Information System (INIS)

Nithyadevi, Rajan; Thilagam, L.; Karthikeyan, R.; Pal, Usha

2016-01-01

Highlights: • Use of advanced computational code – DRAGON-5 using advanced self shielding model USS. • Testing the capability of DRAGON-5 code for the analysis of light water reactor system. • Wide variety of fuels LEU, MOX and spent fuel have been analyzed. • Parameters such as k ∞ , one, few and multi-group macroscopic cross-sections and fluxes were calculated. • Suitability of deterministic methodology employed in DRAGON-5 code is demonstrated for LWR. - Abstract: Advances in reactor physics have led to the development of new computational technologies and upgraded cross-section libraries so as to produce an accurate approximation to the true solution for the problem. Thus it is necessary to revisit the benchmark problems with the advanced computational code system and upgraded cross-section libraries to see how far they are in agreement with the earlier reported values. Present study is one such analysis with the DRAGON code employing advanced self shielding models like USS and 172 energy group ‘JEFF3.1’ cross-section library in DRAGLIB format. Although DRAGON code has already demonstrated its capability for heavy water moderator systems, it is now tested for light water reactor (LWR) and fast reactor systems. As a part of validation of DRAGON for LWR, a VVER computational benchmark titled “Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel-Volume 3” submitted by the Russian Federation has been taken up. Presently, pincell and assembly calculations are carried out considering variation in fuel temperature (both fresh and spent), moderator temperatures and boron content in the moderator. Various parameters such as infinite neutron multiplication (k ∞ ) factor, one group integrated flux, few group homogenized cross-sections (absorption, nu-fission) and reaction rates (absorption, nu-fission) of individual isotopic nuclides are calculated for different reactor states. Comparisons of results are made with the reported Monte Carlo

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

Mox fuel utilization in ATR

OpenAIRE

下村 和生; 川太 徳夫

1987-01-01

ATR, a heavy-water moderated boiling-light-water cooled reactor developed in Japan, is a unique reactor with out-standing flexibility regarding nuclear fuel utilization, because it has superior properties concerning the utilization of plutonium, recovered uranium and depleted uranium. The development of this type of reactor is expected to contribute both to the stable supply of energy and to the establishment of plutonium utilization in Japan. Much effort has been and will be made on the deve...

Safety aspects of water chemistry in light water reactors

International Nuclear Information System (INIS)

1988-12-01

The goals of the water chemistry control programmes are to maximize operational safety and the availability and operating life of primary system components, to maximize fuel integrity, and to control radiation buildup. To achieve these goals an effective corporate policy should be developed and implemented. Essential management responsibilities are: Recognizing of the long-term benefits of avoiding or minimizing: a) system corrosion; b) fuel failure; and c) radiation buildup. The following control or diagnostic parameters are suitable performance indicators: for PWR primary coolant circuits: pH of reactor water (by operating temperature); Concentration of chlorides in reactor water; Hydrogen (or oxygen) in reactor water. For PWR secondary coolant circuits: pH in feedwater; Cation productivity in steam generator blowdown; Iron concentration in feedwater; Oxygen concentration in condensate. And BWR coolant circuits: Conductivity of reactor water; Concentration of chlorides in reactor water; Iron concentration in feedwater; Copper concentration in feedwater. The present document represents a review of the developments in some Member States on how to implement a reasonable water chemistry programme and how to assess its effectiveness through numerical indicators. 12 figs, 20 tabs

Reactor waste heat utilization and district heating reactors. Nuclear district heating in Sweden - Regional reject heat utilization schemes and small heat-only reactors

International Nuclear Information System (INIS)

Hannerz, K.; Larsson, Y.; Margen, P.

1977-01-01

A brief review is given of the current status of district heating in Sweden. In future, district heating schemes will become increasingly interesting as a means of utilizing heat from nuclear reactors. Present recommendations in Sweden are that large reactors should not be located closer than about 20 km from large population centres. Reject heat from such reactors is cheap at source. To minimize the cost of long distance hot water transmission large heat rates must be transmitted. Only areas with large populations can meet this requirement. The three areas of main interest are Malmoe/Lund/Helsingborg housing close to 0.5 million; Greater Stockholm housing 1 to 1.5 million and Greater Gothenburg housing about 0.5 million people. There is an active proposal that the Malmoe/Lund/Helsingborg region would be served by a third nuclear unit at Barsebaeck, located about 20 km from Malmoe/Lund and supplying 950 MW of base load heat. Preliminary proposals for Stockholm involve a 2000 MW heat supply; proposals for Gothenburg are more tentative. The paper describes progress on these proposals and their technology. It also outlines technology under development to increase the economic range of large scale heat transport and to make distribution economic even for low heat-density family housing estates. Regions apart from the few major urban areas mentioned above require the adoption of a different approach. To this end the development of a small, simple low-temperature reactor for heat-only production suitable for urban location has been started in Sweden in close contact with Finland. Some results of the work in progress are presented, with emphasis on the safety requirements. An outline is given in the paper as to how problems of regional heat planning and institutional and legislative issues are being approached

Feed water control device in a reactor

International Nuclear Information System (INIS)

Okutani, Tetsuro.

1984-01-01

Purpose: To prevent substantial fluctuations of the water level in a nuclear reactor and always keep a constant standard level under any operation condition. Constitution: When the causes for fluctuating the reactor water level is resulted, a certain amount of correction signal is added to a level deviation signal for the difference between the reactor standard level and the actual reactor water level to control the flow rate of the feed water pump depending on the addition signal. If reactor scram should occur, for instance, a level correction signal changing stepwise depending on a scram signal is outputted and added to the level deviation signal. As the result, the flow rate of feed water sent into the reactor just after the scram is increased, whereby the lowering in the reactor water level upon scram can be decreased as compared with the case where no such level compensation signal is inputted. (Kamimura, M.)

Implementation of utilities operation and maintenance experience into the European pressurized water reactor design

International Nuclear Information System (INIS)

Zaiss, W.; Lallier, M.

1999-01-01

Since 1992 Electricite de France EDF and German Utilities GU work together with Nuclear Power International NPI, a subsidiary of Framatome and Siemens, in the development of the future European Pressurized Water Reactor EPR. The EPR is an evolutionary concept, based on the French N4 plants and the German KONVOI plants. From the beginning, experienced operation and maintenance people from the precursor plants participate at the design process. Their experience will lead to a plant, which is not only characterised by low investment costs, but also by good operability, high availability and low operation and maintenance costs. No expensive back-fittings should be necessary after commissioning, to reach these availability and maintenance targets. The utility specialists give design requirements for outage performance, system design, and layout. These design requirements are really determining the system performances, and not what was design basis before. It does not necessarily lead to system increases. Mainly it is a shifting of the emphasis to other items. There are even cases, where the system performances can be reduced. Mostly very small modifications, which are nearly cost neutral when implemented early in the design, have big impact on the further operation. If there are big cost influences, a sound balance between investment and gained availability is made together with the designers. There is very fruitful discussion between designers and operators, which is highly estimated by both sides. In this frame also new, revolutionary ideas are coming up, which are going mostly in the direction of investment cost reduction, without loosing operation freedom. It is the first time in Europe, that designers and operators are working so close together. It is also the first time, that the management and the decision making is dominated by the utilities. (author)

Water feeding method upon reactor isolation

International Nuclear Information System (INIS)

Sasaki, Koichi; Takahara, Kuniaki; Hamamura, Kenji; Arakawa, Masahiro.

1990-01-01

The present invention concerns a method of feeding water upon reactor isolation in a plural loop type reactor having a plurality of reactor cooling systems. Water can be injected to a plurality of pools even if the pressure between the pools is not balanced and the water level in the reactor cooling system is optimally controlled. That is, water can be injected in accordance with the amount required for each of the pools by setting the opening of a turbine inlet steam control valve to somewhat higher than the cooling system pressure of the highest pressure loop. Water feeding devices upon reactor isolation were required by the same number as that for the reactor cooling systems. Whereas since pumps and turbines are used in common without worsening the water injection controllability to each of the loops according to the method of this invention and, accordingly, the cost performance can be improved. Further, since the opening degree of the turbine inlet steam control valve is controlled while making the difference pressure constant between the turbine inlet pressure and the pump exhaust pressure, the amount of the turbine exhausted steams can be reduced and, further, water injection controllability of the flow rate control valve in the injection line is improved. (I.S.)

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)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Water injection device for reactor container

International Nuclear Information System (INIS)

Sakaki, Isao.

1996-01-01

A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

Thorium utilization in heavy water moderated Accelerator Driven Systems

International Nuclear Information System (INIS)

Bajpai, Anil; Degweker, S.B.; Ghosh, Biplab

2011-01-01

Research on Accelerator Driven Systems (ADSs) is being carried out around the world primarily with the objective of waste transmutation. Presently, the volume of waste in India is small and therefore there is little incentive to develop ADS based waste transmutation technology immediately. With limited indigenous U availability and the presence of large Th deposits in the country, there is a clear incentive to develop Th related technologies. India also has vast experience in design, construction and operation of heavy water moderated reactors. Heavy water moderated reactors employing solid Th fuels can be self sustaining, but the discharge burnups are too low to be economical. A possible way to improve the performance such reactors is to use an external neutron source as is done in ADS. This paper discusses our studies on Th utilization in heavy water moderated ADSs. The study is carried out at the lattice level. The time averaged k-infinity of the Th bundle from zero burnup up to the discharge burnup is taken to be the same as the core (ensemble) averaged k-infinity. For the purpose of the analysis we have chosen standard PHWR and AHWR assemblies. Variation of the pitch and coolant (H 2 O/D 2 O) are studied. Both, the once through cycle and the recycling option are studied. In the latter case the study is carried out for various enrichments (% 233 U in Th) of the recycled Th fuel bundles. The code DTF as modified for lattice and burnup calculations (BURNTRAN) was used for carrying out the study. The once through cycle represents the most attractive ADS concept (Th burner ADS) possible for Th utilization. It avoids reprocessing of Th spent fuel and in the ideal situation the use of any fissile material either initially or for sustaining itself. The gain in this system is however rather low requiring a high power accelerator and a substantial fraction of the power generated to be fed back to the accelerator. The self sustaining Th-U cycle in a heavy moderated ADS

Study on core design for reduced-moderation water reactors

International Nuclear Information System (INIS)

Okubo, Tsutomu

2002-01-01

The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

Study on core design for reduced-moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

Water-immersion type ship reactor

International Nuclear Information System (INIS)

Okada, Hiroki; Yamamura, Toshio.

1996-01-01

In a water immersion-type ship reactor in which a water-tight wall is formed around a pressure vessel by way of an air permeable heat insulation layer and immersing the wall under water in a reactor container, a pressure equalizing means equipped with a back flow check valve and introducing a gas in a gas phase portion above the water level of the container into a water tight wall and a relief valve for releasing the gas in the water tight wall to the reactor container are disposed on the water tight wall. When the pressure in the water tight wall exceeds the pressure in the container, the gas in the water tight wall is released from the relief valve to the reactor container. On the contrary, when the pressure in the container exceeds the pressure in the water tight wall, the gas in the gas phase portion is flown from the pressure equalizing means equipped with a back flow check valve to the inside of the water tight wall. Thus, a differential pressure between both of them is kept around 0kg/cm 2 . A large differential pressure is not exerted on the water tight wall thereby capable of preventing rupture of them to improve reliability, as well as the thickness of the plate can be decreased thereby enabling to moderate the design for the pressure resistance and reduce the weight. (N.H.)

Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test

International Nuclear Information System (INIS)

Cowell, B.S.

1997-06-01

This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy

The safety of light water reactors

International Nuclear Information System (INIS)

Pershagen, B.

1986-04-01

The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

Present and possible utilization of PUSPATI reactor

International Nuclear Information System (INIS)

Gui Ah Auu.

1983-01-01

The utilization of PUSPATI TRIGA Mark II Reactor (PTR) has increased reasonably well since its commissioning last year. PTR was used mainly for training of operators, neutron flux measurements and neutron activation analysis. However, the present utilization data indicates that further increase in PTR utilization to include teaching and the usage of the beam ports is desirable. Some possible areas of PTR applications in the future in relevance to our needs are also described in this paper. (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

Water level monitoring device in nuclear reactor

International Nuclear Information System (INIS)

Miura, Kiyohide; Otake, Tomohiro.

1988-01-01

Purpose: To monitor the water level in a pressure vessel of BWR type nuclear reactors at high accuracy by improving the compensation functions. Constitution: In the conventional water level monitor in a nuclear reactor, if the pressure vessel is displaced by the change of the pressure in the reactor or the temperature of the reactor water, the relative level of the reference water head in a condensation vessel is changed to cause deviation between the actual water level and the indicated water level to reduce the monitoring accuracy. According to the invention, means for detecting the position of the reference water head and means for detection the position in the condensation vessel are disposed to the pressure vessel. Then, relative positional change between the condensation vessel and the reference water head is calculated based on detection sinals from both of the means. The water level is compensated and calculated by water level calculation means based on the relative positional change, water level signals from the level gage and the pressure signals from the pressure gage. As a result, if the pressure vessel is displaced due to the change of the temperature or pressure, it is possible to measure the reactor water level accurately thereby remakably improve the reliability for the water level control in the nuclear reactor. (Horiuchi, T.)

TRIGA 14 MW Research Reactor Status and Utilization

International Nuclear Information System (INIS)

Barbos, D.; Ciocanescu, M.; Paunoiu, C.

2016-01-01

Institute for Nuclear Research is the owner of the largest family TRIGA research reactor, TRIGA14 MW research reactor. TRIGA14 MW reactor was designed to be operated with HEU nuclear fuel but now the reactor core was fully converted to LEU nuclear fuel. The full conversion of the core was a necessary step to ensure the continuous operation of the reactor. The core conversion took place gradually, using fuel manufactured in different batches by two qualified suppliers based on the same well qualified technology for TRIGA fuel, including some variability which might lead to a peculiar behaviour under specific conditions of reactor utilization. After the completion of the conversion a modernization program for the reactor systems was initiated in order to achieve two main objectives: safe operation of the reactor and reactor utilization in a competitive environment to satisfy the current and future demands and requirements. The 14 MW TRIGA research reactor operated by the Institute for Nuclear Research in Pitesti, Romania, is a relatively new reactor, commissioned 37 years ago. It is expected to operate for another 15-20 years, sustaining new fuel and testing of materials for future generations of power reactors, supporting radioisotopes production through the development of more efficient new technologies, sustaining research or enhanced safety, extended burn up and verification of new developments concerning nuclear power plants life extension, to sustain neutron application in physics research, thus becoming a centre for instruction and training in the near future. A main objective of the TRIGA14MW research reactor is the testing of nuclear fuel and nuclear material. The TRIGA 14 MW reactor is used for medical and industrial radioisotopes production ( 131 I, 125 I, 192 Ir etc.) and a method for 99 Mo- 99 Tc production from fission is under development. For nuclear materials properties investigation, neutron radiography methods have been developed in the INR. The

Suppression device for the reactor water level lowering

International Nuclear Information System (INIS)

Kasuga, Hajime; Kasuga, Hiroshi.

1984-01-01

Purpose: To suppress the lowering in the reactor water level so as to avoid unnecessary actuation of ECCS upon generation of transient changes which forecasts the lowering of the reactor water level in a BWR type reactor. Constitution: There are provided a water level suppression signal generator for generating a water level suppression signal upon generation of a transient change signal which forecasts the water level lowering in a nuclear reactor and a recycling flow rate controller that applies a recycling flow rate control signal to a recycling pump drive motor by the water level lowering suppression signal. The velocity of the recycling pump is controlled by a reactor scram signal by way of the water level lowering suppresion signal generator and a recycling flow rate controller. Then, the recycling reactor core flow rate is decreased and the void amount in the reactor is transiently increased where the water level tends to increase. Accordingly, the water level lowering by the scram is moderated by the increasing tendency of the water level. (Ikeda, J.)

Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-07-15

This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

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

The thorium fuel cycle in water-moderated reactor systems

International Nuclear Information System (INIS)

Critoph, E.

1977-01-01

Current interest in the thorium cycle, as an alternative to the uranium cycle, for water-moderated reactors is based on two attractive aspects of its use - the extension of uranium resources, and the related lower sensitivity of energy costs to uranium price. While most of the scientific basis required is already available, some engineering demonstrations are needed to provide better economic data for rational decisions. Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. There appear to be no major feasibility problems associated with the use of thorium, although development is required in the areas of fuel testing and fuel management. The use of thorium cycles implies recycling the fuel, and the major uncertainties are in the associated costs. Experience in the design and operation of fuel reprocessing and active-fabrication facilities is required to estimate costs to the accuracy needed for adequately defining the range of conditions economically favourable to thorium cycles. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An ''inventory'' of uranium of between 1 and 2Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium), is some two decades

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)

Radiological control aspects of the fabrication of the Light Water Breeder Reactor core (LWBR Development Program)

International Nuclear Information System (INIS)

Schultz, B.G.

1979-05-01

A description is presented of the radiological control aspects of the fabrication of the Light Water Breeder Reactor (LWBR) core. Included are the radiological control criteria applied for the design and use of fabrication facilities, the controls and limits imposed to minimize radiaion exposure to personnel, and an evaluation of the applied radiological program in meeting the program objectives. The goal of the LWBR program is to develop the technology to breed in light water reactors so that nuclear fuel may be used significantly more efficiently in these reactors. This technology is being developed by designing and fabricating a breeder reactor core, utilizing thoria (ThO 2 ) and binary thoria--urania (ThO 2 - 233 UO 2 ) fuel, to be operated in the existing pressurized water reactor plant owned by the Department of Energy at Shippingport, Pennsylvania

Pressurized water reactor flow arrangement

International Nuclear Information System (INIS)

Gibbons, J.F.; Knapp, R.W.

1980-01-01

A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

A plan of reactor physics experiments for reduced-moderation water reactors with MOX fuel in TCA

International Nuclear Information System (INIS)

Shimada, Shoichiro; Akie, Hiroshi; Suzaki, Takenori; Okubo, Tutomu; Usui, Shuji; Shirakawa, Toshihisa; Iwamura, Takamiti; Kugo, Teruhiko; Ishikawa, Nobuyuki

2000-06-01

The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors which aim at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. Critical Experiments performed so far in Eualope and Japan were reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of the equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX fuel rods used in the experiments are obtained by calculations and the modification of the equipment for the experiments are shown. (author)

Heavy water moderated tubular type nuclear reactor

International Nuclear Information System (INIS)

Oohashi, Masahisa.

1986-01-01

Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

Pressurized water reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

Modeling a Packed Bed Reactor Utilizing the Sabatier Process

Science.gov (United States)

Shah, Malay G.; Meier, Anne J.; Hintze, Paul E.

2017-01-01

A numerical model is being developed using Python which characterizes the conversion and temperature profiles of a packed bed reactor (PBR) that utilizes the Sabatier process; the reaction produces methane and water from carbon dioxide and hydrogen. While the specific kinetics of the Sabatier reaction on the RuAl2O3 catalyst pellets are unknown, an empirical reaction rate equation1 is used for the overall reaction. As this reaction is highly exothermic, proper thermal control is of the utmost importance to ensure maximum conversion and to avoid reactor runaway. It is therefore necessary to determine what wall temperature profile will ensure safe and efficient operation of the reactor. This wall temperature will be maintained by active thermal controls on the outer surface of the reactor. Two cylindrical PBRs are currently being tested experimentally and will be used for validation of the Python model. They are similar in design except one of them is larger and incorporates a preheat loop by feeding the reactant gas through a pipe along the center of the catalyst bed. The further complexity of adding a preheat pipe to the model to mimic the larger reactor is yet to be implemented and validated; preliminary validation is done using the smaller PBR with no reactant preheating. When mapping experimental values of the wall temperature from the smaller PBR into the Python model, a good approximation of the total conversion and temperature profile has been achieved. A separate CFD model incorporates more complex three-dimensional effects by including the solid catalyst pellets within the domain. The goal is to improve the Python model to the point where the results of other reactor geometry can be reasonably predicted relatively quickly when compared to the much more computationally expensive CFD approach. Once a reactor size is narrowed down using the Python approach, CFD will be used to generate a more thorough prediction of the reactors performance.

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

Nuclear fuel for light water reactors

International Nuclear Information System (INIS)

Etemad, A.

1976-01-01

The goal of the present speech is to point out some of the now-a-day existing problems related to the fuel cycle of light water reactors and to foresee their present and future solutions. Economical aspects of nuclear power generation have been considerably improving, partly through technological advancements and partly due to the enlargement of unit capacity. The fuel cycle, defined in the course of this talk, discusses the exploration, mining, ore concentration, purification, conversion, enrichment, manufacturing of fuel elements, their utilization in a reactor, their discharge and subsequent storage, reprocessing, and their re-use or disposal. Uranium market in the world and the general policy of several uranium owning countries are described. The western world requirement for uranium until the year 2000, uranium resources and the nuclear power programs in the United States, Australia, Canada, South Africa, France, India, Spain, and Argentina are discussed. The participation of Iran in a large uranium enrichment plant based on French diffusion technology is mentioned

Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors

Science.gov (United States)

Gordon, Barry; Garcia, Susan

Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

Water simulation of sodium reactors

International Nuclear Information System (INIS)

Grewal, S.S.; Gluekler, E.L.

1981-01-01

The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system

Controlling hydrogen behavior in light water reactors

International Nuclear Information System (INIS)

Cullingford, H.S.; Edeskuty, F.J.

1981-01-01

In the aftermath of the incident at Three Mile Island Unit 2 (TMI-2), a new and different treatment of the Light Water Reactor (LWR) risks is needed for public safety because of the specific events involving hydrogen generation, transport, and behavior following the core damage. Hydrogen behavior in closed environments such as the TMI-2 containment building is a complex phenomenon that is not fully understood. Hence, an engineering approach is presented for prevention of loss of life, equipment, and environment in case of a large hydrogen generation in an LWR. A six-level defense strategy is described that minimizes the possibility of ignition of released hydrogen gas and otherwise mitigates the consequences of hydrogen release. Guidance is given to reactor manufacturers, utility companies, regulatory agencies, and research organizations committed to reducing risk factors and insuring safety of life, equipment, and environment

Emergency water supply facility for nuclear reactor

International Nuclear Information System (INIS)

Karasawa, Toru

1998-01-01

Water is stored previously in an equipment storage pit disposed on an operator floor of a reactor building instead of a condensate storage vessel. Upon occurrence of an emergency, water is supplied from the equipment storage pit by way of a sucking pipeline to a pump of a high pressure reactor core water injection circuit and a pump of a reactor-isolation cooling circuit to supply water to a reactor. The equipment storage pit is arranged in a building so that the depth thereof is determined to keep the required amount of water by storing water at a level lower than the lower end of a pool gate during normal operation. Water is also supplied from the equipment storage pit by way of a supply pipeline to a spent fuel storage pool on the operation floor of the reactor building. Namely, water is supplied to the spent fuel storage pool by a pump of a fuel pool cooling and cleaning circuit. This can eliminate a suppression pool cleaning circuit. (I.N.)

Study of Pu consumption in Advanced Light Water Reactors

International Nuclear Information System (INIS)

1993-01-01

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

Status of neutron beam utilization at the Dalat nuclear research reactor

International Nuclear Information System (INIS)

Dien, Nguyen Nhi; Hai, Nguyen Canh

2003-01-01

The 500-kW Dalat nuclear research reactor was reconstructed from the USA-made 250-kW TRIGA Mark II reactor. After completion of renovation and upgrading, the reactor has been operating at its nominal power since 1984. The reactor is used mainly for radioisotope production, neutron activation analysis, neutron beam researches and reactor physics study. In the framework of the reconstruction and renovation project of the 1982-1984 period, the reactor core, the control and instrumentation system, the primary and secondary cooling systems, as well as other associated systems were newly designed and installed by the former Soviet Union. Some structures of the reactor, such as the reactor aluminum tank, the graphite reflector, the thermal column, horizontal beam tubes and the radiation concrete shielding have been remained from the previous TRIGA reactor. As a typical configuration of the TRIGA reactor, there are four neutron beam ports, including three radial and one tangential. Besides, there is a large thermal column. Until now only two-neutron beam ports and the thermal column have been utilized. Effective utilization of horizontal experimental channels is one of the important research objectives at the Dalat reactor. The research program on effective utilization of these experimental channels was conducted from 1984. For this purpose, investigations on physical characteristics of the reactor, neutron spectra and fluxes at these channels, safety conditions in their exploitation, etc. have been carried out. The neutron beams, however, have been used only since 1988. The filtered thermal neutron beams at the tangential channel have been extracted using a single crystal silicon filter and mainly used for prompt gamma neutron activation analysis (PGNAA), neutron radiography (NR) and transmission experiments (TE). The filtered quasi-monoenergetic keV neutron beams using neutron filters at the piercing channel have been used for nuclear data measurements, study on

Experimental utilization of the IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Bitelli, U. d'Utra; Santos, A. dos; Jerez, R.; Diniz, R.; Fanaro, L.C.C.B.; Abe, A.Y.; Moreira, J.M.L.; Fer, N.; Giada, M.R.; Fuga, R.

2003-01-01

This paper aims to show the experimental utilization of the IPEN/MB-01 nuclear reactor during the last fourteen years. The IPEN/MB-01 is a zero-power critical assembly specially designed to measure integral and differential reactor physics parameters to validate calculational methodologies and related nuclear data libraries. Experiments involving determination of spectral indices, critical mass, relative abundance of delayed neutrons, the inversion point of the isothermal reactivity coefficient and burnable poison are considered the most important experiments. Current experiments at IPEN/MB-01 reactor are also commented. (author)

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

Boiling water reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

Recent advances in the utilization and the irradiation technology of the refurbished BR2 reactor

International Nuclear Information System (INIS)

Dekeyser, J.; Benoit, P.; Decloedt, C.; Pouleur, Y.; Verwimp, A.; Weber, M.; Vankeerberghen, M.; Ponsard, B.

1999-01-01

Operation and utilization of the materials testing reactor BR2 at the Belgian Nuclear Research Centre (SCK·CEN) has since its start in 1963 always followed closely the needs and developments of nuclear technology. In particular, a multitude of irradiation experiments have been carried out for most types of nuclear power reactors, existing or under design. Since the early 1990s and increased focus was directed towards more specific irradiation testing needs for light water reactor fuels and materials, although other areas of utilization continued as well (e.g. fusion reactor materials, safety research, ...), including also the growing activities of radioisotope production and silicon doping. An important milestone was the decision in 1994 to implement a comprehensive refurbishment programme for the BR2 reactor and plant installations. The scope of this programme comprised very substantial studies and hardware interventions, which have been completed in early 1997 within planning and budget. Directly connected to this strategic decision for reactor refurbishment was the reinforcement of our efforts to requalify and upgrade the existing irradiation facilities and to develop advanced devices in BR2 to support emerging programs in the following fields: - LWR pressure vessel steel, - LWR irradiation assisted stress corrosion cracking (IASCC), - reliability and safety of high-burnup LWR fuel, - fusion reactor materials and blanket components, - fast neutron reactor fuels and actinide burning, - extension and diversification of radioisotope production. The paper highlights these advances in the areas of BR2 utilisation and the ongoing development activities for the required new generation of irradiations devices. (author)

Reactor Safety Commission Code of Practice for Pressurized Water Reactors

International Nuclear Information System (INIS)

1990-01-01

The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

Method of operating heavy water moderated reactors

International Nuclear Information System (INIS)

Masuda, Hiroyuki.

1980-01-01

Purpose: To enable stabilized reactor control, and improve the working rate and the safety of the reactor by removing liquid poison in heavy water while maintaining the power level constant to thereby render the void coefficient of the coolants negative in the low power operation. Method: The operation device for a heavy water moderated reactor comprises a power detector for the reactor, a void coefficient calculator for coolants, control rods inserted into the reactor, a poison regulator for dissolving poisons into or removing them out of heavy water and a device for removing the poisons by the poison regulator device while maintaining the predetermined power level or inserting the control rods by the signals from the power detector and the void coefficient calculator in the high temperature stand-by conditions of the reactor. Then, the heavy water moderated reactor is operated so that liquid poisons in the heavy water are eliminated in the high temperature stand-by condition prior to the start for the power up while maintaining the power level constant and the plurality of control rods are inserted into the reactor core and the void coefficient of the coolants is rendered negative in the low power operation. (Seki, T.)

Study relating to the physico-chemical behaviour of heavy water in nuclear reactors

International Nuclear Information System (INIS)

Chenouard, J.; Dirian, G.; Roth, E.; Vignet, P.; Platzer, R.

1959-01-01

Chemical and isotope pollution, and radiolytic decomposition are the two most important ways in which heavy water becomes degraded in nuclear reactors. Chemical pollution has led to the creation of ion exchange purification loops specially designed for reactors: the report contains a description in detail of the application of this purification method in CEA research reactors, including the analysis required, results obtained, and their interpretation. The intelligence obtained on radiolytic decomposition with the same facilities is also discussed, as well as the recombination apparatus and control equipment utilized. Finally, investigation to date in the CEA on recombination circuits for power reactors is also discussed. (author) [fr

Advances in light water reactor technologies

CERN Document Server

Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

2010-01-01

""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

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

Service to the Electric Utility Industry by the Ford Nuclear Reactor, University of Michigan

International Nuclear Information System (INIS)

Burn, R.R.; Simpson, P.A.; Cook, G.M.

1993-01-01

Since 1977, the staff of the University of Michigan's Ford Nuclear Reactor has been providing irradiation, testing, analytical, and training services to electric utilities and to suppliers of the nuclear electric utility industry. This paper discusses the reactor's irradiation facilities; reactor programs and utilization; materials testing programs; neutron activation analysis activities; and training programs conducted

Investigation on innovative water reactor for flexible fuel cycle (FLWR). (1) Conceptual design

International Nuclear Information System (INIS)

Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiko; Ohnuki, Akira; Iwamura, Takamichi

2005-01-01

A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI) in order to ensure sustainable energy supply in the future based on the well-experienced Light Water Reactor (LWR). The concept aims at effective and flexible utilization of uranium and plutonium resources through plutonium multiple recycling by two stages. In the first stage, the FLWR core realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The core in the second stage represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the core concepts in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of natural uranium resource, or establishment of economical reprocessing technology of MOX spent fuel. The FLWR is essentially a BWR-type reactor, and its core design is characterized by use of hexagonal-shaped fuel assemblies with the triangular-lattice fuel rod configuration of highly enriched MOX fuel, control rods with Y-shaped blades, and a short and flat core design. Detailed investigations have been performed on the core design, in conjunction with the other related studies such as on thermal hydraulics in the tight lattice core including experimental activities, and the results obtained so far have shown the proposed concept is feasible and promising. (author)

Needs of nuclear data for advanced light water reactor

International Nuclear Information System (INIS)

Chaki, Masao

2008-01-01

Hitachi has been developing medium sized ABWRs as a power source that features flexibility to meet various market needs, such as minimizing capital risks, providing a timely return on capital investments, etc. Basic design concepts of the medium sized ABWRs are 1) using the current ABWR design which has accumulated favorable construction and operation histories as a starting point; 2) utilizing standard BWR fuels which have been fabricated by proven technology; 3) achieving a rationalized design by suitably utilizing key components developed for large sized reactors. Development of the medium sized ABWRs has proceeded in a systematic, stepwise manner. The first step was to design an output scale for the 600MWe class reactor (ABWR-600), and the next step was to develop an uprating concept to extend this output scale to the 900MWe class reactor (ABWR-900) based on the rationalized technology of the ABWR-600 for further cost savings. In addition, Hitachi and MHI developed an ultra small reactor, 'Package-Reactor'. About the nuclear data, for the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO 2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO 2 cores and the MOX cores. (author)

Pressurised water reactor operation

International Nuclear Information System (INIS)

Birnie, S.; Lamonby, J.K.

