Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

International Nuclear Information System (INIS)

Hayafune, Hiroki; Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Ohshima, Jun; Ito, Takaya

2011-01-01

JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)

Present status and future perspectives of research and test reactor in Japan

International Nuclear Information System (INIS)

Kaneko, Yoshihiko; Kaieda, Keisuke

2000-01-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

Present status and future perspectives of research and test reactor in Japan

Energy Technology Data Exchange (ETDEWEB)

Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2000-10-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

Data list of nuclear power plants in Japan

International Nuclear Information System (INIS)

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

1993-01-01

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

A review of fast reactor progress in Japan, March 1979

Energy Technology Data Exchange (ETDEWEB)

Tomabechi, K

1979-07-01

The fast reactor development project in Japan will be continued in the next fiscal year, from April 1979 through March 1980, at a similar scale of effort both in budget and personnel, to those of the fiscal year of 1978. The total budget for LMFBR development for the next fiscal year is approximately 24 billion Yen, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast reactor development in the PNC is approximately 500, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, approval for power increase from presently approved 50 MWt to 75 MWt with the present core and also to 100 MWt with a modified core in the future was granted by the regulatory authority in September 1978. Two operational cycles at 50 MWt have been completed very recently and preparation for power increase to 75 MWt is being made. With respect to the prototype fast breeder reactor MONJU, progress toward construction is being made and an environmental impact statement of MONJU filed last autumn is being reviewed by the concerned authorities. By the new atomic energy law recently made effective in Japan, the tasks of the former Japan Atomic Energy Commission were split into two and the Atomic Energy Safety Commission was newly established on 4th October 1978 in order to deal with nuclear safety problems in the country. All other problems are treated by the Atomic Energy Commission, as before. Highlights and topics of the fast reactor development activities in the past twelve months are summarized in this paper.

Progress on development of nuclear power in Japan

International Nuclear Information System (INIS)

Anon.

2000-01-01

Since three Laws on the nuclear power were published 45 years has passed. Now, development on nuclear power in Japan is at an emergent state. In Japan, 51 units of commercial nuclear reactors with 44.917 GW are in operation, occupy about 37% of total electric power generation, and is positioned at an essential basic energy source supporting economical society in Japan. However, an accident occurred at Tokai Works of the JCO Co., Ltd., one of the uranium reconversion company, on September 30, 1999, was the first critical accident in Japan, and became the worst case in history on development of nuclear power in Japan, because of forming three heavy radiation disabled persons (One of them was dead) in its operators. This was a big crisis with relation to existence on development of nuclear power in Japan, by which anxiety and distrust of the Japanese against the nuclear power were amplified rapidly. On the other side, for Japan short in energy sources and of a big energy consumption, in order to intend for a long term to carry out energy security, global environmental conservation, and sustainable maintenance of essential growth, it remains to be one of important optional methods to further promote nuclear power generation and to establish nuclear fuel cycle. Here were described on progress on peaceful applications of nuclear power in Japan, progress on the field of nuclear power in Japan (from 1955 to 1999), progress on Tokai nuclear power station, introduction of nuclear power generation and effort on its domestic production. (G.K.)

A review of fast reactor program in Japan. April 1997 - March 1998

International Nuclear Information System (INIS)

1998-01-01

This report describes the development and activities on fast reactor in Japan for the period of April 1997 - March 1998. During this period, two important results were drawn by the Special Committee on Fast Breeder Reactors (FBRs) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) Reform Committee, respectively. The Special Committee on FBRs discussed on the future FBR development in Japan including the Prototype FBR 'Monju' operation, and proposed its conclusion as the final report to the Japan Atomic Energy Commission (JAEC) on December 1, 1997. The PNC Reform Committee reviewed PNC's management and safety assurance system, and recommended to reform PNC to a new organization. Each committee result is outlined in this report. The Experimental Fast Reactor 'Joyo' operated 30th - 32nd cycle. In parallel with the operation, the Joyo Upgrading Program (MK-III program) is in progress. Five MK-III driver fuel subassemblies were loaded to the core in the 32nd cycle. Monju comprehensive safety review, which was started in December 1996, was continued through 1997, and was completed in March 1998. The DFBR Plant Optimization (phase 2) design study was launched by the Japan Atomic Power Company (JAPC) with goal of constructing FBR plant that achieves both reliability and economy from FY 1997 for three years. Research and development works are underway under the discussion and coordination of the Japanese FBR R and D Steering Committee, which is composed of PNC, JAPC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). (author)

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

International Nuclear Information System (INIS)

Izumi, Fumio; Harayama, Yasuo

1981-08-01

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

Seismic resistance design of nuclear power plant building structures in Japan

International Nuclear Information System (INIS)

Kitano, Takehito

1997-01-01

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

Seismic resistance design of nuclear power plant building structures in Japan

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

The development of breeder reactors in Japan

International Nuclear Information System (INIS)

Segawa, M.

1984-01-01

In the framework of a global analysis of the various available sources of energy, Japan has reserved a prominent place to the nuclear energy and, in the long-term view, to the breeder reactor which will be due for commercial deployment in 20)10. To achieve these objectives, three stages are envisaged, one of the experimental reactor Joyo (in service), one of the demonstration reactor Monju (its construction has been decided), and one of the pre-commercial reactor (due to be taken in hand at the beginning of the Nineties). Efforts will be made in parallel concerning the fuel cycle [fr

JTEC panel on nuclear power in Japan. Final report

International Nuclear Information System (INIS)

Hansen, K.F.; Behnke, W.B.; Cousin, S.B.; Evans, E.A.; Olander, D.R.

1990-10-01

The report examines the status and direction of nuclear power-related research and development in Japan in six areas: the nuclear fuel cycle, nuclear materials, instrumentation and control technology, CAD/CAM, nuclear safety research, and nuclear plant construction. Overall findings suggest that the nuclear power industry in Japan is at an advanced state of development; in particular, Japan is now technologically self-sufficient. Long-term goals of the Japanese program include closure of the complete fuel cycle and pursuit of the liquid metal fast breeder reactor as the future base system

Decommissioning and decontamination of licensed reactor facilities and demonstration nuclear power plants

International Nuclear Information System (INIS)

Lear, G.; Erickson, P.B.

1975-01-01

Decommissioning of licensed reactors and demonstration nuclear power plants has been accomplished by mothballing (protective storage), entombment, and dismantling or a combination of these three. The alternative selected by a licensee seems to be primarily based on cost. A licensee must, however, show that the decommissioning process provides adequate protection of the health and safety of the public and no adverse impact on the environment. To date the NRC has approved each of the alternatives in the decommissioning of different facilities. The decommissioning of small research reactors has been accomplished primarily by dismantling. Licensed nuclear power plants, however, have been decommissioned primarily by being placed in a mothballed state in which they continue to retain a reactor license and the associated licensee responsibilities

Present status and future outlook of nuclear power generation in Japan

International Nuclear Information System (INIS)

Kunikazu Aisaka

1987-01-01

The structure of energy consumption in Japan is heavily dependent on imported oil, therefore Japan has been making its greatest effort in developing nuclear power among other alternatives of oil. The capacity factor of the nuclear power plants in Japan marked 76% in FY 1986, exceeding 70% level for the past several years. The share of nuclear power is expected to increase steadily in the future. Future scale of the nuclear power generation is projected as 62,000 MW in year 2000 and as 137,000 MW in 2030. Nuclear power is expected to produce 58% of the nation's total power generation in 2030. Under the present circumstances, Janpan is executing a nuclear energy policy based on the following guidelines: 1. Promoting the safety advancement program; 2. Improving LWR technologies; 3. Program on use of plutonium in thermal reactors; 4. Advanced thermal reactors (ATRs); 5. Promotion of FBR development; 6. Nuclear fuel cycle. (Liu)

A review of fast reactor program in Japan

International Nuclear Information System (INIS)

Matsuno, Y.

1982-01-01

The fast breeder reactor development project in Japan has been in progress for the past twelve months and will be continued this fiscal year, from April 1982 through March 1983, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1981. The 1982 year budget for R and D work and for construction of a prototype fast breeder reactor MONJU is approximately 20 and 27 billion yen respectively, excluding wages for the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaged in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, power increase from 50 MWt to 75 MWt was made in July 1979 and six operational cycles at 75 MWt were completed in December 1981. With respect to the prototype reactor MONJU, progress toward construction has been made and an environmental impact statement of the reactor was approved by the authorities concerned, and the licensing of the first step was completed at the end of 1981. Preliminary design studies of a large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

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

A comparative analysis of the domestic and foreign licensing processes for power and non-power reactors

International Nuclear Information System (INIS)

Joe, J. C.; Youn, Y. K.; Kim, W. S.; Kim, H. J.

2003-01-01

The System-integrated Modular Advanced Reactor (SMART), a small to medium sized integral type Pressurized Water Reactor (PWR) has been developed in Korea. Now, SMART-P, a 1/5 scaled-down of the SMART, is being developed for the purpose of demonstrating the safety and performance of SMART design. The SMART-P is a first-of-a-kind reactor which is utilized for the research and development of a power reactor. Since the licensing process of such a reactor is not clearly specified in the current Atomic Energy Act, a comparative survey and analysis of domestic and foreign licensing processes for power and non-power reactors has been carried out to develop the rationale and technical basis for establishing the licensing process of such a reactor. The domestic and foreign licensing processes of power and non-power reactors have been surveyed and compared, including those of the U.S.A., Japan, France, U.K., Canada, and IAEA. The general trends in nuclear reactor classification, licensing procedures, regulatory technical requirements, and other licensing requirements and regulations have been investigated. The results of this study will be used as the rationale and technical basis for establishing the licensing process of reactors at development stage such as SMART-P

Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

Energy Technology Data Exchange (ETDEWEB)

Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

2014-05-01

It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

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

International Nuclear Information System (INIS)

Yamamoto, T.; Kasai, S.

2011-01-01

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

A review of fast reactor programme in Japan

Energy Technology Data Exchange (ETDEWEB)

Masuno, Y [Experimental Fast Reactor Division, O-arai Engineering Center, PNC (Japan); Bando, S [Project Planning and Management Division, PNC, Minato-ku, Tokyo (Japan)

1981-05-01

The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report.

A review of fast reactor programme in Japan

International Nuclear Information System (INIS)

Masuno, Y.; Bando, S.

1981-01-01

The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report

A report of the overall working group of the AEC Committee on Development of Advanced Power Reactors

International Nuclear Information System (INIS)

1979-01-01

The AEC Committee on Development of Advanced Power Reactors was set up in April, 1978, following on the previous AEC Special Committee on Development of Advanced Power Reactors, in order to study on the complementary power reactors between current LWRs and future FBRs. The subjects of study by the overall working group are the status of advanced power reactors in views of the nuclear fuel cycle, the impacts on industries, the selection of reactor types under present international circumstances, and the evaluation of advanced power reactors in their technology and economy. The following matters are described: evaluations in view of the nuclear fuel cycle, i.e. the features of the ATR of Japan and CANDU reactors of Canada; international problems concerning nuclear nonproliferation and securing of uranium; problems in the diversification of power reactor types concerning the expenditure and technology; problems of technology in the ATR of Japan, CANDU reactors of Canada and Pu utilization for LWRs; and the economy of D 2 O power reactors, i.e. the ATR of Japan and CANDU reactors of Canada. (J.P.N.)

Decontamination and radioactivity measurement on building surfaces related to dismantling of Japan power demonstration reactor (JPDR)

International Nuclear Information System (INIS)

Hatakeyama, Mutsuo; Tachibana, Mitsuo; Yanagihara, Satoshi

1997-12-01

In the final stage of dismantling activities for decommissioning a nuclear power plant, building structures have to be demolished to release the site for unrestricted use. Since building structures are generally made from massive reinforced concrete materials, it is not a rational way to treat all concrete materials arising from its demolition as radioactive waste. Segregation of radioactive parts from building structures is therefore indispensable. The rational procedures were studied for demolition of building structures by treating arising waste as non-radioactive materials, based on the concept established by Nuclear Safety Commission, then these were implemented in the following way by the JPDR dismantling demonstration project. Areas of the JPDR facilities are categorized into two groups : possibly contaminated areas, and possibly non-contaminated areas, based on the document of the reactor operation. Radioactivity on the building surfaces was then measured to confirm that the qualitative categorization is reasonable. After that, building surfaces were decontaminated in such a way that the contaminated layers were removed with enough margin to separate radioactive parts from non-radioactive building structures. Thought it might be possible to demolish the building structures by treating arising waste as non-radioactive materials, confirmation survey for radioactivity was conducted to show that there is no artificial radioactive nuclides produced by operation in the facility. This report describes the procedures studied on measurement of radioactivity and decontamination, and the results of its implementation in the JPDR dismantling demonstration project. (author)

Japan: The Experimental Fast Reactor JOYO. Profile 12

International Nuclear Information System (INIS)

2017-01-01

The experimental fast reactor JOYO of the Japan Atomic Energy Agency (JAEA) is the first sodium-cooled fast reactor (SFR) in Japan. JOYO attained its initial criticality as a breeder core (MK-I core) in 1977. During the MK-I operation, which consisted of two 50 MWt and six 75 MWt duty cycles, the basic characteristics of plutonium (Pu) and uranium (U) mixed oxide (MOX) fuel core and sodium cooling system were investigated and the breeding performance was verified. In 1983, the reactor increased its thermal output up to 100 MWt in order to start the irradiation tests of fuels and materials to be used mainly for other SFRs. Thirty-five duty cycle operations and many irradiation tests were successfully carried out using the MK-II core by 2000. The core was then modified to the MK-III core in 2003. In order to obtain higher fast neutron flux, the core was modified from one region core to two region core with different Pu fissile contents. Accordingly, the reactor power increased up to 140 MWt together with a renewal of intermediate heat exchangers (IHXs) and dump heat exchangers (DHXs). The rated power operation of the MK-III core started in 2004. The MK-III core has been used for the irradiation tests of fuels and materials for future SFRs and other R&D fields like innovative nuclear energy systems and technologies as well. This powerful neutron irradiation flux has an advantage especially for high burn-up fuel irradiation and material irradiation with high neutron dose. This paper shows the outline of the irradiation irradiation irradiation irradiation irradiation capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop innovative nuclear energy systems and technologies.

The current uranium exploration activities of the Power Reactor and Nuclear Fuel Development Corporation (PNC), Japan

International Nuclear Information System (INIS)

Miyada, H.

2001-01-01

As of November 1996, Japan's total installed commercial nuclear power generation capacity was 42 GW(e), accounting for 34% of total electric energy generation. By 2010, Japan intends to have an installed electricity generation capacity of 70.5 GW(e). This will increase the country's demand for nat Ural uranium from 7,700 t U in 1994 (13% of the world consumption) to 13,800 t U in 2010 (17%-19% of the world projected consumption). However, Japan's known uranium resources at Ningyo-Toge and Tono deposits, are estimated at roughly only 6,600 t U. The Long-term Programme for Research, Development and Utilization of Nuclear Energy (adopted in 1994) calls for diversification through long-term purchasing contracts, independent exploration and involvement in mining vent Ures, with the objective of ensuring independence and stability in Japan's development and utilization of nuclear energy. The Power Reactor and Nuclear Fuel Development Corporation (PNC) has been commissioned to carry out the task of independent exploration. PNC is carrying out exploration projects in Canada, Australia, USA and China targeting unconformity related type deposits with an eye to privatizing them. Currently about 40,000 t U of uranium resources are held by PNC. PNC has been carrying out the following related activities: (1) Reference surveys on uranium resources to delineate the promising areas; (2) Development of uranium exploration technology; (3) Information surveys on the nuclear industries to project long-term supply and demand; (4) International Cooperation programme on uranium exploration with Asian countries. (author)

Reactor construction programme in Japan

Energy Technology Data Exchange (ETDEWEB)

NONE

1959-04-15

In September last year, the Japanese Government requested the International Atomic Energy Agency to supply three tons of natural uranium for a research reactor, and the Agency has now arranged for its sale to Japan. The metal will be supplied in ingot form and after fabrication it will be used as fuel in a reactor of the natural uranium, heavy water type. The uranium will be obtained from Canada and sold to Japan by IAEA. The Agency had invited tenders for its supply, and after considering the tenders received, the Agency's Board of Governors decided that the Canadian offer to the Agency of three tons of natural uranium free of charge should be accepted and that the selling price to Japan should be US $35. 50 per kilogramme. The price takes into account Article XIV/E of the Agency's Statute which says that the Agency shall establish a scale of charges (including those for storage and handling) for materials furnished to Member States, and that the scale shall be designed to produce revenues to meet expenses in connexion with materials acquired by the Agency and costs of materials and services provided by it under agreements with one or more members. This is the first operation of its kind to be undertaken by the Agency, and the reactor for which the supply is being made will be the first in Japan to be constructed by Japanese scientists and technicians. IAEA's Board of Governors has given the necessary approval to the reactor project for which the Agency is providing assistance

Fusion-power demonstration

International Nuclear Information System (INIS)

Henning, C.D.; Logan, B.G.; Carlson, G.A.; Neef, W.S.; Moir, R.W.; Campbell, R.B.; Botwin, R.; Clarkson, I.R.; Carpenter, T.J.

1983-01-01

As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment

Fusion power demonstration

International Nuclear Information System (INIS)

Henning, C.D.; Logan, B.G.

1983-01-01

As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment

Seismic design of equipment and piping systems for nuclear power plants in Japan

International Nuclear Information System (INIS)

Minematsu, Akiyoshi

1997-01-01

The philosophy of seismic design for nuclear power plant facilities in Japan is based on 'Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981' (referred to as 'Examination Guide' hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in 'Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association'. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

Seismic design of equipment and piping systems for nuclear power plants in Japan

Energy Technology Data Exchange (ETDEWEB)

Minematsu, Akiyoshi [Tokyo Electric Power Co., Inc. (Japan)

1997-03-01

The philosophy of seismic design for nuclear power plant facilities in Japan is based on `Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981` (referred to as `Examination Guide` hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in `Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association`. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

International Nuclear Information System (INIS)

Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

2000-01-01

Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

A review of fast reactor program in Japan - April 1983

International Nuclear Information System (INIS)

Matsuno, Y.

1983-01-01

The fast breeder reactor development project in Japan has been in progress during the past twelve months and will be continued in the next fiscal year, from April 1983 through March 1984, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1982. Concerning the experimental fast reactor, JOYO, all the scheduled testings and operations were completed by the end of 1981 and from the beginning of 1982 the change-out work from Mark-I core to Mark-II core has been continued for 11 months. The initial criticality on the Mark-II core was achieved on 22 Nov. 1982 and after 3 months low power physics tests the reactor power was raised up to 100% (100 MWt) in the middle of March 1983. With respect to the prototype reactor MONJU, progress toward construction has been made and the licensing of the second step will be completed in the first half of 1983. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

Recent development of nuclear power in Japan and instrumentation and control system and control room equipment for advanced light water reactors

International Nuclear Information System (INIS)

Wakayama, N.

1992-01-01

This paper was provided for the 13th IAEA/IWG-NPPCI Meeting and aims to introduce an outline of recent development of nuclear power in Japan and some topics in the field of nuclear power plant control and instrumentation. Forty units of nuclear power plants are in operation in Japan and five units of BWRs and six PWRs are under construction. Construction of prototype FBR Monju have almost completed an construction of High-Temperature Engineering Test Reactor, HTTR, started in March 1991. In parallel of those, extensive effort has been carried out to develop the third generation LWRs which are called Advanced BWR (ABWR) and Advanced PWR (APWR). Two Advanced BWRs are under safety review for construction. Instrumentation and control system of these Advanced LWRs adopts integrated digital I and C system, optical multiplexing signal transmission, fault tolerant control systems and software logic for reactor protection and safety systems and enhances plant control performance and provides human-friendly operation and maintenance environments. Main control room of these Advanced LWRs, comprised with large display panels and advanced console, has special futures such as one-man sit-down operation, human friendly man-machine interface, high level automation in operation and maintenance. (author). 7 refs, 9 figs, 1 tab

What is nuclear power in Japan?