1987-01-01

The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

Progress report on joint utilization of the Musashi reactor, 10(1985)

International Nuclear Information System (INIS)

1986-06-01

The results of the research by those who commonly utilized the reactor in Musashi Institute of Technology and others in fiscal year 1985 were compiled as a report. The number of application was one case of medical treatment, three cases related to living things, five experiments and 34 cases of general irradiation, and all were adopted. The number of actual utilization was all except one case related to living things. The actual result of the daily utilization from September, 1985 to March, 1986 is shown. The gists of 42 reports are collected. In fiscal year 1986, 44 researches are adopted for the common utilization of the reactor and others. The connection between the Musashi reactor and the large computer center in University of Tokyo was increased to 9600 bps, and 4800 bps are used for carrying out remote jobs as before, while 4800 bps are used for utilizing four TSS terminal equipments. (Kako, I.)

Reactor water quality degradation suppressing method upon reactor start up

International Nuclear Information System (INIS)

Maeda, Katsuharu.

1993-01-01

Preceding to reactor start-up, vacuum degree in a condenser is increased, and after the vacuum degree has been increased sufficiently, a desalting tower is inserted. Then, water feed to the reactor is started and the reactor is operated so that water is supplied gradually. Thus, dissolved oxygen in the feedwater and condensates is kept low and an entire organic carbon leaching rate from resins in the condensate desalting tower is reduced. Further, since feedwater is gradually supplied after the start-up, the entire organic carbon brought into the reactor is decomposed by heat and radiation and efficiently removed by a reactor coolant cleanup system. As a result, corrosion of stainless steel or the like is suppressed, as well as integrity of fuels can be maintained. Further, degradation of water quality can be suppressed effectively not by additionally putting the condensate desalting towers to in-service in accordance with the increase of the feedwater flow rate accompanying the power up but by previously putting the condensate desalting towers to in-service. (N.H.)

High converter pressurized water reactor with heavy water as a coolant

International Nuclear Information System (INIS)

Ronen, Y.; Reyev, D.

1983-01-01

There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

Summary report of the 7th reduced-moderation water reactor workshop

International Nuclear Information System (INIS)

Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao

2005-08-01

As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in Japan Atomic Energy Research Institute (JAERI). The workshop on RMWRs is aiming at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors, and has been held every year since 1998. The 7th workshop was held on March 5, 2004 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The program of the workshop was composed of 5 lectures and an overall discussion time. The workshop started with the lecture by JAERI on the status and future program of PMWR research and development, followed by the two presentations by JAERI and Japan Nuclear Cycle Development Institute, respectively, on the investigation and evaluation of water cooled reactor in Feasibility Study Program on Commercialized Fast Reactor Systems. The lectures were also made on the Japan's nuclear fuel cycle and scenarios for RMWRs deployment by JAERI, and on the next generation reactor development activity by Hitachi, Ltd. The main subjects of the overall discussion time were Na cooled fast reactor, deployment effects of RMWRs and the future plan of the RMWR research and development. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as of the discussion time. In addition in the Appendices, there are included presentation handouts of each lecture, program of the workshop and the participants list. (author)

Proceedings of the 2000 workshop on the utilization of research reactors

International Nuclear Information System (INIS)

2001-11-01

The 2000 Workshop on the Utilization of Research Reactors, which is the ninth Workshop on the theme of research reactor utilization, was held in Taejon, Korea from November 20 to 24. This Workshop was executed based on the agreement in the First Coordinator's Meeting of Forum for Nuclear Cooperation in Asia (FNCA) held in Tokyo, March 2000. The Workshop consists of three groups under themes of the following fields; 1) Neutron Scattering, 2) Neutron Activation Analysis, and 3) Research Reactor Utilization (BNCT). The total number of participants for the Workshop was about 60 people from 8 countries; Australia, China, Indonesia, Korea, Malaysia, The Philippines, Thailand, Vietnam, and Japan. The 37 of the presented papers are indexed individually. (J.P.N.)

Proceedings of the first symposium on utilization of research reactors and JMTR

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-03-01

The first symposium on utilization of research reactors (JRR-2, JRR-3M, JRR-4) and Japan Materials Testing Reactor (JMTR) in JAERI was held from September 29th to 30th, 1997 at Sannomaru Hotel, Mito. The purpose of this symposium is to announce contribution to progress of scientific technology as well as to promote future utilization of the research reactors and JMTR. During the symposium, 16 reports were presented on nuclear fuel and material, neutron beam experiment, medical irradiation, radioisotope production and neutron activation analysis. The present status of the research reactors and JMTR were also reported. The special lecture titled `JRR-2 and Medical Irradiation` was given by Mr. Nakamura, former editorial writer of Yomiuri. Finally, panel discussion was carried on `The Role of Research Reactors and JMTR in Scientific Technology for the future` actively by the participants and experts in every field of research reactor utilization. 250 people participated in this symposium from universities, national research institutes, private corporations and JAERI. This proceedings briefly summarizes 16 reports, the content of panel discussion and so forth. (J.P.N.)

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

Reactor water clean-up device

International Nuclear Information System (INIS)

Sawa, Toshio; Takahashi, Sankichi; Takashima, Yoshie.

1983-01-01

Purpose: To efficiently eliminate radioactive materials such as iron oxide and cobalt ions with less heat loss by the use of an electrode assembly applied with a direct current. Constitution: In a reactor water clean-up device adapted to pass reactor water through an electrode assembly comprising at least a pair of anode and cathode applied with a direct current to eliminate various types of ions contained in the reactor water by way of the electrolysis or charge neutralization at the anode, the cathode is constituted with a corrosion resistant grid-like or porous metal plate and a layer to the upper portion of the metal plate filled with a plurality of metal spheres of about 1 - 5 mm diameter, and the anode is made of insoluble porous or spirally wound metal material. (Seki, T.)

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

The current status of utilization of research reactors in China

International Nuclear Information System (INIS)

Luzheng, Yuan

2004-01-01

Seminars on utilization of research reactors were held to enhance experience exchanging among institutes and universities in China. The status of CARR (China Advanced Research Reactor) project is briefly described. The progress in BNCT program in China is introduced. (author)

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)

Utility industry evaluation of the Modular High-Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Burstein, S.; Bitel, J.S.; Tramm, T.R.; High, M.D.; Neils, G.H.; Tomonto, J.R.; Weinberg, C.J.

1990-02-01

A team of utility industry representatives evaluated the Modular High Temperature Gas-Cooled Reactor plant design, a current design created by an industrial team led by General Atomics under Department of Energy sponsorship and with support provided by utilities through Gas-Cooled Reactor Associates. The utility industry team concluded that the plant design should be considered a viable application of an advanced nuclear concept and deserves continuing development. Specific comments and recommendations are provided as a contribution toward improving a very promising plant design. 2 refs

Decay ratio estimation in pressurized water reactor

International Nuclear Information System (INIS)

Por, G.; Runkel, J.

1990-11-01

The well known decay ratio (DR) from stability analysis of boiling water reactors (BWR) is estimated from the impulse response function which was evaluated using a simplified univariate autoregression method. This simplified DR called modified DR (mDR) was applied on neutron noise measurements carried out during five fuel cycles of a 1300 MWe PWR. Results show that this fast evaluation method can be used for monitoring of the growing oscillation of the neutron flux during the fuel cycles which is a major concern of utilities in PWRs, thus it can be used for estimating safety margins. (author) 17 refs.; 10 figs

Activity report on the utilization of research reactors. Japanese fiscal year, 2002

International Nuclear Information System (INIS)

2004-08-01

During the fiscal year 2002, the Tokai Research Establishment research reactors carried out 7 cycles of joint use reactor operation at JRR-3 and 39 cycles at JRR-4. The research reactors are being utilized for various purposes including experimental studies such as neutron scattering, prompt gamma analysis, neutron radiography and medical irradiation (BNCT), and irradiation utilization such as neutron activation analysis of various samples, Irradiation Test of Reactor Materials and fission track. This volume contains 279 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analysis, reactor materials, prompt gamma analysis, and others, submitted by the users in JAERI and from other organizations. (author)

Activity report on the utilization of research reactors. Japanese fiscal year, 2003

International Nuclear Information System (INIS)

2005-09-01

During the fiscal year 2003, the Tokai Research Establishment research reactors carried out 8 cycles of joint use reactor operation at JRR-3 and 42 cycles at JRR-4. The research reactors are being utilized for various purposes including experimental studies such as neutron scattering, prompt gamma analysis, neutron radiography and medical irradiation (BNCT), and irradiation utilization such as neutron activation analysis of various samples, Irradiation Test of Reactor Materials and fission track. This volume contains 246 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analysis, reactor materials, prompt analysis, and others, submitted by the users in JAERI and from other organizations. (author)

Development of radiation resistant structural materials utilizing fission research reactors in Japan (Role of research reactors)

International Nuclear Information System (INIS)

Shikama, T.; Tanigawa, H.; Nozawa, T.; Muroga, T.; Aoyama, T.; Kawamura, H.; Ishihara, M.; Ito, C.; Kaneda, S.; Mimura, S.

2009-01-01

Structural materials for next-generation nuclear power systems should have a good radiation resistance, where the expected accumulation dose will largely exceed 10 dpa. Among several candidate materials, materials of five categories, 1. Austenitic steels, including high nickel alloys, 2. Low activation ferritic martensitic steels, 3. ODS steels (austenitic and ferritic), 4. Vanadium based alloys, 5. Silicon carbide composites (SiC/SiCf). All have been most extensively studied in Japan, in collaboration among industries, national institutes such as Japan Atomic Energy Agency (JAEA), National Institute for Fusion Science (NIFS) and National Institute for Materials Science (NIMS), and universities. The high nickel base alloys were studied for their low swelling behaviors mainly by the NIMS and the austenitic steels are studied for their reliable engineering data base and their reliable performance in irradiation environments mainly by the JAEA, mainly for their application in the near-term projects such as the ITER and the Sodium Cooled Fast Reactors. The most extensive studies are now concentrated on the Low Activation Ferritic Marsensitic steels and ODS steels, for their application in a demonstration fusion reactor and prototype sodium cooled fast reactors. Fundamental studies on radiation effects are carried out, mainly utilizing Japan Materials Testing Rector (JMTR) with its flexible irradiation ability, up to a few dpa. For higher dpa irradiation, a fast test reactor, JOYO is utilized up to several 10s dpa. Some international collaborations such as Japan/USA and Japan/France are effective to utilize reactors abroad, such as High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory, and sodium cooled high flux fast reactors in France. Silicon carbide based composites are extensively studied by university groups led by Kyoto University and the JAEA. For their performance in heavy irradiation environments, the Japan/USA collaboration plays an important role

Auxiliary water supply device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In the device of the present invention, a cooling condensation means is disposed to a steam discharge channel of a turbine for driving pumps to directly return condensates to the reactor, so that the temperature of the suppression pool water is not elevated. Namely, the cooling condensation means for discharged steams is disposed to the discharge channel of the turbine. The condensate channel from the cooling condensation means is connected to a sucking side of the turbine driving pump. With such a constitution, when the reactor is isolated from a main steam system, reactor scram is conducted. Although the reactor water level is lowered by the reactor scram, the lowering of the reactor water level is prevented by supplementing cooling water by the turbine driving pump using steams generated in the reactor as a power source. The discharged steams after driving the turbine are cooled and condensated by the cooling condensation means by way of the discharge channel and returned to the reactor again by way of the condensate channel. With such procedures, since the temperature of suppression pool water is not elevated, there is no need to operate other cooling systems. In addition, auxiliary water can be supplied for a long period of time. (I.S.)

The heavy water accountancy for research reactors in JAERI

International Nuclear Information System (INIS)

Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

1998-11-01

The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is used as moderator, coolant and reflector about 16 tons heavy water. The JRR-3M is a light water cooled and moderated pool type research reactor with a thermal power of 20MW and its is used as reflector about 7.3 tons heavy water. In the JRR-4, which is a light water cooled swimming pool type research reactor with the maximum thermal power of 3.5MW, about 1 ton heavy water is used to supply fully thermalized neutrons with a neutron beam experiment of facility. The heavy water was imported from U.S.A., CANADA and Norway. Parts of heavy water is internationally controlled materials, therefore management of heavy water is necessary for materials accountancy. This report described the change of heavy water inventories in each research reactors, law and regulations for accounting of heavy water in JAERI. (author)

Thorium Fuel Performance in a Tight-Pitch Light Water Reactor Lattice

International Nuclear Information System (INIS)

Kim, Taek Kyum; Downar, Thomas J.

2002-01-01

Research on the utilization of thorium-based fuels in the intermediate neutron spectrum of a tight-pitch light water reactor (LWR) lattice is reported. The analysis was performed using the Studsvik/Scandpower lattice physics code HELIOS. The results show that thorium-based fuels in the intermediate spectrum of tight-pitch LWRs have considerable advantages in terms of conversion ratio, reactivity control, nonproliferation characteristics, and a reduced production of long-lived radiotoxic wastes. Because of the high conversion ratio of thorium-based fuels in intermediate spectrum reactors, the total fissile inventory required to achieve a given fuel burnup is only 11 to 17% higher than that of 238 U fertile fuels. However, unlike 238 U fertile fuels, the void reactivity coefficient with thorium-based fuels is negative in an intermediate spectrum reactor. This provides motivation for replacing 238 U with 232 Th in advanced high-conversion intermediate spectrum LWRs, such as the reduced-moderator reactor or the supercritical reactor

Water cooled nuclear reactor

International Nuclear Information System (INIS)

1975-01-01

A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

Report on results of utilizing nuclear reactor, April, 1983 - March, 1984

International Nuclear Information System (INIS)

1984-01-01

The reactor in the Institute for Atomic Energy, Rikkyo University, is usually operated for 6 hours a day. It has been operated for 21 years without large trouble. The time of operation in fiscal year 1983 was 551 hours, and the cumulative power output was 47,511 kWh. From December, 1983, to March, 1984, the reactor was shut down, and the repair works for the secondary cooling system were carried out, accordingly, the power output and the number of utilization cases decreased. The total number of utilization cases in fiscal year 1983 was 2,702, and the total time of utilization was 8,497 hr 26 min. The utilization by universities accounted for 53.8%. A short period course was held as usual to educate and train 14 students. The radiation control for individuals and controlled areas is reported. The management of solid, gas end liquid wastes and their amounts being preserved in the Institute are reported. The radioactivity in environmental samples was measured 4 times in a year by total beta method, and the data on soil, sea bottom sediment, seawater and marine products are given. The summary of the research made by the personnel of the Institute and other achievement are also described. The execution of research in 1983 owed largely to the scientific research promotion fund of Japan Private School Promotion Foundation. The reactor was effectively utilized by the common utilization of the Rikkyo reactor. (Kako, I.)

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.

Possible physics modifications to CIRUS reactor core for improved reactor utilization

International Nuclear Information System (INIS)

John, Benjamin; Khosla, S.K.; Narain, Rajendra.

1976-01-01

Two fuelling schemes for uprating the neutron flux in CIRUS reactor at Trombay, are studied. One scheme employs enriched uranium-aluminium alloy boosters, the second envisages employing thorium oxide enriched with 0.2% plutonium oxide. It is seen that the second scheme has the potential of in-situ thorium utilization. (M.G.B.)

The present situations and perspectives on utilization of research reactors in Thailand

Science.gov (United States)

Chongkum, Somporn

2002-01-01

The Thai Research Reactor 1/Modification 1, a TRIGA Mark III reactor, went critical on November 7, 1977. It has been playing a central role in the development of both Office of Atomic Energy for Peace (OAEP) and nuclear application in Thailand. It has a maximum power of 2 MW (thermal) at steady state and a pulsing capacity of 2000 MW. The highest thermal neutron flux at a central thimber is 1×10 13 n/cm 2/s, which is extensively utilized for radioisotope production, neutron activation analysis and neutron beam experiments, i.e. neutron scattering, prompt gamma analysis and neutron radiography. Following the nuclear technological development, the OAEP is in the process of establishing the Ongkharak Nuclear Research Center (ONRC). The center is being built in Nakhon Nayok province, 60 km northeast of Bangkok. The centerpiece of the ONRC is a multipurpose 10 MW TRIGA research reactor. Facilities are included for the production of radioisotopes for medicine, industry and agriculture, neutron transmutation doping of silicon, and neutron capture therapy. The neutron beam facilities will also be utilized for applied research and technology development as well as training in reactor operations, performance of experiments and reactor physics. This paper describes a recent program of utilization as well as a new research reactor for enlarging the perspectives of its utilization in the future.

A case study for INPRO methodology based on Indian advanced heavy water reactor

International Nuclear Information System (INIS)

Anantharaman, K.; Saha, D.; Sinha, R.K.

2004-01-01

Under Phase 1A of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) a methodology (INPRO methodology) has been developed which can be used to evaluate a given energy system or a component of such a system on a national and/or global basis. The INPRO study can be used for assessing the potential of the innovative reactor in terms of economics, sustainability and environment, safety, waste management, proliferation resistance and cross cutting issues. India, a participant in INPRO program, is engaged in a case study applying INPRO methodology based on Advanced Heavy Water Reactor (AHWR). AHWR is a 300 MWe, boiling light water cooled, heavy water moderated and vertical pressure tube type reactor. Thorium utilization is very essential for Indian nuclear power program considering the indigenous resource availability. The AHWR is designed to produce most of its power from thorium, aided by a small input of plutonium-based fuel. The features of AHWR are described in the paper. The case study covers the fuel cycle, to be followed in the near future, for AHWR. The paper deals with initial observations of the case study with regard to fuel cycle issues. (authors)

Procedure for operating a heavy water cooled power reactor

International Nuclear Information System (INIS)

Rau, P.; Kumpf, H.

1981-01-01

Nuclear reactors cooled by heavy water usually have equipment for fuel element exchange during operation, with the primary circuit remaining contained. This fuel element exchange equipment is expensive and complicated in many respects. According to the invention, the heavy water is therefore replaced by light water after a certain time of operation in such way that light water is led in and heavy water is led off. After the replacement, at least a quarter of the fuel elements of the reactor core is exchanged with the reactor pressure vessel being open. Then the light water serving as a shielding is replaced by heavy water, with the reactor pressure vessel being closed. The invention is of interest particularly for high-conversion reactors. (orig.) [de

Extensive utilization of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

2003-01-01

Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

Summary of the fourth conference on United States utility experience in reactor noise analysis

International Nuclear Information System (INIS)

Fry, D.N.

1987-01-01

The fourth informal conference on United States utility experience in reactor noise analysis and loose-part monitoring was held at the Northeast Utilities Service Company offices in Hartford, Connecticut, May 12-14, 1987. Host and general chairman for the meeting was J.V. Persio of Northeast Utilities. This conference provided a forum where utilities could share information on reactor noise analysis on an informal basis. There were about 60 attendees at the meeting representing 10 utilities, 3 reactor vendors, 8 consulting organizations, and 4 universities and research laboratories. Twenty-three papers were presented at the conference, dealing with various aspects of loose-part monitoring, neutron noise analysis, and standards activities

Comparison of two thorium fuel cycles for use in light water prebreeder/breeder reactor systems (AWBA Development Program)

International Nuclear Information System (INIS)

Merriman, F.C.; McCoy, D.F.; Boyd, W.A.; Dwyer, J.R.

1983-05-01

Light water prebreeder/breeder conceptual reactor systems have been developed which have the potential to significantly improve the fuel utilization of present generation light water reactors. The purpose of this study is to describe and compare two possible types of thorium fuel cycles for use in these light water prebreeder and breeder concepts. The two types of thorium fuel cycles basically differ in the fuel rod design used in the prebreeder cores and the uranium isotopic concentration of fuel supplied to the breeder cores

Feedwater processing method in a boiling water reactor

Energy Technology Data Exchange (ETDEWEB)

Izumitani, M; Tanno, K

1976-09-06

The purpose of the invention is to decrease a quantity of corrosion products moving from the feedwater system to the core. Water formed into vapor after heated in a reactor is fed to the turbine through a main steam line to drive a generator to return it to liquid-state water in a condenser. The water is then again cycled into the reactor via the condensate pump, desalting unit, low pressure feedwater heater, medium pressure feedwater heater, and high pressure feedwater heater. The reactor water is recycled by a recycling pump. At this time, the reactor water recycled by the recycling pump is partially poured into a middle point between the desalting unit and the low pressure feedwater heater through a reducing valve or the like. With the structure described above, the quantity of the corrosion products from the feedwater system may be decreased by the function of a large quantity of active oxygen contained in the reactor water.

European utilities requirements for future reactors

International Nuclear Information System (INIS)

Roch, M.

1996-01-01

The prospect for future nuclear power plants has led the utilities of seven European countries to launch an effort to define the requirements that should be common to all utilities for the next reactors to be built in Europe. These requirements will ultimately be part of a four-volume document and will cover all aspects of a plant: performance, grid connection, codes and standards, materials, quality assurance, cost, and, of course, safety. The seven European countries - France, the United Kingdom, Germany, Spain, Italy, the Netherlands, and Belgium - issued revision A of Vols 1 and 2 in Nov. 1994, which deal with all the general requirements, not specific to any design, originally issued in March 1994. Comments were requested from most of the nuclear utilities as well as from reactor vendors worldwide. This gave rise to an enormous number of comments, which were duly considered by the European Union. The relevant ones were incorporated into revision B of Vols. 1 and 2, which was issued in Nov 1995, the objective of this revision B being essentially to gain approval from the safety authorities. A particular aspect of the European approach resides in the fact that these European requirements will have to be discussed and agreed on by at least nine safety authorities, i.e., the authorities of the seven counties that launched revision B, plus the authorities of two newcomers, Finland and Sweden, which have just applied for European Union membership

Transactions of the twenty-fifth water reactor safety information meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1997-09-01

This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session.

Transactions of the twenty-fifth water reactor safety information meeting

International Nuclear Information System (INIS)

Monteleone, S.

1997-09-01

This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session

Proceedings (slides) of the OECD/NEA Workshop on Innovations in Water-cooled Reactor Technologies

International Nuclear Information System (INIS)

Spiler, Joze; Kim, Sang-Baik; ); Feron, Fabien; Jaervinen, Marja-Leena; Husse, Julien; ); Ferraro, Giovanni; Bertels, Frank; Denk, Wolfgang; Tuomisto, Harri; Golay, Michael; Buongiorno, J.; Todreas, N.; Adams, E.; Briccetti, A.; Jurewicz, J.; Kindfuller, V.; Srinivasan, G.; Strother, M.; Minelli, P.; Fasil, E.; Zhang, J.; Genzman, G.; Epinois, Bertrand de l'; Kim, Shin Whan; Laaksonen, Jukka; Maltsev, Mikhail; Yu, CHongxing; Powell, David; Gorgemans, Julie; Hopwood, Jerry; Bylov, Igor; Bakhmetyev, Alexander M.; Lepekhin, Andrey N.; Fadeev, Yuriy P.; Bruna, Giovanni; Gulliford, Jim; ); Ham-Su, Rosaura; Thevenot, Caroline; GAUTIER, Guy-Marie; MARSAULT, Philippe; PIGNATEL, Jean-Francois; White, Andrew; )

2015-02-01

New technologies and solutions have been developed over more than thirty years to improve the safety, performance and economics of nuclear power plants. Particular efforts were made in designing systems to prevent or mitigate nuclear accidents and, greatly limit or even avoid any offsite release of radioactivity. Reactor designs developed in the 1980's and later are often referred to as Generation III (Gen III) reactors. They offer enhanced safety compared to earlier Generation II (Gen II) designs, as well as improved performance and economics. Examples of Gen III safety design features include solutions for corium localisation, advanced containment structures, improved emergency core-cooling systems, filtered venting systems, hydrogen risk management solutions, etc. Some of these solutions have also been back-fitted or partially adapted to existing reactors, based on recommendations from regulators or modernisation efforts by the utilities operating these reactors, to bring their level of safety to levels approaching those of the more modern designs. Other innovations found in the latest water-cooled reactor designs include the use of passive safety systems, and often associated with those, a simplification in the design of the reactor. Gen III reactors also feature better economics, for example increased design lifetime up to 60 years, ability to use 100% MOX fuel and operate with higher flexibility, higher thermal efficiencies and reduced staff requirements. Modularity is often quoted as a feature of some Gen III designs as a way of reducing the construction times and simplifying the decommissioning of the plant. The scope of the Workshop includes, inter alia: - Evolution of regulatory and design requirements for commercial water-cooled reactors; - Innovations in water-cooled reactor technologies that allowed significant improvement in the level of safety, with a discussion on advantages and challenges of active vs. passive safety systems; - Innovations under

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)

Analysis of strategies for improving uranium utilization in pressurized water reactors. Annual technical progress report for FY 1980

International Nuclear Information System (INIS)

Sefcik, J.A.; Driscoll, M.J.; Lanning, D.D.

1981-01-01

Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal objective has been the evaluation of suggested improvements on a self-consistent basis, allowing for concurrent changes in dependent variables such as core leakage and batch power histories, which might otherwise obscure the sometimes subtle effects of interest. Two levels of evaluation have been devised: a simple but accurate analytic model based on the observed linear variations in assembly reactivity as a function of burnup; and a numerical approach, embodied in a computer program, which relaxes this assumption and combines it with empirical prescriptions for assembly (or batch) power as a function of reactivity, and core leakage as a function of peripheral assembly power. State-of-the-art physics methods, such as PDQ-7, were used to verify and supplement these techniques

Pressurised water reactor in the UK

International Nuclear Information System (INIS)

Anon.

1982-01-01

Since the Three Mile Island accident there has been much debate about the safety considerations of Pressurised Water Reactors. Their development will continue throughout the world but it will be based upon the lessons learned from that unfortunate accident. In the United Kingdom there is a public enquiry discussing all aspects of the reactor. The papers given in this book provide an informed addition to the literature. The design, safety and licensing and construction of a pressurised water reactor system are discussed in detail. Considerations stemming from the Three Mile Island accident are presented

Experimental determination of lattice parameters for 2% enriched uranium heavy water reactor cores

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Takac, S; Markovic, H; Bosevski, T [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

1963-04-15

Systematic measurements of the buckling, infinite multiplication factor and the thermal utilization factor were made on a series of lattices for 2% enriched uranium tubular fuel elements in heavy water. This work represents the first phase of experimental verification of standard theoretical methods used for the determination of reactor parameters.

Effect of water impurities on stress corrosion cracking in a boiling water reactor

International Nuclear Information System (INIS)

Ljungbery, L.G.; Cubicciotti, D

1985-01-01

A series of stress corrosion tests, including corrosion potential and water chemistry measurements, has been performed in the Swedish Ringhals-1 boiling water reactor. Tests have been run under reactor start-up and reactor power operation with normal reactor water conditions and with alternate water chemistry in which hydrogen is added to the feedwater to suppress stress corrosion cracking. During one alternate water chemistry test, there was significant intergranular corrosion cracking of sensitized stainless specimens. It is shown that nitrate and sulfate, arising from an accidental resin intrusion, are likely causes. Nitrate increases the oxidizing power of the water, and sulfate enhances cracking under oxidizing conditions. During another test under start-up conditions, enhanced transgranular stress corrosion cracking in low alloy steels and possibly initiation of cracking in a nickel base alloy was observed as a result of resin intrusion into the reactor water. The intrusion produced acid and sulfate, which are believed to enhance hydrogen cracking conditions

Guidebook on quality control of mixed oxides and gadolinium bearing fuels for light water reactors

International Nuclear Information System (INIS)

1991-02-01

Under the coverage of an efficient quality assurance system, quality control in nuclear fuel fabrication is an essential element to assure the reliable performance of all its components in service. Incentives to increase fuel performance, by extending reactor cycles or achieving higher burnups and, in some countries to use recycled plutonium in light water reactors (LWRs) necessitated the development of new types of fuels. In the first case, due to higher uranium enrichments, a burnable neutron absorber was integrated to the fuel pellets. Gadolinia was found to form a solid solution with Uranium dioxide and, to present a burnup rate which matches fissile uranium depletion. (U,Gd)O 2 fuels which have been successfully used since the seventies, in boiling water reactors have more recently found an increased utilization, in pressurized water reactors. This amply justifies the publication of this TECDOC to encourage authorities, designers and manufacturers of these types of fuel to establish a more uniform, adapted and effective system of control, thus promoting improved materials reliability and good performance in advanced fuel for light water reactors. The Guidebook is subdivided into four chapters written by different authors. A separate abstract was prepared for each of these chapters. Refs, figs and tabs

Future directions in boiling water reactor design

International Nuclear Information System (INIS)

Wilkins, D.R.; Hucik, S.A.; Duncan, J.D.; Sweeney, J.I.

1987-01-01

The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuver-ability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proven features from BWR designs in Europe, Japan and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electrohydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling netwoek; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced trubine/generator with 52'' last stage buckets; and advanced radwaste technology. The ABWR is ready for lead plant application in Japan, where it is being developed as the next generation Japan standard BWR under the guidance and leadership of The Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. In the United States it is being adapted to the needs of US utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the US Nuclear Regulatory Commission for certification as a preapproved US standard BWR under the US Department of Energy's ALWR Design Verification Program. These cooperative Japanese and US programs are expected to establish the ABWR as a world class BWR for the 1990's...... (author)

Activity report on the utilization of research reactors. Japanese Fiscal Year, 1998

Energy Technology Data Exchange (ETDEWEB)

Toyoda, Masayuki; Koyama, Yoshimi [eds.; Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-03-01

This is the second issue of the activity report on the utilization of research reactors in the fields of neutron beam experiments, neutron activation analysis, radioisotope production, etc., performed during Japanese Fiscal Year 1998 (April 1, 1998 - March 31, 1999). All reports in this volume were described by users from JAERI and also users from the other organizations, i.e., universities, national research institutes and private companies, who have utilized our research reactor utilization facilities for the purpose of the above studies. (author)

Activity report on the utilization of research reactors. Japanese Fiscal Year, 1999

International Nuclear Information System (INIS)

Toyoda, Masayuki

2001-03-01

This is the second issue of the activity report on the utilization of research reactors in the fields of neutron beam experiments, neutron activation analysis, radioisotope production, etc., performed during Japanese Fiscal Year 1999 (April 1, 1999 - March 31, 2000). All reports in this volume were described by users from JAERI and also users from the other organizations, i.e., universities, national research institutes and private companies, who have utilized our research reactor utilization facilities for the purpose of the above studies. (author)

Developing strategic plans for effective utilization of research reactors

International Nuclear Information System (INIS)

Ridikas, Danas

2015-01-01

Strategic plans are indispensable documents for research reactors (RRs) to ensure their efficient, optimized and well managed utilization. A strategic plan provides a framework for increasing utilization, while helping to create a positive safety culture, a motivated staff, a clear understanding of real costs and a balanced budget. A strategic plan should be seen as an essential tool for a responsible manager of any RR, from the smallest critical facility to the largest reactor. Results and lessons learned are shown from the IAEA efforts to help the RR facilities developing strategic plans, provide review and advise services, organize national and regional stakeholder/user workshops, prepare further guidance and recommendations, document and publish guidance documents and other supporting materials.