Science.gov (United States)

Suzuki, Toshikazu

2011-03-01

The aggressive use of such non-fossil energy as the atomic energy with high power density and energy production efficiency is an indispensable choice aiming at the low-carbon society. There is a trial calculation that the carbon dioxide emission of 40000 ton can be suppressed by nuclear power generation by one ton of uranium. The basis of nuclear research after the Second World War in Japan was established by the researchers learnt in Argonne National Laboratory. In 2010, NPPs under operation are 54 units and the total electric generating power is 48.85GW. The amount of nuclear power generation per person of the people is 0.38kW in Japan, and it is near 0.34kW of the United States. However, the TMI accident and the Chernobyl disaster should have greatly stagnated the nuclear industry of Japan although it is not more serious than the United States. A lot of Japanese unconsciously associate a nuclear accident with the atomic bomb. According to the investigation which Science and Technology Agency carried out to the specialist in 1999, ``What will be the field where talent should be emphatically sent in the future?'' the rank of nuclear technology was the lowest in 32 fields. The influence of the nuclear industry stagnation was remarkable in the education. The subject related to the atomic energy of a university existed 19 in 1985 that was the previous year of the Chernobyl disaster decreased to 7 in 2003. In such a situation, we have to rely on the atomic energy because Japan depends for 96% of energy resources on import. The development of the fuel reprocessing and the fast breeder reactor has been continued in spite of a heavy failure. That is the only means left behind for Japan to be released from both fossil fuel and carbon dioxide.

Promotion of research and development for fast breeder reactors in Japan

International Nuclear Information System (INIS)

1989-01-01

The ''Long-term Program for Development and Utilization of Nuclear Energy'' by Japan Atomic Energy Commission (JAEC) plans to develop fast breeder reactors (FBRs) as the mainstay of Japan's future nuclear power generation. For the development of the FBR, the main principle is to pursue the establishment of FBR plutonium utilization systems, which are superior to the LWR uranium utilization systems. The above program also states as follows: ''Incessant efforts, with proper co-operation of the public and private sectors, will be required for a considerably long time for FBR development, and the concrete measures to be implemented in this connection will be discussed, hereafter, in continuation at the Advisory Committee on FBR Development Program of the Atomic Energy Commission''. It was clearly stated in the ''Long-term Program'' that utilities would play a leading role in the design, construction, and operation of DFBRs to follow ''Monju''. They have entrusted Japan Atomic Power Company (JAPC) with the above tasks. Technical verification and demonstration tests for seismic design and feasibility studies of main components are being conducted with government support. Regarding the nuclear fuel cycle, PNC has been developing FBR fuel reprocessing technologies. Currently, the Recycle Equipment Test Facility (RETF) is being designed to conduct engineering-scale tests. A pilot plant is planned to start operation after 2000 assuming positive results of the tests. (author)

A review of fast reactor programme in Japan

International Nuclear Information System (INIS)

Matsuno, Y.; Bando, S.

1981-03-01

The fast breeder reactor development project in Japan made progress in the past year, and will be continued in the next fiscal 1981. The scale of efforts both in budget and personnel will be similar to those in fiscal 1980. The budget for R and D works and for the construction of the fast breeder prototype reactor ''Monju'' will be approximately 20 billion yen and 27 billion yen, respectively, excluding the wage of the personnel concerned. The number of the technical personnel currently engaging in fast breeder reactor development in the Power Reactor and Nuclear Fuel Development Corp. is about 530. As for the experimental fast reactor ''Joyo'', three operational cycles at 75 MWt have been completed in August, 1980, and the fourth cycle has started in March, 1981. As for the prototype reactor ''Monju'', progress was made toward the construction, and the environmental impact statement on the reactor was approved by the authorities concerned. The studies on the preliminary design of large LMFBRs have been made by the PNC and also by power companies. The design study carried out by the PNC is concerned with a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of the commissioning of ''Monju''. The highlights and topics in the development activities for fast breeder reactors in the past twelve months are summarized in this report. (Kako, I.)

Future of nuclear power in Japan - Development of next Generation LWRs

Energy Technology Data Exchange (ETDEWEB)

Yamada, Eiji; Yamamoto, T.; Kurosaki, K.; Ohga, Y.; Tsuzuki, K.; Kasai, S.; Tanaka, T.

2010-09-15

Japan's energy policies have been to decrease the oil portion and dependence on the Middle East for energy security, as well as satisfy environmental requirement. The report of 2008 targeted reducing GHG emission by 60-80% before 2050, and highlighted ''Cool Earth-Innovative Energy Technology Program'' featuring 21 innovative technologies. In this context nuclear power is expected as a core power source. In April 2008, ''Next Generation Light Water Power Reactor Development Program'' was launched with the IAE as the core organization in alliance with Japan's major vendors and in collaboration with METI and power utilities for the future of nuclear power.

A review of fast reactor program in Japan. April 2000 - March 2001

International Nuclear Information System (INIS)

Nagata, Takashi; Yamashita, Hidetoshi

2001-01-01

This report describes the development and activities on fast reactors in Japan thru April 2000 to March 2001. During this period, the most important result of the Japanese Fast Reactor Project was that the first phase 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was completed at the end of March 2001, and the second phase study has just started in order to narrow down the candidate concepts selected in the first phase for next stage. In the Experimental Fast Reactor 'Joyo', the 35 th rated power operation was completed by the end of May 2000. The 13 th periodical inspection and reconstruction works for the Joyo upgrading program (MK-III) were started on the beginning of June 2000. The modification of the cooling system is underway. In the Prototype Fast Breeder Reactor 'Monju', countermeasures against sodium leakage have already been drawn up based on 'Monju' comprehensive safety review. The Japan Atomic Energy Commission (JAEC) has issued a new 'Long-term Program for Research, Development and Utilization of Nuclear Energy' in November 2000. (author)

Fusion power demonstration - a baseline for the mirror engineering test reactor

International Nuclear Information System (INIS)

Henning, C.D.; Logan, B.G.; Neef, W.S.

1983-01-01

Developing a definition of an engineering test reactor (ETR) is a current goal of the Office of Fusion Energy (OFE). As a baseline for the mirror ETR, the Fusion Power Demonstration (FPD) concept has been pursued at Lawrence Livermore National Laboratory (LLNL) in cooperation with Grumman Aerospace, TRW, and the Idaho National Engineering Laboratory. Envisioned as an intermediate step to fusion power applications, the FPD would achieve DT ignition in the central cell, after which blankets and power conversion would be added to produce net power. To achieve ignition, a minimum central cell length of 67.5 m is needed to supply the ion and alpha particles radial drift pumping losses in the transition region. The resulting fusion power is 360 MW. Low electron-cyclotron heating power of 12 MW, ion-cyclotron heating of 2.5 MW, and a sloshing ion beam power of 1.0 MW result in a net plasma Q of 22. A primary technological challenge is the 24-T, 45-cm bore choke coil, comprising a copper hybrid insert within a 15 to 18 T superconducting coil

Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

International Nuclear Information System (INIS)

Wakai, Takashi; Machida, Hideo; Yoshida, Shinji; Xu, Yang; Tsukimori, Kazuyuki

2014-01-01

This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J IC , and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins

Waste inventory, waste characteristics and waste repositories in Japan

International Nuclear Information System (INIS)

Shimooka, K.

1997-01-01

There are two types of repositories for the low level radioactive wastes in Japan. One is a trench type repository only for concrete debris generated from the dismantling of the research reactor. According to the safety assurance system, Japan Atomic Energy Research Institute (JAERI) has disposed of the concrete debris arose from the dismantling of the Japan Power Demonstration Reactor (JPDR). The other type is the concreted pit with engineered barriers. Rokkasho Low Level Radioactive Waste Disposal Center has this type of repository mainly for the power plant wastes. Japan Nuclear Fuel Ltd. (JNFL) established by electric power companies is the operator of the LLW disposal project. JNFL began the storage operation in 1992 and buried approximately 60,000 drums there. Two hundred thousand drums of uniformly solidified, waste may be buried ultimately. 4 refs, 3 tabs

The summary of the nuclear power development in Japan

International Nuclear Information System (INIS)

Mizoguchi, Kenzo; Hirose, Yasuo; Fukai, Yuzo; Hada, Mikio; Ogawa, Nagao.

1980-01-01

A quarter of century has elapsed since the development of atomic energy was started in Japan. At present, the scale of nuclear power generation reached the operation of 22 plants with about 15.12 million kW capacity, and 12% of the total installation capacity for power generation. Efforts have been exerted to bring up the domestic technologies gradually, while importing and digesting quickly the foreign technologies. Now in LWRs, the equipments of nearly 100% can be produced by the domestic technologies, moreover, the technologies have reached such level that they can be exported to foreign countries. In the last five years, the improvement and standardization of LWR technologies have been promoted. The development of the reactors of new types has been continued by the domestic technologies. According to the long term plan, the nuclear power generation of 53 million kW is expected by 1990, but various problems such as the location of nuclear power stations and nuclear fuel cycle remain, and considerable difficulty is expected in its materialization. The history of nuclear power generation in Japan, the features and progress of LWRs, the photographs and the main specifications of notable nuclear power plants, and the future perspectives of LWRs, the reactors of new types, nuclear fusion and nuclear fuel cycle are described. (Kako, I.)

Dismantling of JPDR begins: to demonstrate advanced technology

Energy Technology Data Exchange (ETDEWEB)

1986-12-01

The first dismantling of the Japan Power Demonstration Reactor (JPDR, BWR, 90 MWt, 12.5 MWe) began on December 4, 1986, claiming the attention of nuclear interests in Japan and overseas. The Japan Atomic Energy Research Institute undertook the project as the second phase of the six year program for dismantling the JPDR at the Tokai Research Establishment. It is the demonstration of the technology developed in the first phase of the program from 1981 to 1986, aiming at establishing a total system for dismantling commercial nuclear power plants in the furture. At the ceremony for the beginning of dismantling held on December 4 at the site, a special switch was operated to fire a gas burner, and cutting of the upper head of the reactor pressure vessel on the service floor of the reactor building began. The long term program on the development and utilization of nuclear energy in 1982 decided the basic policy on reactor decommissioning. Under this policy, in July, 1984, the nuclear subcommittee of the Advisory Committee for Energy set up the guideline for standardized decommissioning suitable to the actual situation in Japan. The schedule of the program, the development of eight fundamental techniques, disassembling techniques, decontamination, measurement and robotics are described. (Kako, I.).

Advanced Demonstration and Test Reactor Options Study

International Nuclear Information System (INIS)

Petti, David Andrew; Hill, R.; Gehin, J.; Gougar, Hans David; Strydom, Gerhard; Heidet, F.; Kinsey, J.; Grandy, Christopher; Qualls, A.; Brown, Nicholas; Powers, J.; Hoffman, E.; Croson, D.

2017-01-01

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power's share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy's (DOE's) broader commitment to pursuing an 'all of the above' clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate 'advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy'. Advanced reactors are

The status and prospects for the fossil-fired and nuclear power industry in Japan

International Nuclear Information System (INIS)

Miyahara, S.

1994-01-01

Power plant capacity in Japan amounts to about 200 GW, of which 180 GW belong to the electricity supply industry. 60% are installed in fossil-fired power stations, 19% in nuclear power stations and 21% in hydro-electric power stations. Key engineering techniques for power production from fossil fuels are supercritical steam conditions and combined cycle power plant technology. Crucial points for nuclear power generation are the development of the advanced light water reactor, the commericialization of the fast breeder reactor and the installation of a closed nuclear fuel cycle. (orig.) [de

Handling of spent fuel from research reactors in Japan

International Nuclear Information System (INIS)

Kanda, K.

1997-01-01

In Japan eleven research reactors are in operation. After the 19th International Meeting on Reduced Enrichment for Research Reactors and Test Reactors (RERTR) on October 6-10, 1996, Seoul, Korea, the Five Agency Committee on Highly Enriched Uranium, which consists of Science and Technology Agency, the Ministry of Education, Science and Culture, the Ministry of Foreign Affairs, Japan Atomic Energy Research Institute (JAERI) and Kyoto University Research Reactor Institute (KURRI) met on November 7,1996, to discuss the handling of spent fuel from research reactors in Japan. Advantages and disadvantages to return spent fuel to the USA in comparison to Europe were discussed. So far, a number of spent fuel elements in JAERI and KURRI are to be returned to the US. The first shipment to the US is planned for 60 HEU elements from JMTR in 1997. The shipment from KURRI is planned to start in 1999. (author)

Effort to grapple with improvement of security and reliability of nuclear power plant. Actions of the Japan Atomic Power Company

International Nuclear Information System (INIS)

Ishiguma, Kazuo

2012-01-01

Following the Great Tohoku Earthquake in 2011, Tokai No.2 reactor was shut down automatically. Three of emergency diesel generators worked automatically at loss-of-offsite-power and began to work the cooling system of reactor. The reactor could be kept stable and safe in cold state by management of power from the gas turbine electric generator and power source car. Actions of Japan Atomic Power Company (JAPC) for cold shutdown and Tsunami were stated. Inspection results after the earthquake and testimony of staff was described. Countermeasure of improvement of safety of nuclear power station is explained by ensuring of power source and water supply, crisis management system, countermeasure of accident, ensuring, and training of workers, and action for better understanding of reliance. (S.Y.)

Fast reactors in nuclear power

Energy Technology Data Exchange (ETDEWEB)

Kazachkovskii, O

1981-02-01

The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

Challenge to the enhancement of LWRs in Japan

International Nuclear Information System (INIS)

Inoue, T.

1996-01-01

Japan was way behind western countries in nuclear power development. This is partly because of the defeat in the second world war which bound Japan to the ban on nuclear power development. As a result, Japan was obliged to take the policy of introducing the power reactor, and its technology, from overseas advanced countries in order to promote the development of power reactors for its own country. (orig.)

Advanced Demonstration and Test Reactor Options Study

Energy Technology Data Exchange (ETDEWEB)

Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-01-01

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy�

A prospect of fast reactor and related fuel cycle in Japan

Energy Technology Data Exchange (ETDEWEB)

Nagata, Takashi [Japan Atomic Energy Agency, Ibaraki (Japan)

2009-04-15

JAEA has launched a new project 'Fast Reactor Cycle Technology Development'(FaCT) in cooperation with electric utilities. In this FaCT project, a combination of 'the Japanese sodium cooled loop type fast reactor with oxide fuel, the advanced aqueous reprocessing, and the simplified palletizing fuel fabrication systems' is adopted, where many innovative technologies with technical challenging issues are actively used in order to provide significant improvements in economic competitiveness, and enhancement of safety and reliability, sustainability, and nonproliferation. Fast reactor cycle technology will provide harmonic solutions for global issues of energy resources and environments, and is expected to contribute to sustainable development of the future society. Therefore, it was selected as one of key technologies of national importance in the third term (JPY2006-2010) 'Science and Technology Basic Plan' in March 2006 in Japan. The 'Nuclear Energy National Plan' in August 2006 states start up of a demonstration FR by around 2025 and deployment of a commercial FR before 2050, and start operating fuel cycle facilities when these reactors achieve consistency. Accordingly, we will decide about the adoption of innovative technologies by judging their applicability by 2010, and present the conceptual designs of commercial and demonstration FR cycle facilities by 2015 with the R and D plans to realize. In developing the FR cycle, 5 Party council, which consists of MEXt, MITI, electricity utilities, manufacturers, and JAEA, was established in July 2006 for moving forward on the commercialization smoothly. In this framework, users' requirements for the future R and D, a scenario of transition from light water reactor cycle to sodium cooled FR cycle, international collaboration, development schedule, demonstration steps, and so on are discussed. In this presentation, a prospect concerning the system design features of JSFR and a

A prospect of fast reactor and related fuel cycle in Japan

International Nuclear Information System (INIS)

Nagata, Takashi

2009-01-01

JAEA has launched a new project 'Fast Reactor Cycle Technology Development'(FaCT) in cooperation with electric utilities. In this FaCT project, a combination of 'the Japanese sodium cooled loop type fast reactor with oxide fuel, the advanced aqueous reprocessing, and the simplified palletizing fuel fabrication systems' is adopted, where many innovative technologies with technical challenging issues are actively used in order to provide significant improvements in economic competitiveness, and enhancement of safety and reliability, sustainability, and nonproliferation. Fast reactor cycle technology will provide harmonic solutions for global issues of energy resources and environments, and is expected to contribute to sustainable development of the future society. Therefore, it was selected as one of key technologies of national importance in the third term (JPY2006-2010) 'Science and Technology Basic Plan' in March 2006 in Japan. The 'Nuclear Energy National Plan' in August 2006 states start up of a demonstration FR by around 2025 and deployment of a commercial FR before 2050, and start operating fuel cycle facilities when these reactors achieve consistency. Accordingly, we will decide about the adoption of innovative technologies by judging their applicability by 2010, and present the conceptual designs of commercial and demonstration FR cycle facilities by 2015 with the R and D plans to realize. In developing the FR cycle, 5 Party council, which consists of MEXt, MITI, electricity utilities, manufacturers, and JAEA, was established in July 2006 for moving forward on the commercialization smoothly. In this framework, users' requirements for the future R and D, a scenario of transition from light water reactor cycle to sodium cooled FR cycle, international collaboration, development schedule, demonstration steps, and so on are discussed. In this presentation, a prospect concerning the system design features of JSFR and a summary of the above R and D progresses for

Summary of the progress of reactor physics in Japan reviewing the activities related to NEA Committee on Reactor Physics

International Nuclear Information System (INIS)

Hirota, Jitsuya

1984-09-01

The progress of fast and thermal reactor physics, fusion neutronics and shielding researches in these twenty years can be clearly recognized in the reviews of reactor physics activities in Japan which had been perpared by the Special Committee on Reactor Physics: the joint committee under Atomic Energy Society of Japan and JAERI. Many topics of those discussed at the NEACRP meetings concerned fast reactor physics. Information exchange on the topics such as adjustment of group cross sections by integral data, central worth discrepancy, sodium void effect and heterogeneous core stimulated the researches in Japan. And achievements in Japan including those in the JAERI Fast Critical Facility FCA were reported and contributed largely to the international co-operation. In addition, the contribution from Japan was also made concerning a study of fusion blanket. Among various specialists' meetings recommended by NEACRP, those on nuclear data and benchmarks for reactor shielding were often held since 1973 and helpful to the progress of shielding researches in Japan. The Third Specialists' Meeting on Reactor Noise (SMORN-III) was held in Tokyo in 1981, indicating the recent progress in safety-related applications of reactor noise analysis. The NEACRP benchmark tests were quite useful to the progress of reactor physics in Japan, which included the benchmark calculations of BWR lattice cell, key parameters and burn-up characteristics of a large LMFBR, FBR and PWR shielding, and so on. It may be noted that the benchmark test on reactor noise analysis methods was successfully conducted by Japan in connection with SMORN-III. In addition, the co-operation was positively made to the compilation of light water lattice data, and the preparation of reviews on actinide production and burn-up, and blanket physics. (J.P.N.)

Nuclear power development in Japan

International Nuclear Information System (INIS)

Sugawara, A.