Developing strategic plans for effective utilization of research reactors

Energy Technology Data Exchange (ETDEWEB)

Ridikas, Danas [International Atomic Energy Agency, Vienna (Austria). Dept. of Nuclear Sciences and Applications

2015-12-15

Strategic plans are indispensable documents for research reactors (RRs) to ensure their efficient, optimized and well managed utilization. A strategic plan provides a framework for increasing utilization, while helping to create a positive safety culture, a motivated staff, a clear understanding of real costs and a balanced budget. A strategic plan should be seen as an essential tool for a responsible manager of any RR, from the smallest critical facility to the largest reactor. Results and lessons learned are shown from the IAEA efforts to help the RR facilities developing strategic plans, provide review and advise services, organize national and regional stakeholder/user workshops, prepare further guidance and recommendations, document and publish guidance documents and other supporting materials.

Alternative water injection device to reactor equipment facility

International Nuclear Information System (INIS)

Yamashita, Masahiro.

1995-01-01

The device of the present invention injects water to the reactor and the reactor container continuously for a long period of time for preventing occurrence of a severe accident in a BWR type reactor and maintaining the integrity of the reactor container even if the accident should occur. Namely, diesel-driven pumps disposed near heat exchangers of a reactor after-heat removing system (RHR) are operated before the reactor is damaged by the after heat to cause reactor melting. A sucking valve disposed to a pump sucking pipeline connecting a secondary pipeline of the RHR heat exchanger and the diesel driving pump is opened. A discharge valve disposed to a pump discharge pipeline connecting a primary pipeline of the RHR heat exchanger and the diesel driving pump is opened. With such procedures, sea water is introduced from a sea water taking port through the top end of the secondary pipeline of the RHR heat exchanger and water is injected into the inside of the pressure vessel or the reactor container by way of the primary pipeline of the RHR heat exchanger. As a result, the reactor core is prevented from melting even upon occurrence of a severe accident. (I.S.)

Activity report on the utilization of research reactors. Japanese fiscal year, 2004

International Nuclear Information System (INIS)

2006-03-01

During the fiscal year 2004, the Nuclear Science Research Institute research reactors carried out 7 cycles of joint use reactor operation at JRR-3 and 41 cycles at JRR-4. The research reactors are being utilized for various purposes including experimental studies such as neutron scattering, prompt gamma analysis, neutron radiography and medical irradiation (BNCT), and irradiation utilization such as neutron activation analysis of various samples, RI production, Irradiation Test of Reactor Materials and fission track, advanced Science Research. This volume contains 235 activity reports, which are categorized into the fields of neutron scattering (10 subcategories), neutron radiography, neutron activation analysis, RI Production, prompt gamma analysis, and others, submitted by the users in JAEA and from other organizations. (author)

To the analysis of reactor noise in boiling water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1972-01-01

The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

Transactions of the nineteenth water reactor safety information meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1991-10-01

This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately

HOMOGENEOUS NUCLEAR POWER REACTOR

Science.gov (United States)

King, L.D.P.

1959-09-01

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

Emergency cooling of presurized water reactor

International Nuclear Information System (INIS)

Sykora, D.

1981-01-01

The method described of emergency core cooling in the pressurized water reactor is characterized by the fact that water is transported to the disturbed primary circuit or direct to the reactor by the action of the energy and mass of the steam and/or liquid phase of the secondary circuit coolant, which during emergency core cooling becomes an emergency cooling medium. (B.S.)

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

Light and heavy water replacing system in reactor container

International Nuclear Information System (INIS)

Miyamoto, Keiji.

1979-01-01

Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

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.

Twenty years of chemistry associated with the needs and utilization of nuclear reactors at the 'Boris Kidric' Institute of nuclear sciences, Vinca, Yugoslavia

International Nuclear Information System (INIS)

1969-01-01

This publication covers nine review papers on the following topics related to the needs and utilization of nuclear reactors in the Boris Kidric Institute of nuclear sciences during previous twenty years: radiochemistry, hot atom chemistry, isotope production, spent nuclear fuel reprocessing, chemistry of transuranium elements; liquid radioactive waste processing, purification of reactor coolant water by inorganic ion exchangers, research related to deuterium concentration processes, and chemical dosimetry at the RA reactor [sr

Aspects on thorium utilization in heavy water reactors

International Nuclear Information System (INIS)

1978-11-01

Some of the main problems of the Th - PHWR cycles are analyzed. With respect to the burnup limitations introduced by SSET cycle conditions and the burnup sensitivities of this reference cycle, estimates are presented using an integrated neutron-heavy element balance method. A PHWR of 1 GW(e) very similar to the CANDU current design was selected. In the case of 0.5% uranium losses, 11000 MWD/tHE and 13000 MWD/tHE were considered for U-235, respectively, for the Pu initialization of the cycle, the corresponding inventory being 4 t U-235/GW(e) and 5 t Pu (with 72% fissile content) per GW(e) for one year delay time between reactor out to reactor in, 66% capacity factor, 27 MW (fission)/tHE medium specific power. The following aspects are also analyzed: Safety problems associated with low delayed neutron fraction values; High and intermediate burnup fuel elements conceptual problems; Specific problems of thorium reprocessing; Specific problems for radioactive wastes and thorium storage; U-232 content evaluations and related fuel fabrication problems

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

Development and optimization of water treatment reactors using TiO2-modified polymer beads with a refractive index identical to that of water

Science.gov (United States)

Myoga, Arata; Iwashita, Ryutaro; Unno, Noriyuki; Satake, Shin-ichi; Taniguchi, Jun; Yuki, Kazuhisa; Seki, Yohji

2018-06-01

Various water purification reactors were constructed using beads of TiO2-coated MEXFLON, which is a fluoropolymer exhibiting a refractive index identical to that of water. The performance of these reactors was evaluated in a recirculation experiment utilizing an aqueous solution of methylene blue. Reactor pipes (length = 150 mm, internal diameter = 10 mm) were made of a fluorinated ethylene polymer with a refractive index of 1.338 and contained 206-bead clusters. A UV lamp was used to irradiate eight reactor pipes surrounding it. The above-mentioned eight bead-packed pipes were connected both in series and in parallel, and the performances of these two reactor types were compared. A pseudo-first-order rate constant of 0.70 h- 1 was obtained for the series connection, whereas the corresponding value for the parallel connection was 1.5 times smaller, confirming the effectiveness of increasing the reaction surface by employing a larger number of beads.

Development and optimization of water treatment reactors using TiO2-modified polymer beads with a refractive index identical to that of water

Science.gov (United States)

Myoga, Arata; Iwashita, Ryutaro; Unno, Noriyuki; Satake, Shin-ichi; Taniguchi, Jun; Yuki, Kazuhisa; Seki, Yohji

2018-03-01

Various water purification reactors were constructed using beads of TiO2-coated MEXFLON, which is a fluoropolymer exhibiting a refractive index identical to that of water. The performance of these reactors was evaluated in a recirculation experiment utilizing an aqueous solution of methylene blue. Reactor pipes (length = 150 mm, internal diameter = 10 mm) were made of a fluorinated ethylene polymer with a refractive index of 1.338 and contained 206-bead clusters. A UV lamp was used to irradiate eight reactor pipes surrounding it. The above-mentioned eight bead-packed pipes were connected both in series and in parallel, and the performances of these two reactor types were compared. A pseudo-first-order rate constant of 0.70 h- 1 was obtained for the series connection, whereas the corresponding value for the parallel connection was 1.5 times smaller, confirming the effectiveness of increasing the reaction surface by employing a larger number of beads.

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

Volatile fatty acid formation and utilization in anaerobic sulphidogenic batch reactors

CSIR Research Space (South Africa)

Greben, HA

2006-05-01

Full Text Available four stirred batch-test reactors (2 l) were operated, fed with artificial SO4 rich (1700 mg/l) feed water and tap water (controls). The reactors received sulphate reducing bacteria, compost bacteria and grass cuttings. The experimental period was 25...

Self-Sustaining Thorium Boiling Water Reactors

Directory of Open Access Journals (Sweden)

Ehud Greenspan

2012-10-01

Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

Fuel cycle options for light water reactors and heavy water reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1999-11-01

In the second half of the 20th century nuclear power has evolved from the research and development environment to an industry that supplies 16% of the world's electricity. By the end of 1997, over 8500 reactor-years of operating experience had been accumulated. Global environmental change, and the continuing increase in global energy supply required to provide increasing populations with an improving standard of living, make the contribution from nuclear energy even more important for the next century. For nuclear power to achieve its full potential and make its needed contribution, it must be safe, economical, reliable and sustainable. All of these factors can be enhanced by judicious choice and development of advanced fuel cycle options. The Technical Committee Meeting (TCM) on Fuel Cycle Options for Light Water Reactors and Heavy Water Reactors was hosted by Atomic Energy of Canada Limited (AECL) on behalf of the Canadian Government and was jointly conducted within the frame of activities of the IAEA International Working Group on Advanced Technologies for Light Water Reactors (IWG-LWR) and the IAEA International Working Group on Advanced Technologies for Heavy Water Reactors (IWG-HWR). The TCM provided the opportunity to have in-depth discussions on important technical topics which were highlighted in the International Symposium on Nuclear Fuel Cycle and Reactor Strategies: Adjusting to New Realities, held in Vienna, 3-6 June 1997. The main results and conclusions of the TCM were presented as input for discussion at the first meeting of the IAEA newly formed International Working Group on Fuel Cycle Options

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.)

TA-2 Water Boiler Reactor Decommissioning Project

International Nuclear Information System (INIS)

Durbin, M.E.; Montoya, G.M.

1991-06-01

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

Reactor performance calculations for water reactors

International Nuclear Information System (INIS)

Hicks, D.

1970-04-01

The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

Present status of operation and utilization of Kyoto University Reactor, KUR

International Nuclear Information System (INIS)

Kimura, Itsuro

1988-01-01

The Research Reactor Institute was established as an inter-university research institute in 1963. The main installation of the Institute is the KUR, a light water moderated, tank type reactor of 5,000 kW. In addition, a 46 MeV electron linear accelerator and a gamma ray irradiation facility with 10,000 Ci Co-60 are actively used for research. In 1974, Kyoto University Critical Assembly (KUCA) was constructed, and it has been used for research and education. The Reactor Utilization Center and the Fundamental Research Laboratory for Neutron Therapy were established in 1975 and 1976, respectively. Approximately 200 people work there, of them, some 80 do research and education, including 13 professors and 12 associate professors. All the experimental facilities of the Institute are available for the cooperative research projects of other universities and public research institutions in the fields of natural science and engineering, medical science, agriculture and forestry, fishery and stock-raising, environment science, cultural science and others. As a rule, the KUR is operated for about 70 hours from Tuesday morning to Friday evening every week. The annual examination by the government is carried out in spring. The total operation time was about 45,000 hours as of the end of 1987. The recent topics are reported. (Kako, I.)

Transactions of the Twenty-First Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Monteleone, S.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session

Transactions of the Twenty-First Water Reactor Safety Information Meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.

Transactions of the twenty-second water reactor safety information meeting

International Nuclear Information System (INIS)

1994-10-01

This report contains summaries of papers on reactor safety research to be presented at the 22nd Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 24--26, 1994. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session. Individual papers have been cataloged separately

Reliability assurance programme guidebook for advanced light water reactors

International Nuclear Information System (INIS)

2001-12-01

To facilitate the implementation of reliability assurance programmes (RAP) within future advanced reactor programmes and to ensure that the next generation of commercial nuclear reactors achieves the very high levels of safety, reliability and economy which are expected of them, in 1996, the International Atomic Energy Agency (IAEA) established a task to develop a guidebook for reliability assurance programmes. The draft RAP guidebook was prepared by an expert consultant and was reviewed/modified at an Advisory Group meeting (7-10 April 1997) and at a consults meeting (7-10 October 1997). The programme for the RAP guidebook was reported to and guided by the Technical Working Group on Advanced Technologies for Light Water Reactors (TWG-LWR). This guidebook will demonstrate how the designers and operators of future commercial nuclear plants can exploit the risk, reliability and availability engineering methods and techniques developed over the past two decades to augment existing design and operational nuclear plant decision-making capabilities. This guidebook is intended to provide the necessary understanding, insights and examples of RAP management systems and processes from which a future user can derive his own plant specific reliability assurance programmes. The RAP guidebook is intended to augment, not replace, specific reliability assurance requirements defined by the utility requirements documents and by individual nuclear steam supply system (NSSS) designers. This guidebook draws from utility experience gained during implementation of reliability and availability improvement and risk based management programmes to provide both written and diagrammatic 'how to' guidance which can be followed to assure conformance with the specific requirements outlined by utility requirements documents and in the development of a practical and effective plant specific RAP in any IAEA Member State

Grey water treatment in UASB reactor at ambient temperature.

Science.gov (United States)

Elmitwalli, T A; Shalabi, M; Wendland, C; Otterpohl, R

2007-01-01

In this paper, the feasibility of grey water treatment in a UASB reactor was investigated. The batch recirculation experiments showed that a maximum total-COD removal of 79% can be obtained in grey-water treatment in the UASB reactor. The continuous operational results of a UASB reactor treating grey water at different hydraulic retention time (HRT) of 20, 12 and 8 hours at ambient temperature (14-24 degrees C) showed that 31-41% of total COD was removed. These results were significantly higher than that achieved by a septic tank (11-14%), the most common system for grey water pre-treatment, at HRT of 2-3 days. The relatively lower removal of total COD in the UASB reactor was mainly due to a higher amount of colloidal COD in the grey water, as compared to that reported in domestic wastewater. The grey water had a limited amount of nitrogen, which was mainly in particulate form (80-90%). The UASB reactor removed 24-36% and 10-24% of total nitrogen and total phosphorus, respectively, in the grey water, due to particulate nutrients removal by physical entrapment and sedimentation. The sludge characteristics of the UASB reactor showed that the system had stable performance and the recommended HRT for the reactor is 12 hours.

Utilization of large electromagnetic pumps in the fast breeder reactors

International Nuclear Information System (INIS)

Deverge, C.; Lefrere, J.P.; Peturaud, P.; Sauvage, M.

1984-04-01

After an overview concerning the induction annular electromagnetic pumps and the dimensioning methods usually utilized, development of these components for a fast breeder integrated reactor is considered: - utilization of cooled EMP in the intermediate circuit, - utilization of immersed pumps, coupled with the intermediate exchanger, for the primary pumping; dimensioning, energetic aspects, and effects on the power plant geometrical configurations [fr

The CANDUR Reactor - The Practical Path to RU and TH use in Nuclear Reactors

International Nuclear Information System (INIS)

Kuran, Sermet; Yang, Dezi

2012-01-01

The CANDU heavy water reactor has unrivalled flexibility for using a variety of fuels, such as Natural Uranium (NU), Low Enriched Uranium (LEU), Recycled Uranium (RU), Mixed Oxide (MOX), and Thorium (Th). Recently, this unique CANDU reactor feature attracted considerable attention due to favourable commercial, environmental and strategic needs. This paper summarizes the solid progress over the last three years and outlines CANDU Energy Incorporated's (CEI) multi-stage vision of utilizing various fuels in currently operational and new build CANDU reactors. In CEI's fuel-cycle vision, CANDU reactors will operate in conjunction with other reactor types and use advanced fuels to produce more energy and ensure the most efficient and least costly method of utilizing Light Water Reactor (LWR) used fuel. With this vision and the tandem goal of systematic adoption of Thorium based fuels, CANDU reactors will be a strong technology partner in ensuring the availability of long-term stable resources for nuclear power plants

Research and materials irradiation reactors

International Nuclear Information System (INIS)

Ballagny, A.; Guigon, B.

2004-01-01

Devoted to the fundamental and applied research on materials irradiation, research reactors are nuclear installations where high neutrons flux are maintained. After a general presentation of the research reactors in the world and more specifically in France, this document presents the heavy water cooled reactors and the water cooled reactors. The third part explains the technical characteristics, thermal power, neutron flux, operating and details the Osiris, the RHF (high flux reactor), the Orphee and the Jules Horowitz reactors. The last part deals with the possible utilizations. (A.L.B.)

Water and Regolith Shielding for Surface Reactor Missions

Science.gov (United States)

Poston, David I.; Ade, Brian J.; Sadasivan, Pratap; Leichliter, Katrina J.; Dixon, David D.