1994-01-01

Energy situation in Japan and Japan's strategy for stable supply of energy are discussed. Benefits of nuclear power in comparison with other energy sources is considered. History of nuclear power development in Japan, modern status and future trends are described. 6 figs

Problems in Siting Nuclear Power Plants in Japan and Efforts to Solve Them

Energy Technology Data Exchange (ETDEWEB)

Inouye, T. [Ministry of International Trade and Industry, Tokyo (Japan)

1967-09-15

The rapidly growing demand for energy in Japan will require a total capacity of 30 to 40 thousand MW(e) in nuclear power by 1985. Materialization of this development programme must naturally be supported by securing the requisite sites for the nuclear power plants. The following factors make siting of nuclear power plants more difficult in Japan than in any other country: a small, densely populated territory with little level land, that is already completely utilized for agricultural and/or industrial purposes; small rivers and an active marine-product industry developed along most of the seacoasts, both of which create difficult cooling-water problems; frequent earthquakes; and the fear of possible radioactivity, which prevails in the only nation in the world to have suffered from the atomic bomb. There are at present four nuclear power plants in operation or under construction in Japan with a total capacity of about 1.3 thousand MW(e). However, the plants in these construction programmes have been sited on the basis of taking the easiest course available although there were several possible solutions to choose from. It is pointed out here that the long-range nuclear power development programme will call for a fundamental solution to enable siting a large number of power plants under the adverse conditions in Japan. Accordingly, a study was made, which included quantitative analyses of reactor siting factors and suggested measures for solving the siting problems. The analyses were based on nuclear power plant sites assumed to be located on the seacoast and characterized by low-population density, desirable geology and favourable topography. It was assumed that seacoast siting was more economical than inland siting. Although the study was made by a general survey using maps, it was shown that approximately 10% of the total coastline areas would be eligible for reactor siting, but most of these areas in this case are located in the northern part of Japan, far from

Evolution of the Fusion Power Demonstration tandem mirror reactor configuration

International Nuclear Information System (INIS)

O'Toole, J.A.; Lousteau, D.C.

1985-01-01

This paper gives a presentation of the evolution of configurations proposed for tandem mirror Fusion Power Demonstration (FPD) machines. The FPD study was undertaken to scope the mission as well as the technical and design requirements of the next tandem mirror device. Three configurations, entitled FPD I, II, and III were studied. During this process new systems were conceived and integrated into the design, resulting in a significantly changed overall machine configuration. The machine can be divided into two areas. A new center cell configuration, minimizing magnetic field ripple and thus maximizing center cell fusion power, features a semicontinuous solenoid. A new end cell has evolved which maintains the required thermal barrier in a significantly reduced axial length. The reduced end cell effective length leads to a shorter central cell length being required to obtain minimum ignition conditions. Introduced is the concept of an electron mantle stabilized octopole arrangement. The engineering features of the new end cell and maintenance concepts developed are influenced to a great extent by the octopole-based design. The new ideas introduced during the FPD study have brought forth a new perspective of the size, design, and maintenance of tandem mirror reactors, making them more attractive as commercial power sources

Some problems on domestic technology development from a point of fabricator of nuclear power plant. [Japan

Energy Technology Data Exchange (ETDEWEB)

Watamori, T [Hitachi Ltd., Tokyo (Japan)

1976-06-01

During past 20 years, the nuclear power industry in Japan has introduced foreign technology, digested it in a short period, and continued to research and develop domestic technology. Now, 95% of the machinery and equipments for nuclear power generation with light water reactors can be produced domestically, and some technologies are going to be exported. However, the nuclear power industry is still in a severe environment. The progress of the development of nuclear power plants passed the periods of organizational preparation, the construction of research reactors, the import of foreign technologies and reactors for practical use, and the construction of domestically produced reactors for practical use. The supplying capacity of the nuclear power industry in Japan reached 6 units of 1,000 MW yearly, but in order to meet the long term plan of nuclear power generation, this capacity must be further enhanced. The problems in the promotion of domestic production are the establishment of independent technologies, the promotion of standardization, the strengthening of business basis, the upbringing of relating enterprises, and the acceleration of national projects. Since the energy crisis, the trend of filling up energy demand with nuclear power generation became conspicuous, but for the expansion of export, the problems of safety guarantee, nuclear fuel cycle, and financial measures must be solved with government aid.

Present status of Japan materials testing reactor

Energy Technology Data Exchange (ETDEWEB)

Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

Present status of Japan materials testing reactor

International Nuclear Information System (INIS)

Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi

2012-01-01

The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

Foundations of a long-term strategy for nuclear power development in Japan

International Nuclear Information System (INIS)

Murata, H.

1975-01-01

A long-term strategy for nuclear power developments in Japan is proposed. The situation in the world has greatly changed in the recent years due to the rise in oil prices as well as the considerable concern about the environmental problems caused by the nuclear power plants. Stress is being placed on the harmonization with the environmental protection rather than on the economical generation of the nuclear power. In order to meet the future requirements, five systems are given for the short, medium and long ranges beyond the year 2000. For the final stage a system is proposed that combines fusion-fission hybrid reactors with very high temperature gas cooled reactors to supply clean energy. (author)

Urgent Safety Measures in Japan after Great East Japan Earthquake

International Nuclear Information System (INIS)

Taniura, Wataru; Otani, Hiroyasu

2012-01-01

Due to tsunami triggered by the Great East Japan Earthquake, the operating and refueling reactor facilities at Fukushima Dai-ichi and Dai-ni Nuclear Power Plants caused a nuclear hazard. Given the fact, Japanese electric power companies voluntarily began to compile various urgent measures against tsunami. And then the Nuclear and Industrial Safety Agency (NISA) ordered the licensees to put into practice the voluntarily compiled urgent safety measures, in order to ensure the effectiveness of the means for recovering cooling functions along with avoiding the release of radioactive substances to the possible minimum, even if a huge tsunami following a severe earthquake hits nuclear power plants. The following describes the state and the effect of the urgent safety measures implemented for 44 reactors (under operation) and 1 reactor (under construction) in Japan and also describes the measures to be implemented by the licensees of reactor operation in the future.

The ORNL fusion power demonstration study

International Nuclear Information System (INIS)

Shannon, T.E.; Steiner, D.

1978-01-01

In this paper, we review the design approach developed in the ORNL Fusion Power Demonstration Study [1]. The major emphasis of this study is in the application of current and near-term technology as the most logical path to near-term demonstration of tokamak fusion power. In addition we are pursuing a number of concepts to simplify the tokamak reactor to be more acceptable to the utility industry as a future source of energy. The discussion will focus on the areas having the greatest overall impact on reactor feasibility: 1) overall size and power output, 2) remote maintenance considerations, 3) electrical power supplies, 4) blanket design; and 5) economics. The tokamak device, by nature of its configuration and pulsed operation, is an exceptionally complex engineering design problem. We have concluded that innovative design concepts are essential to cope with this basic complexity. We feel that the feasibility of tokamak fusion power has been significantly improved by these design approaches. (author)

Japan

International Nuclear Information System (INIS)

Huttner, Kevin; Suzuki, Tatsujiro

1987-01-01

The Japanese nuclear power programme began with reactors imported from the United States. A natural uranium heavy water reactor using domestic materials was started in 1958. Subsequent progress was with light water reactors imported from the United States. Domestic reactor development was of a fast breeder reactor and an advanced thermal reactor. By March 1986 there were 32 commercial power plants in operation which produced approximately 23% of the electricity consumed in Japan. Ten more are under construction and six more are planned. Their location and comparative generating cost are tabulated. Energy demand and targets for nuclear power generation are discussed. The FBR advanced thermal reactor and high temperature reactor programmes are summarized. The Japanese nuclear fuel cycle - uranium prospecting, enrichment, reprocessing, the development of mixed oxide fuels, thermal recycling and radioactive waste management is also discussed. (U.K.)

Safety aspect of digital reactor protection system in Japan

International Nuclear Information System (INIS)

Ogiso, Zen-Ichi

1998-01-01

It was early in 1980's that the digital controllers were first applied to nuclear power plant in japan. After that, their application area had been expanding gradually, reaching to the overall integrated digital system including the safety system in Kashiwazaki-Kariwa units 6 and 7. The software for computer-based systems has been produced using the graphical language ''POL'' in Japanese nuclear power plants. It is the fundamental principle that the reliability of the software should be assured through the properly managed quality assurance. The POL-based system is fitted to this principle. In applying POL-based systems to safety system, the MITI, Ministry of International Trade and Industry, identified the licensing issues as the regulatory body, while the utilities had developed the digital technology feasible to the safety application. Through the activities, a specific industrial design guide for the software important to safety was established and the adequacy of the technology was certified through the demonstration tests of the integrated system. In the safety examination of the digital reactor protection system of K-6/7, the application of POL were approved. The POL-based systems in nuclear power plants were successful design and production process of the POL-based systems. This paper describes the activities in licensing and maintaining the computer-based systems by the utilities and manufacturers as well as the MITI. (author)

Development of demonstration advanced thermal reactor

Energy Technology Data Exchange (ETDEWEB)

Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige

1982-08-01

The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported.

Development of demonstration advanced thermal reactor

International Nuclear Information System (INIS)

Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige.

1982-01-01

The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

Maintenance management of nuclear power reactors at the stage of research and development

International Nuclear Information System (INIS)

Takaya, Shigeru; Chikazawa, Yoshitaka; Kubo, Shigenobu; Hayashida, Kiichi; Tagawa, Akihiro; Yamashita, Atsushi

2016-07-01

A maintenance management required to nuclear power reactors at the R and D stage was discussed in this report. It is the most important to ensure safety of nuclear power plants by taking account of characteristics of nuclear power reactors at the R and D stage. In addition, it is needed to establish a system of maintenance management technologies suitable for reactor types. In this report, objectives of maintenance management of nuclear power reactors at the R and D stage were clarified. Next, requirements and consideration for maintenance management of nuclear power reactors at the R and D stage were discussed according to the objectives. 'Code for Maintenance at Nuclear Power Plants' and 'Guide for Maintenance at Nuclear Power Plants' published by the Japan Electric Association were refereed in the discussion. Then, a draft of codes for maintenance management of nuclear power plants at the R and D stage was newly proposed. Finally, an example that the draft codes were applied to components containing sodium, typical components of sodium-cooled fast reactor, was presented. (author)

Project WAGR: the UK demonstration project for power reactor decommissioning - a review of the tools used to dismantle the reactor core

International Nuclear Information System (INIS)

Benest, T.G.