2006-01-01

This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density.

Water and Regolith Shielding for Surface Reactor Missions

International Nuclear Information System (INIS)

Poston, David I.; Sadasivan, Pratap; Dixon, David D.; Ade, Brian J.; Leichliter, Katrina J.

2006-01-01

This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density

Thermohydraulic relationships for advanced water cooled reactors

International Nuclear Information System (INIS)

2001-04-01

This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

Thermohydraulic relationships for advanced water cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-04-01

This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

Survey of natural-circulation cooling in U.S. pressurized water reactors

International Nuclear Information System (INIS)

Boyack, B.E.

1985-01-01

Literature describing natural circulation analyses, experiments, and plant operation have been obtained from the Nuclear Regulatory Commission, reactor vendors, utility-sponsored research groups, utilities, national laboratories, and foreign sources. These have been reviewed and significant results and conclusions identified. Three modes of natural-circulation cooling are covered: single phase, two-phase, and reflux condensation. Single-phase natural circulation is amply verified by plant operational data, test data from scaled experimental facilities, and analysis with assessed computer codes. Ample evidence also exists that two-phase natural circulation can successfully cool pressurized water reactors. This mode occurs during certain events such as small-break loss-of-coolant accidents. The data base for reflux condensation is primarily from tests in scaled experimental facilities. There are no plant operational data and only limited assessment of thermal-hydraulic systems codes has been performed. Further work is needed before this mode of natural circulation can be confidently used

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)

An improved water cooled nuclear reactor and pressuriser assembly

International Nuclear Information System (INIS)

Gardner, F.J.; Strong, R.

1991-01-01

A water cooled nuclear reactor is described which comprises a reactor core, a primary water coolant circuit and a pressuriser arranged as an integral unit in a pressure vessel. The pressure vessel is divided into an upper and a lower chamber by a casing. The reactor core and primary water coolant circuit are arranged in the lower chamber and the pressuriser is arranged in the upper chamber. A plurality of spray pipes interconnect a steam space of the pressuriser with the downcomer of the primary water coolant circuit below a heat exchanger. A plurality of surge ports interconnect a water space of the pressuriser with the primary water coolant circuit. The surge ports have hydraulic diodes so that there is a low flow resistance for water from the water space of the pressuriser to the primary water coolant circuit and high flow resistance in the opposite direction. The spray pipes provide a desuperheating spray of cooled water into the pressuriser during positive volume surges of the primary water coolant. The pressuriser arrangement may also be applied to integral water cooled reactors with separate pressurisers and to dispersed pressurised water reactors. The surge ports also allow water to flow by gravity to the core in an emergency. (author)

Fuel cycle options for light water reactors in Germany

International Nuclear Information System (INIS)

Broecking, D.; Mester, W.

1999-01-01

In Germany 19 nuclear power plants with an electrical output of 22 GWe are in operation. Annually about 450 t of spent fuel are unloaded from the reactors. Currently most of the spent fuel elements are shipped to France and the United Kingdom for reprocessing according to contracts which have been signed since the late 70es. By the amendment of the Atomic Energy Act in 1994 the previous priority for reprocessing of spent nuclear fuel was substituted by a legal equivalency of the reprocessing and direct disposal option. As a consequence some utilities take into consideration the direct disposal of their spent fuel for economical reasons. The separated plutonium will be recycled as MOX fuel in light water reactors. About 30 tons of fissile plutonium will be available to German utilities for recycling by the year 2000. Twelve German reactors are already licensed for the use of MOX fuel, five others have applied for MOX use. Eight reactors are currently using MOX fuel or used it in the past. The spent fuel elements which shall be disposed of without reprocessing will be stored in two interim dry storage facilities at Gorleben and Ahaus. The storage capacities are 3800 and 4200 tHM, respectively. The Gorleben salt dome is currently investigated to prove its suitability as a repository for high level radioactive waste, either in a vitrified form or as conditioned spent fuel. The future development of the nuclear fuel cycle and radioactive waste management depends on the future role of nuclear energy in Germany. According to estimations of the German utilities no additional nuclear power plants are needed in the near future. Around the middle of the next decade it will have to be decided whether existing plants should be substituted by new ones. For the foreseeable time German utilities are interested in a highly flexible approach to the nuclear fuel cycle and waste management keeping open both spent fuel management options: the closed fuel cycle and direct disposal of

Utilization of particle fuels in different reactor concepts

International Nuclear Information System (INIS)

1983-04-01

To date, particle fuel is only used in high temperature reactors (HTR). In this reactor type the particles exist of oxide fuel with a diameter of about 0.5 mm and are surrounded by various coatings in order to safely enclose fission products and decrease the radioactive release into the primary circuit. However, it is felt that fuel based upon spherical particles could have some advantages compared with pellets both on fabrication and in-core behaviour in several reactor concepts. This fuel is now of general interest and there is a high level of research and development activity in some countries. In order to collect, organize additional information and summarize experience on utilization of particle fuels in different reactor concepts, a questionnaire was prepared by IAEA in 1980 and sent to Member States, which might be involved in relevant developments. This survey has been prepared by a group of consultants and is mainly based on the responses to the IAEA questionnaire

Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

International Nuclear Information System (INIS)

1978-05-01

An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

Critical Power Response to Power Oscillations in Boiling Water Reactors

International Nuclear Information System (INIS)

Farawila, Yousef M.; Pruitt, Douglas W.

2003-01-01

The response of the critical power ratio to boiling water reactor (BWR) power oscillations is essential to the methods and practice of mitigating the effects of unstable density waves. Previous methods for calculating generic critical power response utilized direct time-domain simulations of unstable reactors. In this paper, advances in understanding the nature of the BWR oscillations and critical power phenomena are combined to develop a new method for calculating the critical power response. As the constraint of the reactor state - being at or slightly beyond the instability threshold - is removed, the new method allows the calculation of sensitivities to different operation and design parameters separately, and thus allows tighter safety margins to be used. The sensitivity to flow rate and the resulting oscillation frequency change are given special attention to evaluate the extension of the oscillation 'detect-and-suppress' methods to internal pump plants where the flow rate at natural circulation and oscillation frequency are much lower than jet pump plants

Franco-German nuclear cooperation: from the 'common product' to the first European pressurized water reactor

International Nuclear Information System (INIS)

Vignon, D.

1999-01-01

It has now been 10 years since Framatome and Siemens decided to collaborate on the design and sales of an advanced nuclear power plant (NPP) model based on pressurized water reactor (PWR) technology. Originally called the 'common product', this model was renamed the European pressurized water reactor when Electricite de France (EDF) and the German electric utilities joined this cooperative development effort in 1992. Since the beginning, this cooperation has been formalized in the framework of an agreement that led to the founding of a joint and equally owned subsidiary, Nucler Power International (NPI), which is reponsible for leading the development of the new model and later handling its export sales

Pressurized water reactor simulator. Workshop material. 2. ed

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

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

Nuclear reactor in deep water

International Nuclear Information System (INIS)

Anon.

1980-01-01

Events during October 1980, when the Indian Point 2 nuclear reactor was flooded by almost 500 000 litres of water from the Hudson river, are traced and the jumble of human errors and equipment failures chronicled. Possible damage which could result from the reactor getting wet and from thermal shock are considered. (U.K.)

Heat insulation device for reactor pressure vessel in water

International Nuclear Information System (INIS)

Nakamura, Heiichiro; Tanaka, Yoshimi.

1993-01-01

Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

Utilization of light water reactors for plutonium incineration

International Nuclear Information System (INIS)

Galperin, A.

1995-01-01

In this work a potential of incineration of excess Pu in LWR's is investigated. In order to maintain the economic viability of the Pu incineration option it should be carried out by the existing power plants without additional investment for plant modifications. Design variations are reduced to the fuel cycle optimization, i.e. fuel composition may be varied to achieve optimal Pu destruction. Fuel mixtures considered in this work were based either on uranium or thorium fertile materials and Pu as a fissile component. The slightly enriched U fuel cycle for a typical pressurized water reactor was considered as a reference case. The Pu content of all fuels was adjusted to assure the identical cycle length and discharged burnup values. An equilibrium cycle was simulated by performing cluster burnup calculations. The material composition data for the whole core was estimated based on the core, fuel and cycle parameters. The annual production of Pu of a standard PWR with 1100 MWe output is about 298 kg. The same core completely loaded with the MOX fuel is estimated to consume 474 kg of Pu, mainly fissile isotopes. The MOX-239 fuel type (pure Pu-239) shows a potential toreduce the initial total Pu inventory by 220 kg/year and fissile Pu inventory by 420 kg/year. TMOX and TMOX-239 are based on Th-232 as a fertile component of the fuel, instead of U-238. The amount of Pu destroyed per year for both cases is significantly higher than that of U-based fuels. Especially impressive is the reduction in fissile Pu inventory: more than 900 kg/year. (author)

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)

Report on the 9th workshop on the innovative water reactor for flexible fuel cycle

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao

2006-07-01

The research on Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been performed in JAEA for the development of future innovative reactor. The workshop on FLWRs has been held every year since 1998 aiming at information exchange with other organizations such as universities, laboratories, utilities and vendors. The 9th workshop was held on March 1, 2006 under the joint auspices of JAEA and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 64 participants. The workshop began with presentation entitled 'Activities on Nuclear Science and Engineering Research and Collaboration with Industry in JAEA', followed by presentations entitled 'Progress of Research and Development on FLWR' and 'On Final Report of Feasibility Study (phase 2) on Commercialized FBR Cycle Systems'. Then two lectures followed: 'Core and Fuel Design on Super Light Water Reactor' by Tokyo University and 'Recent trends on the Development of Next Generation Nuclear Reactor' by Institute of Applied Energy. This report summarizes the lectures of the workshop. (author)

Advance reactor and fuel-cycle systems--potentials and limitations for United States utilities

International Nuclear Information System (INIS)

Zebroski, E.L.; Williams, R.F.

1979-01-01

This paper reviews the potential benefits and limitations of advance reactor and fuel-cycle systems for United States utilities. The results of the review of advanced technologies show that for the near and midterm, the only advance reactor and fuel-cycle system with significant potential for United States utilities is the current LWR, and evolutionary, not revolutionary, enhancements. For the long term, the liquid-metal breeder reactor continues to be the most promising advance nuclear option. The major factors leading to this conclusion are summarized

Reprocessing technology for present water reactor fuels

International Nuclear Information System (INIS)

McMurray, P.R.

1977-01-01

The basic Purex solvent extraction technology developed and applied in the U.S. in the 1950's provides a well-demonstrated and efficient process for recovering uranium and plutonium for fuel recycle and separating the wastes for further treatment and packaging. The technologies for confinement of radioactive effluents have been developed but have had limited utilization in the processing of commercial light water reactor fuels. Technologies for solidification and packaging of radioactive wastes have not yet been demonstrated but significant experience has been gained in laboratory and engineering scale experiments with simulated commercial reprocessing wastes and intermediate level wastes. Commercial scale experience with combined operations of all the required processes and equipment are needed to demonstrate reliable reprocessing centers

Operator Support System for Pressurized Water Reactor

International Nuclear Information System (INIS)

Wei Renjie; Shen Shifei

1996-01-01

Operator Support System for Pressurized Water Reactor (OSSPWR) has been developed under the sponsorship of IAEA from August 1994. The project is being carried out by the Department of Engineering Physics, Tsinghua University, Beijing, China. The Design concepts of the operator support functions have been established. The prototype systems of OSSPWR has been developed as well. The primary goal of the project is to create an advanced operator support system by applying new technologies such as artificial intelligence (AI) techniques, advanced communication technologies, etc. Recently, the advanced man-machine interface for nuclear power plant operators has been developed. It is connected to the modern computer systems and utilizes new high performance graphic displays. (author). 6 refs, 4 figs

Proceedings of the 1997 workshop on the utilization of research reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-10-01

The 1997 Workshop on the Utilization of Research Reactors, which is the sixth Workshop on the theme of research reactor utilization was held in Bandung, Indonesia from November 6 to 13. This Workshop was executed based on the agreement in the Eighth International conference for Nuclear Cooperation in Asia (ICNCA) held in Tokyo, March 1997. The whole Workshop consists of the preceding Sub-workshop carried out the demonstration experiment of Radioisotope Production, and the Workshop on the theme of three fields (Neutron Scattering, Radioisotope production, Safe Operation and Maintenance of Research Reactor). The total number of participants for the workshop was about 100 people from 8 countries, i.e. China, Indonesia, Korea, Malaysia, Philippine, Thailand, Vietnam and Japan. It consists of the papers for Sub-workshop, Neutron Scattering, Radioisotope Production, Safe Operation and Maintenance of research reactor, and summary reports. The 53 of the presented papers are indexed individually. (J.P.N.)

Safety aspects of pressurised water reactors

International Nuclear Information System (INIS)

1985-01-01

This submission to the Health and Safety Executive has been prepared by the Institution of Professional Civil Servants (IPCS) as a contribution to the debate on safety aspects associated with Pressurized Water Reactors (PWRs). Although supporting an energy policy which includes the development of nuclear power, assurances are sought on a number of safety issues if it is decided that this should be generated by a PWR-type reactor. These issues are listed. In particular the following are mentioned: the wider publication of design information, the use of elastic-plastic fracture mechanics as the basis for determining pressure vessel integrity, the failure rate of steam generating units, water coolant quality control, greater investigation of two-phase flow accident conditions, the components of the reactor cooling system and training of reactor personnel in the understanding of LOCA effects. (U.K.)

Device for controlling water supply to nuclear reactor

International Nuclear Information System (INIS)

Iwasaki, Toshio.

1974-01-01

Object: To smoothly control automatic water supply for realizing stable operation of a nuclear reactor by providing a flow rate limiting signal selection circuit and a preferential circuit in a water supply control device for a nuclear reactor wherein the speed of a recirculation pump may be changed in two-steps. Structure: Opening angle signals for a water supply regulating valve are controlled by a nuclear reactor water level signal, a vapor flow rate signal and a supplied water flow rate signal through an adder and an adjuster in response to a predetermined water level setting signal. When the water in the reactor is maintained at a predetermined level, a selection circuit receives a water pump condition signal for selecting one of the signals from a supplied water rate limiting signal generator generating signals for indicating whether one or two water supply pumps are operated. A low value preferential circuit passes the lower of the values generated from the selection circuit and the adder. The selection circuit receives a recirculation pump condition signal and selects either one of the signals from the supplied water flow rate limiting signal generator operated at high speed or low speed. A high value preferential circuit passes the higher value

Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.

1989-01-01

Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

Research reactor utilization. Summary reports of three study group meetings: Irradiation techniques at research reactors, held in Istanbul 15-19 November 1965; Research reactor operation and maintenance problems, held in Caracas 6-10 December 1965; and Research reactor utilization in the Far East, held in Lucas Heights 28 February - 4 March 1966

International Nuclear Information System (INIS)

1967-01-01

The three sections of this book, which are summary reports of three Study Group meetings of the IAEA: Irradiation techniques at research reactors, Istanbul, 15-19 November 1965; Research reactor operation and maintenance problems, Caracas, 6-10 December 1965; and Research reactor utilization in the Far East, Lucas Heights, Australia, 28 February - 4 March 1966. These meetings were the latest in a series designed to promote efficient utilization of research reactors, to disseminate information on advances in techniques, to discuss common problems in reactor operations, and to outline some advanced areas of reactor-based research. (author)

Auxiliary equipment for cooling water in a reactor

International Nuclear Information System (INIS)

Konno, Yasuhiro; Sakairi, Toshiaki.

1975-01-01

Object: To effectively make use of pressure energy of reactor water, which has heretofore been discarded, to enable supply of emergency power supply of high reliability and to prevent spreading of environmental contamination. Structure: Sea water pumped by a sea water supply pump is fed to a heat exchanger. Reactor water carried through piping on the side to be cooled is removed in heat by the heat exchanger to be cooled and returned, and then again returned to the reactor. On the other hand, sea water heated by the heat exchanger is fed to a water wheel to drive the water wheel, after which it is discharged into a discharging path. A generator may be directly connected to the water wheel to use the electricity generated by the generator as the emergency power source. (Kamimura, M.)