2008-01-01

The United Kingdom Atomic Energy Authority (UKAEA) has built and operated a wide range of nuclear facilities since the late 1940. UKAEA mission is to restore the environment of its sites in a safe and secure manner. This restoration includes the decommissioning of a number of redundant research and power reactors. The Windscale Advanced Gas-cooled Reactor (WAGR) was the UK prototype Advanced gas cooled reactor and became the forerunner of a family of 14 reactors built to generate cheaper and more efficient electricity in the UK. WAGR was constructed between 1957 and 1961 and was a carbon dioxide cooled, graphite moderated reactor using uranium oxide fuel in stainless steel cans. The reactor consisted of a graphite moderator housed in a cylindrical reactor vessel with hemispherical ends. The reactor and associated heat exchangers were enclosed in the iconic spherical containment building regularly used by the media in the UK as an illustration of the nuclear industry. The reactor first produced power in August 1962 and achieved full design output in 1963. It operated at an electrical output of 33 MW (E) for 18 years (average load factor of 75%). In 1981 the reactor was shut down after satisfactory completion of all the research and development objectives. In anticipation of the UK likely nuclear decommissioning needs the UKAEA decided to decommission WAGR to the International Atomic Energy Agency (IAEA) stage 3 (restoration of the area occupied by the facility to a condition of unrestricted re-usability) as the national demonstration exercise for power reactor decommissioning. Since 1998 the UKAEA and its contractors have been undertaking the dismantling of the reactor core components and pressure vessel in a series of 10 campaigns. These contain neutron activated components expected to produce dose rates well in excess of 1 Sv/hr. To carry out the work UKAEA installed an 8M remote dismantling machine (RDM) a waste recovery and transport system and a shielded waste

Analysis on the Electric Power Supply - Demand Measures of Japan in 2011 Summer after Earthquake and Tsunami

International Nuclear Information System (INIS)

Lee, Y. E.; Chang, H. S.

2011-01-01

Only 12 of 54 nuclear reactors are in operation as of September 1, 2011 in the wake of the earthquake and tsunami in Japan. The share of nuclear power in the nation's installation capacity fell to about 14% in August from about 30% before March 11, 2011. Government or many of research institutes estimated that the power supply system in Japan would fall to the minus reserve margin, if the nuclear power stations could not be restarted as scheduled. However, the current situation of power supply system in Japan is less severe than expected before, because the power companies and public have engaged in various diligent efforts to boost supply capacity or reduce demand in response to the electric power crisis. This paper aims to analyze the how much Japan electric power supply system depends on the nuclear power, what kinds of countermeasures of electric power supply-demand are taken by electricity companies in summer time to avoid the blackouts and why the saving electricity in Japan could be possible unlike Korea. Insights from this paper would be taken into account in the long term energy planning, even though the further study in depth should be followed

Nuclear energy in Japan

International Nuclear Information System (INIS)

Guillemard, B.

1978-01-01

After having described the nuclear partners in Japan, the author analyzes the main aspects of Japan's nuclear energy: nuclear power plants construction program; developping of light water reactors; fuel cycle politics [fr

[The Chinese nuclear test and 'atoms for peace' as a measure for preventing nuclear armament of Japan: the nuclear non-proliferation policy of the United States and the introduction of light water reactors into Japan, 1964-1968].

Science.gov (United States)

Yamazaki, Masakatsu

2014-07-01

Japan and the United States signed in 1968 a new atomic energy agreement through which US light-water nuclear reactors, including those of the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company, were to be introduced into Japan. This paper studies the history of negotiations for the 1968 agreement using documents declassified in the 1990s in the US and Japan. After the success of the Chinese nuclear test in October 1964, the United States became seriously concerned about nuclear armament of other countries in Asia including Japan. Expecting that Japan would not have its own nuclear weapons, the US offered to help the country to demonstrate its superiority in some fields of science including peaceful nuclear energy to counter the psychological effect of the Chinese nuclear armament. Driven by his own political agenda, the newly appointed Prime Minister Eisaku Sato responded to the US expectation favorably. When he met in January 1965 with President Johnson, Sato made it clear that Japan would not pursue nuclear weapons. Although the US continued its support after this visit, it nevertheless gave priority to the control of nuclear technology in Japan through the bilateral peaceful nuclear agreement. This paper argues that the 1968 agreement implicitly meant a strategic measure to prevent Japan from going nuclear and also a tactic to persuade Japan to join the Nuclear Non -Proliferation Treaty.

Development of remote decontamination technologies improving internal environment of reactor buildings at Fukushima Daiichi Nuclear Power Station

International Nuclear Information System (INIS)

Hotta, Koji; Hayashi, Hirotada; Sakai, Hitoshi

2016-01-01

The reactor buildings at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc., which was seriously damaged by the Great East Japan Earthquake of March 11, 2011, have been highly contaminated by radioactive materials. To safely and efficiently advance the processes related to the forthcoming decommissioning of the reactors, it is necessary to improve the hazardous environment inside the reactor buildings. During the more than four years that have elapsed since the Great East Japan Earthquake, Toshiba has been implementing various measures to reduce the ambient dose rates inside the reactor buildings through decontamination work and participation in a national project for the development of remote decontamination technologies for reactor buildings. A variety of vehicles and technologies to support decontamination work have been developed through these activities, and are significantly contributing to improvement of the environment inside the reactor buildings. (author)

Representativeness elements of an hybrid reactor demonstrator

International Nuclear Information System (INIS)

Kerdraon, D.; Billebaud, A.; Brissot, R.; David, S.; Giorni, A.; Heuer, D.; Loiseaux, J.M.; Meplan, O.

2000-11-01

This document deals with the quantification of the minimum thermal power level for a demonstrator and the definition of the physical criteria which define the representative character of a demonstrator towards a power reactor. Solutions allowing to keep an acceptable flow in an industrial core, have also been studied. The document is divided in three parts: the representativeness elements, the considered solutions and the characterization of the neutrons flows at the interfaces and the dose rates at the outer surface of the vessel. (A.L.B.)

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

in the regions of Republic of Kazakhstan occurs. Southern and western regions import electric power and capacity because of undeveloped circuit of networks. Moreover, power intensity of an industrial-agrarian complex of the country is limited transmission capacity of lines is insufficient; plenty of small consumers are removed from power supply lines. Thus, nuclear stations of medium and low power are the most acceptable for construction in Kazakhstan. Recommendations for the choice of maximum safe, reliable and economically competitive reactors for Kazakhstan have been made in result of the carried out projects' comparison of the power reactors according to 15 criteria of safety and economic competitiveness, with respect to condition and perspectives of Kazakhstan power complex development: Recommended power reactors of medium capacity: - P-600 - passive PWR, developed of the Westinghouse company, USA; - CANDU-6, developed by Atomic Energy of Canada, Limited (AECL), Canada. - MS-600 - Mitsubishi Company, Japan. Recommended reactors of low power: - IRIS - reactor of IV generation developed by the international corporation of 13 organizations from 7 countries; - NPP 'UNITERM' - development NIKIET, Moscow, Russia; - MRX - the Project of sea reactor MRX for civil applications, is developed by the Japanese Research Institute of Atomic Energy (JAERI). The most important advantages of recommended medium and low power reactors are given

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

regions of Republic of Kazakhstan occurs. Southern and western regions import electric power and capacity because of undeveloped circuit of networks. Moreover, power intensity of an industrial-agrarian complex of the country is limited transmission capacity of lines is insufficient; plenty of small consumers are removed from power supply lines. Thus, nuclear stations of medium and low power are the most acceptable for construction in Kazakhstan. Recommendations for the choice of maximum safe, reliable and economically competitive reactors for Kazakhstan have been made in result of the carried out projects' comparison of the power reactors according to 15 criteria of safety and economic competitiveness, with respect to condition and perspectives of Kazakhstan power complex development: Recommended power reactors of medium capacity: - P-600 - passive PWR, developed of the Westinghouse company, USA; - CANDU-6, developed by Atomic Energy of Canada, Limited (AECL), Canada. - MS-600 - Mitsubishi Company, Japan. Recommended reactors of low power: - IRIS - reactor of IV generation developed by the international corporation of 13 organizations from 7 countries; - NPP 'UNITERM' - development NIKIET, Moscow, Russia; - MRX - the Project of sea reactor MRX for civil applications, is developed by the Japanese Research Institute of Atomic Energy (JAERI). The most important advantages of recommended medium and low power reactors are given

A review on the utilization of the Japan materials testing reactor (JMTR)

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Kang, Y. H.; Kim, B. G.; Choo, K. N.; Oh, J. M.; Park, S. J.; Shin, Y. T

1999-04-01

The HANARO has possessed the potential capability for the testing of materials and fuels since the beginning of its operation in 1995. Recently, this reactor has contributed to various activities in nuclear power research in Korea. We need the recent technical data of developed countries to support these activities in nuclear power. Most of the developed countries in nuclear power have more than thirty years' experience in the irradiation test of nuclear fuel and material for performing their complicated in-core measurements of the change of material properties. They also have developed various types of sensors, equipment and techniques. This report describes the status of utilization of the irradiation facilities of the Japan Materials Testing Reactor(JMTR). It also describes the recent efforts of the JMTR in order to develop new irradiation test techniques. It will be our great pleasure for this report to help a broad range of people understand the generic contents (JMTR utilization, new techniques) of the JMTR. (author)

A review on the utilization of the Japan materials testing reactor (JMTR)

International Nuclear Information System (INIS)

Kim, D. H.; Kang, Y. H.; Kim, B. G.; Choo, K. N.; Oh, J. M.; Park, S. J.; Shin, Y. T.