Nuclear fuel, with emphasis on its utilization in pressurized water reactor

International Nuclear Information System (INIS)

Khazaneh, R.; Roshanzamir, M.

1997-01-01

Production processes of nuclear fuel on one hand and using nuclear fuels in reactors, particularly PWR Type reactors on the other hand is investigated. The first chapter reviews the relationship between fuel and reactors; The principals of reactor physics in relation with fuel are described shortly. The second chapter reviews uranium exploration and extraction as well as production of uranium concentrate and uranium dioxides. The third chapter is specified to the different procedures of uranium enrichment. In the fourth chapter, processing of uranium dioxide powder and fuel pellet is described. In the fifth chapter fabrication of fuel rod and fuel assemblies is explained thoroughly. The sixth chapter devoted to the different phenomena which occur ed in fuel structure and can during operational time of reactor; damage to fuel rods and developing theoretical models to describe these phenomena and analysis of fuel structure. The seventh chapter discusses how fuel rods are to be experimented during fabrication, operation and development of technology. The eighth chapter explains different fuels such as uranium compounds and mixed oxide fuel of uranium Gadolinium and uranium plutonium and the process of fabrication of zircaloy. In the tenth chapter, fuel reprocessing is investigated and the difficulties of developing this technology is referred

Utilization of fission reactors for fusion engineering testing

International Nuclear Information System (INIS)

Deis, G.A.; Miller, L.G.

1985-01-01

Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful

Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-06-01

Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO₂) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

Conceptual designing of reduced-moderation water reactor with heavy water coolant

Energy Technology Data Exchange (ETDEWEB)

Hibi, Kohki; Shimada, Shoichiro; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi; Wada, Shigeyuki

2001-12-01

The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06-1.11.

Thermodynamic analysis of the use a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water-splitting cycle for hydrogen production

International Nuclear Information System (INIS)

Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

2008-01-01

Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved 'steam' parameters (outlet temperatures up to 625degC and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among 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 an intensive heat supply at temperatures higher than 550-600degC. 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 of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, 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 the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the 'nuclear' heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of

Channel type reactors with supercritical water coolant. Russian experience

International Nuclear Information System (INIS)

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

2003-01-01

Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

Proceedings of the 1998 workshop on the utilization of research reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-10-01

The 1998 Workshop on the Utilization of Research Reactors, which is the seventh Workshop on the theme of research reactor utilization was held in Yogyakarta and Serpong, Indonesia from February 8 to 14. This Workshop was executed based on the agreement in the Ninth International Conference for Nuclear Cooperation in Asia (ICNCA) held in Tokyo, March 1998. The whole Workshop consists of the Workshop on the theme of following three fields, 1) Neutron Scattering, 2) Neutron Activation analysis and 3) Safe Operation and Maintenance of Research Reactor, and the Sub-workshop carried out the experiment of Neutron Activation analysis. The total number of participants for the workshop was about 100 people from 8 countries, i.e. Australia, China, Indonesia, Korea, Malaysia, Thailand, Vietnam and Japan. The 38 papers are indexed individually. (J.P.N.)

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)

Current utilization of research reactor on radioisotopes production in China

International Nuclear Information System (INIS)

Liu Yishu

2000-01-01

The main technical parameters of the four research reactors and their current utilization status in radioisotope manufacture and labeling compounds preparation are described. The radioisotopes, such as Co-60 sealed source, Ir-192 sealed source, γ-knife source, I-131, I-125, Sm-153, P-32 series products, In-113m generator, Tc-99m gel generator, Re-188 gel generator, C-14, Ba-131, Sr-89, 90 Y, etc., and their labeling compounds prepared from the reactor produced radionuclides, such as I-131-MIBG, I-131-Hippure, I-131-capsul, Sm-153-EDTMP, Re-186-HEDP, Re-186-HA, C-14-urea, and radioimmunoassay kits etc. are presented as well. Future development plan of radioisotopes and labeling compounds in China is also given. Simultaneously, the possibility and methods of bilateral or multilateral co-operation in utilization of research reactor, personnel and technology exchange of radioisotope production and labeling compounds is also discussed. (author)

Current utilization of research reactor on radioisotopes production in China

Energy Technology Data Exchange (ETDEWEB)

Liu Yishu [Nuclear Power Institute of China, Chengdu (China)

2000-10-01

The main technical parameters of the four research reactors and their current utilization status in radioisotope manufacture and labeling compounds preparation are described. The radioisotopes, such as Co-60 sealed source, Ir-192 sealed source, {gamma}-knife source, I-131, I-125, Sm-153, P-32 series products, In-113m generator, Tc-99m gel generator, Re-188 gel generator, C-14, Ba-131, Sr-89, {sup 90}Y, etc., and their labeling compounds prepared from the reactor produced radionuclides, such as I-131-MIBG, I-131-Hippure, I-131-capsul, Sm-153-EDTMP, Re-186-HEDP, Re-186-HA, C-14-urea, and radioimmunoassay kits etc. are presented as well. Future development plan of radioisotopes and labeling compounds in China is also given. Simultaneously, the possibility and methods of bilateral or multilateral co-operation in utilization of research reactor, personnel and technology exchange of radioisotope production and labeling compounds is also discussed. (author)

Reactor vessel pressure transient protection for pressurized water reactors

International Nuclear Information System (INIS)

Zech, G.

1978-09-01

During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

Integral Inherently Safe Light Water Reactor (I2S-LWR)

International Nuclear Information System (INIS)

Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

2017-01-01

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

Expert system for control rod programming of boiling water reactors

International Nuclear Information System (INIS)

Fukuzaki, T.; Yoshida, K.; Kobayashi, Y.; Matsuura, H.; Hoshi, K.

1986-01-01

Control rod programming, one of the main tasks in reactor core management of boiling water reactors (BWRs), can be successfully accomplished by well-experienced engineers. By use of core performance evaluation codes, their knowledge plays the main role in searching through optimal control rod patterns and exposure points for adjusting notch positions and exchanging rod patterns. An expert system has been developed, based on a method of knowledge engineering, to lighten the engineer's load in control rod programming. This system utilizes an inference engine suited for planning/designing problems, and stores the knowledge of well-experienced engineers in its knowledge base. In this report, the inference engine, developed considering the characteristics of the control rod programming, is introduced. Then the constitution and function of the expert system are discussed

Utility requirements for safety in the passive advanced light-water reactor

International Nuclear Information System (INIS)

Marston, T.U.; Layman, W.H.; Bockhold, G. Jr.

1993-01-01

The objective of the passive plant design is to use passive systems to replace all the active engineered safety systems presently used in light-water reactors. The benefits derived from such an approach to safety design are multiple. First, it is expected that a passive design approach will significantly simplify the overall plant design, including a reduction in the number of components, and reduce the operation and maintenance burden. Second, it is expected that the overall safety and reliability of the passive systems will be improved over active systems, which will result in extremely low risk to public health and safety. Third, challenges to the operating staff will be minimized during transient and emergency conditions, which will reduce the uncertainty associated with human behavior. Finally, it is expected that reliance on passive safety features will lead to a better understanding by the general public and recognition that a major improvement in public safety has been achieved

Results of Operation and Utilization of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Le Vinh Vinh; Duong Van Dong; Nguyen Xuan Hai; Pham Ngoc Son; Cao Dong Vu

2014-01-01

The Dalat Nuclear Research Reactor (DNRR) with the nominal power of 500 kW was reconstructed and upgraded from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The renovated reactor was put into operation on 20 March 1984. It was designed for the purposes of radioisotope production (RI), neutron activation analysis (NAA), basic and applied researches, and nuclear education and training. During the last 30 years of operation, the DNRR was efficiently utilized for producing many kinds of radioisotopes and radiopharmaceuticals used in nuclear medicine centers and other users in industry, agriculture, hydrology and scientific research; developing a combination of nuclear analysis techniques (INAA, RNAA, PGNAA) and physic-chemical methods for quantitative analysis of about 70 elements and constituents in various samples; carrying out experiments on the reactor horizontal beam tubes for nuclear data measurement, neutron radiography and nuclear structure study; and establishing nuclear training and education programs for human resource development. This paper presents the results of operation and utilization of the DNRR. In addition, some main reactor renovation projects carried out during the last 10 years are also mentioned in the paper. (author)

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

Thorium utilization: conversion ratio and fuel needs in thermal reactors

International Nuclear Information System (INIS)

Oosterkamp, W.J.

1975-01-01

As a preparatory study for thorium utilization in thermal reactors a study has been made of the fuel comsumption in existing reactor types. A quantitative description is given of the influence of enrichment, burnup, amount of structural material, choise of coolant and control requirements on the convertion ratio. The enrichment is an important factor and a low fuel comsumption can be achieved by increasing the enrichment

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

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)

Heat exchangers in heavy water reactor systems

International Nuclear Information System (INIS)

Mehta, S.K.

1988-01-01

Important features of some major heat exchange components of pressurized heavy water reactors and DHRUVA research reactor are presented. Design considerations and nuclear service classifications are discussed

Fundamentals of pressurized water reactors

International Nuclear Information System (INIS)

Murray, L.

1982-01-01

In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

Research and development of super light water reactors and super fast reactors in Japan

International Nuclear Information System (INIS)

Oka, Y.; Morooka, S.; Yamakawa, M.; Ishiwatari, Y.; Ikejiri, S.; Katsumura, Y.; Muroya, Y.; Terai, T.; Sasaki, K.; Mori, H.; Hamamoto, Y.; Okumura, K.; Kugo, T.; Nakatsuka, T.; Ezato, K.; Akasaka, N.; Hotta, A.

2011-01-01

Super Light Water Reactors (Super LWR) and Super Fast Reactors (Super FR) are the supercritical- pressure light water cooled reactors (SCWR) that are developed by the research group of University of Tokyo since 1989 and now jointly under development with the researchers of Waseda University, University of Tokyo and other organizations in Japan. The principle of the reactor concept development, the results of the past Super LWR and Super FR R&D as well as the R&D program of the Super FR second phase project are described. (author)

Water-chemical regime of a fast reactor ower complex

International Nuclear Information System (INIS)

Musikhin, R.N.; Piskunov, E.M.; Samarkin, A.A.; Yurchenko, D.S.

1983-01-01

Some peculiarities of water-chemical regime of a power compleX in Shevchenko are considered. The complex comprises a desalination unit, a gas-masout heating-and-power plant and the BN-350 reactor. The compleX is used for the production of electric and thermal energy and fresh water. The power complex peculiarity is the utilization of disalinated seawater in a technological cycle along with highly mineralized seawater with a total salt content of 13.5 g/l (for cooling) in heat exchanges. A regime of ammoniacal correction of feed water was used as a basic water-chemical regime in the initial period of the BN-350 steam generator operation. Deposits composed mainly of iron oxide slime were observed on steam generator surfaces during the operation under these conditions. A conclusion is made that the regime with chelating agent providing steam generator safe operation without chemical cleaning is the most expedient one

Advances in heavy water reactors

International Nuclear Information System (INIS)

1994-03-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

Membrane steam reforming of natural gas for hydrogen production by utilization of medium temperature nuclear reactor

International Nuclear Information System (INIS)

Djati Hoesen Salimy

2010-01-01

The assessment of steam reforming process with membrane reactor for hydrogen production by utilizing of medium temperature nuclear reactor has been carried out. Difference with the conventional process of natural gas steam reforming that operates at high temperature (800-1000°C), the process with membrane reactor operates at lower temperature (~500°C). This condition is possible because the use of perm-selective membrane that separate product simultaneously in reactor, drive the optimum conversion at the lower temperature. Besides that, membrane reactor also acts the role of separation unit, so the plant will be more compact. From the point of nuclear heat utilization, the low temperature of process opens the chance of medium temperature nuclear reactor utilization as heat source. Couple the medium temperature nuclear reactor with the process give the advantage from the point of saving fossil fuel that give direct implication of decreasing green house gas emission. (author)

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.

Pakistan research reactor and its utilization

International Nuclear Information System (INIS)

Iqbal Hussain Qureshi; Naeem Ahmad Khan.

1983-01-01

The 5 MW enriched uranium fuelled, light water moderated and cooled Pakistan Research reactor became critical on 21st December, 1965 and was taken to full power on 22nd June, 1966. Since then is has been operated for about 23000 hours till 30th June, 1983 without any major break down. It has been used for the studies of neutron cross-sections, nuclear structure, fission physics, structure of material, radiation damage in crystals and semiconductors, studies of geological, biological and environmental samples by neutron activation techniques, radioisotope production, neutron radiography and for training of scientists, engineers and technicians. In the paper we have described briefly the facility of Pakistan Research Reactor and the major work carried around it during the last decade. (author)

Review of light--water reactor safety studies. Volume 3 of health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

International Nuclear Information System (INIS)

Nero, A.V.; Farnaam, M.R.K.

1977-01-01

This report summarizes and compares important studies of light-water nuclear reactor safety, emphasizing the Nuclear Regulatory Commission's Reactor Safety Study, work on risk assessment funded by the Electric Power Research Institute, and the Report of the American Physical Society study group on light-water reactor safety. These reports treat risk assessment for nuclear power plants and provide an introduction to the basic issues in reactor safety and the needs of the reactor safety research program. Earlier studies are treated more briefly. The report includes comments on the Reactor Safety Study. The manner in which these studies may be used and alterations which would increase their utility are discussed

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)

Proceedings of the 1999 workshop on the utilization of research reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-10-01

The 1999 workshop on the utilization of reactors, which is the eighth workshop on the theme of research reactor utilization was held at JAERI Tokai and Mito Plaza Hotel, in Japan from November 25 to December 2. This workshop was executed based on the agreement in the Tenth International conference for Nuclear Cooperation in Asia (ICNCA) held in Tokyo, March 1999. The whole workshop consists of the workshop on the theme of following three fields, 1) neutron scattering, 2) radioisotope production and 3) safe operation and maintenance of research reactor, and the sub-workshop carried out the experiments of small angle neutron scattering. The total number of participants for the workshop was about 70 people from 9 countries, i.e. Australia, China, Indonesia, Korea, Malaysia, The Philippines, Thailand, Vietnam and Japan. The 37 of the presented papers are indexed individually. (J.P.N.)

Plutonium and minor actinides utilization in Thorium molten salt reactor

International Nuclear Information System (INIS)

Waris, Abdul; Aji, Indarta K.; Novitrian,; Kurniadi, Rizal; Su'ud, Zaki

2012-01-01

FUJI-12 reactor is one of MSR systems that proposed by Japan. The original FUJI-12 design considers Th/ 233 U or Th/Pu as main fuel. In accordance with the currently suggestion to stay away from the separation of Pu and minor actinides (MA), in this study we evaluated the utilization of Pu and MA in FUJI-12. The reactor grade Pu was employed in the present study as a small effort of supporting THORIMS-NES scenario. The result shows that the reactor can achieve its criticality with the Pu and MA composition in the fuel of 5.96% or more.

Safety reviews of next-generation light-water reactors

International Nuclear Information System (INIS)

Kudrick, J.A.; Wilson, J.N.

1997-01-01

The Nuclear Regulatory Commission (NRC) is reviewing three applications for design certification under its new licensing process. The U.S. Advanced Boiling Water Reactor (ABWR) and System 80+ designs have received final design approvals. The AP600 design review is continuing. The goals of design certification are to achieve early resolution of safety issues and to provide a more stable and predictable licensing process. NRC also reviewed the Utility Requirements Document (URD) of the Electric Power Research Institute (EPRI) and determined that its guidance does not conflict with NRC requirements. This review led to the identification and resolution of many generic safety issues. The NRC determined that next-generation reactor designs should achieve a higher level of safety for selected technical and severe accident issues. Accordingly, NRC developed new review standards for these designs based on (1) operating experience, including the accident at Three Mile Island, Unit 2; (2) the results of probabilistic risk assessments of current and next-generation reactor designs; (3) early efforts on severe accident rulemaking; and (4) research conducted to address previously identified generic safety issues. The additional standards were used during the individual design reviews and the resolutions are documented in the design certification rules. 12 refs

Economics and utilization of thorium in nuclear reactors

International Nuclear Information System (INIS)

1978-05-01

Information on thorium utilization in power reactors is presented concerning the potential demand for nuclear power, the potential supply for nuclear power, economic performance of thorium under different recycle policies, ease of commercialization of the economically preferred cases, policy options to overcome institutional barriers, and policy options to overcome technological and regulatory barriers

The Utilization of Dalat nuclear research reactor for education and training purposes

International Nuclear Information System (INIS)

Luong, Ba Vien; Nguyen, Nhi Dien; Le, Vinh Vinh; Nguyen, Xuan Hai

2017-01-01

The Dalat Nuclear Research Reactor (DNRR) with the nominal power of 500 kWt is today the unique one in Vietnam. It was designed for the purposes of radioisotope production, neutron activation analysis, basic and applied researches, and nuclear education and training. With the rising demand in development of human resources for utilization of atomic energy in the country, the DNRR has been playing an important role in the nuclear education and training for students from universities and professionals who are interested in reactor engineering. At present, the Dalat Nuclear Research Institute (DNRI) offers two types of training course utilizing the research reactor: an one-week practical training course is applied for undergraduate students and a two-week training course on reactor engineering is applied for the professionals. This paper presents the reactor facility and experiments performed at the DNRR for education and training purposes. In addition, the co-operation between the DNRI with national and international educational organizations for nuclear human resource development for national and regional demands is also mentioned in the paper. (author)

Light water reactor safety research project

International Nuclear Information System (INIS)

Markoczy, G.; Aksan, S.N.; Behringer, K.; Prodan, M.; Stierli, F.; Ullrich, G.

1980-07-01

The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

Hydriding failure in water reactor fuel elements

International Nuclear Information System (INIS)

Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

1980-01-01

Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

Environmentally assisted cracking in light water reactors

International Nuclear Information System (INIS)

Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Soppet, W.K.; Shack, W.J.

1992-03-01

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with ∼ 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289 degrees C

Thermal-hydraulic R and D infrastructure for water cooled reactors of the Indian nuclear power program

International Nuclear Information System (INIS)

Vijayan, P.K.; Jain, V.; Saha, D.; Sinha, R.K.

2009-01-01

R and D has been the critical ingredient of Indian Nuclear Power Program from the very inception. Approach to R and D infrastructure has been closely associated with the three-stage nuclear power program that was crafted on the basis of available resources and technology in the short-term and energy security in the long-term. Early R and D efforts were directed at technologies relevant to Pressurized Heavy Water Reactors (PHWRs) which are currently the mainstay of Indian nuclear power program. Lately, the R and D program has been steered towards the design and development of advanced and innovative reactors with the twin objective of utilization of abundant thorium and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy, proliferation resistance etc. Advanced Heavy Water Reactor (AHWR) is an Indian innovative reactor currently being developed to realize the above objectives. Extensive R and D infrastructure has been created to validate the system design and various passive concepts being incorporated in the AHWR. This paper provides a brief review of R and D infrastructure that has been developed at Bhabha Atomic Research Centre for thermal-hydraulic investigations for water-cooled reactors of Indian nuclear power program. (author)

Special issue on ageing management and long term operation of light water reactors

International Nuclear Information System (INIS)

Kashima, Koichi; Kanno, Masanori; Kimura, Itsurou

2008-01-01

Ageing management and long term operation of light water reactors became important in Japan and relevant research on physical and materials ageing has been carried out among organizations of government, academia and industry and establishment of technical standards and guideline based on the results is under way. The Japan Energy Policy Institute (JEPI) issued a special number discussing this theme, which consisted of ten reports of experts describing these activities. Main topics were ageing evaluation of reactor components due to neutron irradiation embrittlement and stress corrosion cracking, regulatory evaluation of deteriorations due to ageing with technical information basis, technology development in the area of inspection/monitoring, ageing evaluation and preventive maintenance/repairs, nuclear power plant life management of electric utilities, and advancement of reactor maintenance and inspection. (T. Tanaka)

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

Development of database system on MOX fuel for water reactors (I)

International Nuclear Information System (INIS)

Kikuchi, Keiichi; Nakazawa, Hiroaki; Abe, Tomoyuki; Shirai, Takao

2000-04-01

JNC has been conducted a great number of irradiation tests to develop MOX fuels for Advanced Thermal Reactor and Light Water Reactors. In order to manage irradiation data consistently and to effectively utilize valuable data obtained from the irradiation tests, we commenced construction of database system on MOX fuel for water reactors in 1998 JFY. Collection and selection of irradiation data and relevant fuel fabrication data, design of the database system and preparation of assisting programs have been finished and data registration onto the system is under way according to priority at present. The database system can be operated through the menu screen on PC. About 94,000 records of data on 11 fuel assemblies in total have been registered onto the database up to the present. By conducting registration of the remaining data and some modification of the system, if necessary, the database system is expected to complete in 2000 JFY. The completed database system is to be distributed to relevant sections in JNC by means of CD-R as a media. This report is an interim report covering 1998 and 1999 JFY, which gives the structure explanation and users manual concerning to the prepared database up to the present. (author)

Status of advanced light water reactor designs 2004

International Nuclear Information System (INIS)

2004-05-01

The report is intended to be a source of reference information for interested organizations and individuals. Among them are decision makers of countries considering implementation of nuclear power programmes. Further, the report is addressed to government officials with an appropriate technical background and to research institutes of countries with existing nuclear programmes that wish to be informed on the global status in order to plan their nuclear power programmes including both research and development efforts and means for meeting future. The future utilization of nuclear power worldwide depends primarily on the ability of the nuclear community to further improve the economic competitiveness of nuclear power plants while meeting stringent safety requirements. The IAEA's activities in nuclear power technology development include the preparation of status reports on advanced reactor designs to provide all interested IAEA Member States with balanced and objective information on advances in nuclear plant technology. In the field of light water reactors, the last status report published by the IAEA was 'Status of Advanced Light Water Cooled Reactor Designs: 1996' (IAEA-TECDOC-968). Since its publication, quite a lot has happened: some designs have been taken into commercial operation, others have achieved significant steps toward becoming commercial products, including certification from regulatory authorities, some are in a design optimization phase to reduce capital costs, development for other designs began after 1996, and a few designs are no longer pursued by their promoters. With this general progress in mind, on the advice and with the support of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for Light Water Reactors (LWRs), the IAEA has prepared this new status report on advanced LWR designs that updates IAEA-TECDOC-968, presenting the various advanced LWR designs in a balanced way according to a common outline

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.)

Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community

International Nuclear Information System (INIS)

Lybeck, Nancy J.; Tawfik, Magdy S.; Pham, Binh T.

2011-01-01

Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.

The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

International Nuclear Information System (INIS)

Brand, P.; Wiechers, W.K.

1974-01-01

The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

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)

Transmutation of Americium in Light and Heavy Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Hyland, B.; Dyck, G.R.; Edwards, G.W.R. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada); Ellis, R.J.; Gehin, J.C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee (United States); Maldonado, G.I. [University of Tennessee (Knoxville)/ORNL, Tennessee (United States)

2009-06-15

There is interest worldwide in reducing the burden on geological nuclear fuel disposal sites. In most disposal scenarios the decay heat loading of the surrounding rock limits the capacity of these sites. On the long term, this decay heat is generated primarily by actinides, and a major contributor 100 to 1000 years after discharge from the reactor is {sup 241}Am. One possible approach to reducing the decay-heat burden is to reprocess spent reactor fuel and use thermal spectrum reactors to 'burn' the Am nuclides. The viability of this approach is dependent upon the detailed changes in chemical and isotopic composition of actinide-bearing fuels after irradiation in thermal reactor spectra. The currently available thermal spectrum reactor options include light water-reactors (LWRs) and heavy-water reactors (HWRs) such as the CANDU{sup R} designs. In addition, as a result of the recycle of spent LWR fuel, there would be a considerable amount of potential recycled uranium (RU). One proposed solution for the recycled uranium is to use it as fuel in Candu reactors. This paper investigates the possibilities of transmuting americium in 'spiked' bundles in pressurized water reactors (PWRs) and in boiling water reactors (BWRs). Transmutation of Am in Candu reactors is also examined. One scenario studies a full core fuelled with homogeneous bundles of Am mixed with recycled uranium, while a second scenario places Am in an inert matrix in target channels in a Candu reactor, with the rest of the reactor fuelled with RU. A comparison of the transmutation in LWRs and HWRs is made, in terms of the fraction of Am that is transmuted and the impact on the decay heat of the spent nuclear fuel. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). (authors)

Water quality control device and water quality control method for reactor primary coolant system

International Nuclear Information System (INIS)

Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

1995-01-01

The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

Study of reverse osmosis applicability to light water reactor radwaste processing

International Nuclear Information System (INIS)

Markind, J.; Van Tran, T.

1978-12-01

Objectives were to collect and evaluate documented performance data of existing reverse-osmosis/ultrafiltration processes utilized for treating low-level liquid radioactive wastes, originating from light-water-reactor (LWR) nuclear power plants. Relevant information was collected by communication both written and verbal with membrane experts known to be active in the nuclear industry, and by conducting manual and computer searches. The generated information was evaluated on the basis of membrane performance characteristics relevant to nuclear engineering system analysis. 39 figures, 34 tables

Pressurized-water reactors

International Nuclear Information System (INIS)

Bush, S.H.

1983-03-01

An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

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)

High conversion heavy water moderated reactor

International Nuclear Information System (INIS)

Miyawaki, Yoshio; Wakabayashi, Toshio.

1989-01-01

In the present invention, fuel rods using uranium-plutonium oxide mixture fuels are arranged in a square lattice at the same pitch as that in light water cooled reactor and heavy water moderators are used. Accordingly, the volume ratio (Vm/Vf) between the moderator and the fuel can be, for example, of about 2. When heavy water is used for the moderator (coolant), since the moderating effect of heavy water is lower than that of light water, a high conversion ratio of not less than 0.8 can be obtained even if the fuel rod arrangement is equal to that of PWR (Vm/Vf about 2). Accordingly, it is possible to avoid problems caused by dense arrangement of fuel rods as in high conversion rate light water cooled reactors. That is, there are no more troubles in view of thermal hydrodynamic characteristics, re-flooding upon loss of coolant accident, etc., as well as the fuel production cost is not increased. (K.M.)

Flow analysis in a supercritical water oxidation reactor

International Nuclear Information System (INIS)

Oh, C.H.; Kochan, R.J.; Beller, J.M.

1996-01-01

Supercritical water oxidation (SCWO), also known as hydrothermal oxidation (HTO), involves the oxidation of hazardous waste at conditions of elevated temperature and pressure (e.g., 500 C--600 C and 234.4 bar) in the presence of approximately 90% of water and a 10% to 20% excess amount of oxidant over the stoichiometric requirement. Under these conditions, organic compounds are completely miscible with supercritical water, oxygen and nitrogen, and are rapidly oxidized to carbon dioxide and water. The essential part of the process is the reactor. Many reactor designs such as tubular, vertical vessel, and transpiring wall type have been proposed, patented, and tested at both bench and pilot scales. These designs and performances need to be scaled up to a waste throughput 10--100 times that currently being tested. Scaling of this magnitude will be done by creating a numerical thermal-hydraulic model of the smaller reactor for which test data is available, validating the model against the available data, and then using the validated model to investigate the larger reactor performance. This paper presents a flow analysis of the MODAR bench scale reactor (vertical vessel type). These results will help in the design of the reactor in an efficient manner because the flow mixing coupled with chemical kinetics eventually affects the process destruction efficiency

Outline of the advanced boiling water reactor (ABWR)

International Nuclear Information System (INIS)

Hucik, S.A.; Imaoka, T.; Minematsu, A.; Takashima, Y.

1986-01-01

The fundamental design of the Advanced Boiling Water Reactor (ABWR) was completed in December 1985. This design represents the next generation of Boiling Water Reactors (BWR) to be introduced into commercial operation in the 1990s. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technologies and many new improvements based on an extensive accumulation of world-wide experience through design, construction and operation of BWRs. The ABWR development program was initiated in 1978, with subsequent design and test and development programs started in 1981. Most of the development and verification tests of the new features have been completed. The ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. The ABWR objectives were specific improvements such as operating and safety margins, enhanced availability and capacity factor, and reduced occupational exposure while at the same time achieving significant cost reduction in both capital and operating costs. The ABWR is characterized by an improved NSSS including ten internal recirculation pumps, fine motion electric-hydraulic control rod drives, optimized safety and auxiliary systems, advanced control and instrumentation systems, improved turbine-generator with moisture/separator reheater with plant output increased to 1350 MWe, and an integrated reinforced concrete containment vessel and compact Reactor and Turbine Building design. The turbine system also included improvements in the Turbine-Generator, feedwater/heater system, and condensate treatment systems. The radwaste system was also optimized taking advantage of the plant design improvements and advances in radwaste technology. The ABWR is a truly optimal design which utilizes advanced technologies, capabilities, performance improvements, and yet provides an economic advantage. (author)

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)

Screening reactor steam/water piping systems for water hammer

International Nuclear Information System (INIS)

Griffith, P.

1997-09-01

A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made

Localized corrosion problems in water reactors

International Nuclear Information System (INIS)

Coriou, Henri.

1977-01-01

Main localized etching on the structure materials of water reactors are studied: stress corrosion on stainless steel 304 (B.W.R), stress corrosion, 'wall thinning' and denting of Inconel 600 vapor generator tubes (P.W.R.). Some mechanisms are examined and practical exemples in reactors are described. Various possible cures are presented [fr

Brief overview on utilization of research reactors in China

International Nuclear Information System (INIS)

Yuan Luzheng

2003-01-01

The utilization on research reactors in China, such as neutron scattering study, neutron activation analysis and it applications in wide fields, radioisotope production including fission Mo-Tc generators and non-fission Tc-99m generator, BNCT study and many others, is briefly overviewed. (author)

Brief overview on utilization of research reactors in China

Energy Technology Data Exchange (ETDEWEB)

Yuan Luzheng [China Institute of Atomic Energy, Beijing (China)

2003-03-01

The utilization on research reactors in China, such as neutron scattering study, neutron activation analysis and it applications in wide fields, radioisotope production including fission Mo-Tc generators and non-fission Tc-99m generator, BNCT study and many others, is briefly overviewed. (author)

A survey of thorium utilization in thermal power reactors

International Nuclear Information System (INIS)

Oosterkamp, W.J.

1974-01-01

The present status of thorium utilization in thermal reactors HTGR's, HWR's and LWR's has been reviewed. Physics considerations are made to obtain the optimum use of thorium. Existing information on reprocessing and refabrication is given together with the properties of thorium metal and thoria

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

Fuel performance in the Barsebeck boiling water reactors (Unit 1 and 2)

International Nuclear Information System (INIS)

Norman, B.

1979-01-01

Sydkraft is the largest privately owned utility in Sweden. It serves about 20% of the Swedish population with about 12 TWh of electric power per year, of which 64% is nuclear (1978 figures). The two identical 590 MWE ASEA-ATOM boiling water reactors in Barsebeck have been in operation since 1975 and 1977 respectively. Fission product activity in the primary circuits and in the off-gas systems is extremely low and indicate a near perfect fuel condition. Operating restrictions limiting the effect of pellet cladding interaction have been in use since initial start-up and testing. A few events involving rapid power increases above the preconditioned power level have occurred without causing fuel failures. It is believed that an analysis of power reactor operational transients, which did not cause fuel failures, can be useful to design more adequate and less conservative rules for the operation of nuclear reactor cores

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Development of design technology for advanced pressurized water reactor