1999-04-01

The HANARO has possessed the potential capability for the testing of materials and fuels since the beginning of its operation in 1995. Recently, this reactor has contributed to various activities in nuclear power research in Korea. We need the recent technical data of developed countries to support these activities in nuclear power. Most of the developed countries in nuclear power have more than thirty years' experience in the irradiation test of nuclear fuel and material for performing their complicated in-core measurements of the change of material properties. They also have developed various types of sensors, equipment and techniques. This report describes the status of utilization of the irradiation facilities of the Japan Materials Testing Reactor(JMTR). It also describes the recent efforts of the JMTR in order to develop new irradiation test techniques. It will be our great pleasure for this report to help a broad range of people understand the generic contents (JMTR utilization, new techniques) of the JMTR. (author)

The outlook for nuclear power development in Japan

International Nuclear Information System (INIS)

Hiraiwa, Gaishi

1987-01-01

The world economy has entered a new stage of growth--albeit low growth--following painful adjustments in the wake of past oil crises. At the same time, energy demand is expanding at an even slower rate, due to the structural changes in industry and improved efficiency in energy use. Furthermore, progress in the development of alternative energies and technical innovations in both the supply and use of energy have sharpened competition between energy sources. We also aim to improve even further the economy of nuclear power, within the bounds of safety and reliability, to minimize electric power generation costs by optimizing the total system for nuclear power generation including the nuclear fuel cycle. In Japan's long-term strategy for the development of nuclear power, our basic plan is to switch from light-water reactors to fast-breeder reactors (FBR), as the latter use plutonium most efficiently. Every effort is being made to have FBR reactors up and running at an early date. However, given the outlook for the development of their technology and the supply and demand situation for uranium, we estimate that this won't be achieved until 2020 or 2030. With this timetable in mind, it will be important to prepare for the coming age of FBR by mastering the technologies of and establishing the foundation necessary for plutonium utilization. To this end, we plan to expand our use of plutonium to an appropriate scale, at the earliest possible date. (J.P.N.)

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

International Nuclear Information System (INIS)

Yamamoto, Nobuo; Izumi, Fumio.

1995-12-01

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

Research and development for high temperature gas cooled reactor in Japan

International Nuclear Information System (INIS)

Taketani, K.

1978-01-01

The paper describes the current status of High Temperature Gas Cooled Reactor research and development work in Japan, with emphasis on the Experimental Very High Temperature Reactor (Exp. VHTR) to be built by Japan Atomic Energy Research Institute (JAERI) before the end of 1985. The necessity of construction of Exp. VHTR was explained from the points of Japanese energy problems and resources

Review of the activities in Japan

International Nuclear Information System (INIS)

Otake, I.

1982-01-01

The fast breeder reactor development project in Japan has been in progress through.operation of the experimental fast reactor JOYO, design of the prototype fast breeder reactor MONJU and related R and D works. JOYO began operation in mid-1977, increased power from 50 MWt to 75 MWth in July 1979 and operation cycles at 75 MWth are continued at present. With respect to MONJU, which is a 300 MWe plant, progress toward construction has been made and the safety review are started by the concerned authorities. Conceptual design studies of large demonstration fast breeder reactor are also being made by PNC. It is a 1000 MWe, loop type plant

APWR - Mitsubishi, Japan/Westinghouse, USA

International Nuclear Information System (INIS)

Aeba, Y.; Weiss, E.H.

1999-01-01

Nuclear power generated by light water reactors accounts for approximately 1/3 of Japan's power supply. Development of the Advanced Pressurized Water Reactor (APWR) was initiated by five PWR electric power companies (Hokkaido, Kansai, Shikoku, Kyushu and Japan Atomic Power), Mitsubishi Heavy Industries, and Westinghouse, with a view to providing a nuclear power source to meet future energy demand in Japan. The APWR was developed based on the results of the Improvement and Standardization Program, promoted by the Ministry of International Trade and Industry, with reconsideration of the needs of age, such as construction cost reduction, enhanced safety and increased reliability. One of the important concepts of the APWR is its large power rating that decreases the construction cost per unit of electric generation capacity. Though the electric output was lower at the early stage of basic design than it is now, uprating to approximately 1530 MW is achieved based on the results of design progress and high efficiency improvements to the steam turbine and reactor coolant pumps. Furthermore, the APWR remarkably enhances reliability, safety operability and maintainability by introducing new technologies that include a radial reflector and advanced accumulators. The first APWR is planned to be built at Tsuruga No. 3 and No. 4 by the Japan Atomic Power Company and will be the largest commercial operation plant in the early 21st century. (author)

IAEA Issues Report on Mission to Review Japan's Nuclear Power Plant Safety Assessment Process

International Nuclear Information System (INIS)

2012-01-01

Full text: A team of international nuclear safety experts has delivered its report on a mission it conducted from 21-31 January 2012 to review Japan's process for assessing nuclear safety at the nation's nuclear power plants. International Atomic Energy Agency (IAEA) officials delivered the IAEA Mission Report to Japanese officials yesterday and made it publicly available today. Following the 11 March 2011 accident at TEPCO's Fukushima Daiichi Nuclear Power Station, Japan's Nuclear and Industrial Safety Agency (NISA) announced the development of a revised safety assessment process for the nation's nuclear power reactors. At the request of the Government of Japan, the IAEA organized a team of five IAEA and three international nuclear safety experts and visited Japan to review NISA's approach to the Comprehensive Assessments for the Safety of Existing Power Reactor Facilities and how NISA examines the results submitted by nuclear operators. A Preliminary Summary Report was issued on 31 January. 'The mission report provides additional information regarding the team's recommendations and overall finding that NISA's instructions to power plants and its review process for the Comprehensive Safety Assessments are generally consistent with IAEA Safety Standards', said team leader James Lyons, Director of the IAEA's Nuclear Installation Safety Division. National safety assessments and their peer review by the IAEA are a key component of the IAEA Action Plan on Nuclear Safety, which was approved by the Agency's Member States following last year's nuclear accident at Fukushima Daiichi Nuclear Power Station. The IAEA safety review mission held meetings in Tokyo with officials from NISA, the Japanese Nuclear Energy Safety Organization (JNES), and the Kansai Electric Power Company (KEPCO), and the team visited the Ohi Nuclear Power Station to see an example of how Japan's Comprehensive Safety Assessment is being implemented by nuclear operators. In its report delivered today

Development step toward fusion power plant and role of experimental reactor ITER

International Nuclear Information System (INIS)

Hiwatari, Ryouji; Asaoka, Yoshiyuki; Okano, Kunihiko

2005-01-01

The development of fusion energy is going into the experimental reactor stage, and the thermal energy from the fusion reaction will be generated in a plant scale through the ITER (International Thermonuclear Experimental Reactor) project. The remaining critical issue toward the realization of fusion energy is to map out the development strategy. Recently early realization approach as for the fusion energy development is being discussed in Japan, Europe, and the United States. This approach implies that the devices for a Demo reactor and a proto-type reactor as seen in the fast breeder reactor are combined into a single device in order to advance the fusion energy development. On the other hand, a clear development road map for fusion energy hasn't been suggested yet, and whether that early realization approach is feasible or not is still ambiguous. In order to realize the fusion energy as an user-friendly energy system, the suggestion of the development missions and the road map from the user-side point of view is instructive not only to Japanese but also to other country's development policy after the ITER project. In this report, first of all, the development missions from the user's point of view have been structured. Second, the development target required to demonstrate net electric generation and to introduce the fusion energy into the market is investigated, respectively. This investigation reveals that the completion of the ITER reference operation gives the outlook toward the demonstration of net electric generation and that the completion of the ITER advanced operation gives the possibility to introduce the fusion energy into the market. At last, the electric demonstration power plant Demo-CREST and the commercial power plant CREST are proposed to construct the development road map for fusion energy. (author)

Gaseous fuel reactors for power systems

International Nuclear Information System (INIS)

Helmick, H.H.; Schwenk, F.C.

1978-01-01

The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

International nuclear reactor hazard study. Design and operational features, and hazards of commercial nuclear power reactors in the world. Vol. 1

International Nuclear Information System (INIS)

Anderson, R.; Benecke, J.; Donderer, R.; Etemad, S.; Hirsch, H.; Kjellstroem, B.; Large, J.; Scheer, J.; Schneider, M.; Schumacher, H.; Schumacher, O.; Scott, M.; Takagi, J.; Thompson, G.; Torrie, R.; Ziggel, H.

1986-09-01

Project management and scientific coordination was performed by Gruppe Oekologie Hannover. The experts from Canada, the Federal Republic of Germany, France, Great Britain, Sweden, USA, and Japan represented an annulated experience of work on nuclear safety. The panel members analyzed the reactor types of their respective countries. The draft reports on the individual reactor types were discussed. Also included are brief descriptions of the different reactor types with simple drawing, evolution of reactor types and a brief report on nuclear power plants in the world. (DG)

Historical construction costs of global nuclear power reactors

International Nuclear Information System (INIS)

Lovering, Jessica R.; Yip, Arthur; Nordhaus, Ted

2016-01-01

The existing literature on the construction costs of nuclear power reactors has focused almost exclusively on trends in construction costs in only two countries, the United States and France, and during two decades, the 1970s and 1980s. These analyses, Koomey and Hultman (2007); Grubler (2010), and Escobar-Rangel and Lévêque (2015), study only 26% of reactors built globally between 1960 and 2010, providing an incomplete picture of the economic evolution of nuclear power construction. This study curates historical reactor-specific overnight construction cost (OCC) data that broaden the scope of study substantially, covering the full cost history for 349 reactors in the US, France, Canada, West Germany, Japan, India, and South Korea, encompassing 58% of all reactors built globally. We find that trends in costs have varied significantly in magnitude and in structure by era, country, and experience. In contrast to the rapid cost escalation that characterized nuclear construction in the United States, we find evidence of much milder cost escalation in many countries, including absolute cost declines in some countries and specific eras. Our new findings suggest that there is no inherent cost escalation trend associated with nuclear technology. - Highlights: •Comprehensive analysis of nuclear power construction cost experience. •Coverage for early and recent reactors in seven countries. •International comparisons and re-evaluation of learning. •Cost trends vary by country and era; some experience cost stability or decline.

Conceptual design studies of experimental and demonstration fusion reactors

International Nuclear Information System (INIS)

1978-01-01

Since 1973 the FINTOR Group has been involved in conceptual design studies of TOKAMAK-type fusion reactors to precede the construction of a prototype power reactor plant. FINTOR-1 was the first conceptual design aimed at investigating the main physics and engineering constraints on a minimum-size (both dimensions and thermal power) tokamak experimental reactor. The required plasma energy confinement time as evaluated by various power balance models was compared with the values resulting from different transport models. For the reference design, an energy confinement time ten times smaller than neoclassical was assumed. This also implied a rather high (thermally stable) working temperature (above 20 keV) for the reactor. Other relevant points of the design were: circular plasma cross section, single-null axisymmetric divertor; lithium breeder, stainless steel structures, helium coolant; modular blanket and shield structure; copper-stabilized, superconducting Nb-Ti toroidal field and divertor coils; vertical field and transformer coils inside the toroidal coils; vacuum-tight containment vessel. Solutions involving air and iron transformer cores were compared. These assumptions led to a minimum size reactor with a thermal power of about 100MW and rather large dimensions (major radius of about 9m) similar to those of full-scale power reactors considered in other conceptual studies. The FINTOR-1 analysis was completed by the end of 1976. In 1977 a conceptual design of a Demonstration Power Reactor Plant (FINTOR-D) was started. In this study the main working assumptions differing from those of FINTOR-1 are: non-circular plasma cross section; plasma confinement compatible with trapped ion instabilities; cold (gas) blanket sufficient for wall protection (no divertor); wall loading between 1-3MW/m 2 and thermal power of a few GW. (author)

The disappointments for nuclear energy in Japan

International Nuclear Information System (INIS)

2004-01-01

Several dysfunctions are reported in this paper: A reactor (Onagawa) closed after a nitrogen leakage; a small leakage of radioactive water in the nuclear power plant of Mihama assessment raised to five deaths, the operator stops its nuclear power plants for inspection, the Japan face to its ageing nuclear power plants, the truth about the cost of M.O.X., the seven reactors of Japan closed for inspection after cracks and leaks hidden to authorities, Tokai MURA accident. (N.C.)

New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

International Nuclear Information System (INIS)

Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

1996-01-01

The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

Conceptual design study for the demonstration reactor of JSFR. (3) Safety design and evaluation

International Nuclear Information System (INIS)

Tani, Akihiro; Shimakawa, Yoshio; Kubo, Shigenobu; Fujimura, Ken; Yamano, Hidemasa

2011-01-01

This paper describes the result of conceptual safety design and evaluation for the demonstration plant of Japan sodium-cooled fast reactor (JSFR), which was preliminarily conducted for providing information necessary to decide the plant specification for further design study. The plant major specifications except for output power and safety design concept are almost the same as those of the commercial JSFR. A set of safety evaluation for typical design basis events (DBEs) is mainly focused here, which was conducted for the 750 MWe design. Safety analyses for DBEs evaluation were performed on the basis of conservative assumptions using a one-dimensional flow network code with point kinetics. For representative DBEs, transient over power type events and loss of flow type events were analyzed. The long-term loss-of-offsite power event was also calculated to evaluate the natural circulation decay heat removal system. All analytical results showed to meet tentative safety criteria, thus it was confirmed that the safety design concept of JSFR is feasible against DBEs. (author)

Experience with, and programme of, FBR and HWR development in Japan

International Nuclear Information System (INIS)

Iida, M.; Sawai, S.; Nomoto, S.

1983-01-01

Nuclear power generation in Japan is moving forward on the long-term development programme of nuclear power from the LWR to the FBR, essentially in the same way as in other advanced nuclear countries. In this development programme the unique HWR is also included; it can use plutonium produced in LWRs together with depleted uranium before the introduction of commercial FBRs. This report describes the status of the FBR and HWR development project being carried out by the Power Reactor and Nuclear Fuel Development Corporation (PNC) based upon the Long-Term Programme on Research, Development and Utilization of Nuclear Energy in Japan. Operational experience and technical results are shown for the experimental fast reactor JOYO (100 MW(th)), which reached initial criticality in 1977. The status of the 280 MW(e) prototype reactor MONJU, under construction as of 1982, is described. The conceptual design of the subsequent 1000 MW(e) demonstration plant is outlined, as is additional future planning. Research and development results, mainly carried out at Oarai Engineering Center of PNC, are shown. The 165 MW(e) prototype FUGEN is a heavy-water-moderated, boiling-light-water-cooled, pressure-tube-type reactor which uses plutonium mixed-oxide fuel. This report describes the relationship of the fuel cycle to the HWR in Japan and also discusses the operational experience of the prototype FUGEN, which has operated since 1979. Also described is the design of the 600 MW(e) demonstration plant and the programme of related research and development. (author)

Nuclear power supply (Japan Nuclear Safety Institute)

International Nuclear Information System (INIS)

Kameyama, Masashi

2013-01-01

After experienced nuclear disaster occurred on March 11, 2011, role of nuclear power in future energy share in Japan became uncertain because most public seemed to prefer nuclear power phase out to energy security or costs. Whether nuclear power plants were safe shutdown or operational, technologies were requisite for maintaining their equipment by refurbishment, partly replacement or pressure proof function recovery works, all of which were basically performed by welding. Nuclear power plants consisted of tanks, piping and pumps, and considered as giant welded structures welding was mostly used. Reactor pressure vessel subject to high temperature and high pressure was around 200mm thick and made of low-alloy steels (A533B), stainless steels (308, 316) and nickel base alloys (Alloy 600, 690). Kinds of welding at site were mostly shielded-metal arc welding and TIG welding, and sometimes laser welding. Radiation effects on welding of materials were limited although radiation protection was needed for welding works under radiation environment. New welding technologies had been applied after their technical validation by experiments applicable to required regulation standards. Latest developed welding technologies were seal welding to prevent SCC propagation and temper-bead welding for cladding after removal of cracks. Detailed procedures of repair welding of Alloy 600 at the reactor outlet pipe at Oi Nuclear Power Plants unit 3 due to PWSCC were described as an example of crack removal and water jet peening, and then overlay by temper-bead welding using Alloy 600 and clad welding using Alloy 690. (T. Tanaka)

Status of core material development for fast reactor in Japan

International Nuclear Information System (INIS)

Ukai, S.; Shibahara, I.; Nagai, S.

1994-01-01

In the last two decades, extensive efforts have been devoted to the development of mixed-oxide fuel for LMFBR in Japan. For the fuel of the prototype reactor MONJU, drastic improvement in creep rupture strength and swelling resistance were attained by modification within the compositional specification of the standard Type 316 stainless steel (PNC316). For the fuel of future large-scale reactors, extensive research and development program are under way to realize the long life fuel. The candidate material for demonstration reactor is advanced austenitic stainless steel (PNC1520) which intended to modify the composition beyond the Type 316 stainless steel specification. In order to further improve the swelling resistance, the austenitic stainless steel with higher nickel content (High Ni alloy) and ferritic/martensitic steel (PNC-FMS) are developed. In a prospective cladding material for the long life fuel, the development of oxide dispersion strengthened (ODS) ferritic steel is focused to establish the alloying design and fabrication process toward as high as 250dpa. (author)

Current Status and Future Outlook of Nuclear Power Generation in Japan

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Yasuro; Yoshii, Ryosuke

2007-07-01

For Japan, a country poor in natural resources, in light of the tough energy situation in recent times, a National Energy Strategy with energy security at its core was established in May 2006. The key point of the Strategy is nuclear power generation, and the aim is to ensure that nuclear power generation continues to account for 30 to 40 percent or more of total electricity generated even after 2030. The first step to achieving this goal is to make maximum use of existing plants (55 plants, 49580MWe), and the aim is to achieve a 60-year service life by making improvements to plant operation and maintenance, such as extending current monitoring and maintenance of plant condition, and the implementation of plant aging management. In Japan, plant construction has been continuous since the 1970s. The current new plant construction plan (13 plants, 17230MWe) is to be achieved with a concerted, cohesive national effort. In addition, in order to complete the nuclear fuel cycle, a reprocessing plant is being constructed strictly for peaceful use, and construction of a site for disposing of high-level radioactive waste is also proceeding. Development of the next generation light water reactors and fast breeder reactor cycle is also underway. (auth)

Reactor power control device

International Nuclear Information System (INIS)

Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

1998-01-01

The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

A review of fast reactor progress in Japan

Energy Technology Data Exchange (ETDEWEB)

Tomabechi, K [Power Reactor and Nuclear Fuel Development Corporation, Tokyo (Japan)

1978-07-01

The fast reactor development project in Japan is continuing at a slightly increased scale of effort in budget. The total budget for LMFBR development for fiscal year 1978 was 24 billion yen. In August 1977 major industries engaged in LMFBR have set up an office where design work can be jointly conducted. Highlights and topics of the fast reactor development activities cover description of JOYO reactor, its first criticality experiment, and the prototype fast breeder MONJU. Research and development programmes dealt with fission products release and its possible interaction with the soodium coolant, inspection of reactor components, experiments simulating sodium leakage, development of steam generator.

Construction of the advanced boiling water reactor in Japan

International Nuclear Information System (INIS)

Natsume, Nobuo; Noda, Hiroshi

1996-01-01

The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7

Guide to power reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1959-07-15

The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

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)

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

International Nuclear Information System (INIS)

Sawai, Sadamu.

1982-03-01

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

Reactor power measuring device

International Nuclear Information System (INIS)

Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

1993-01-01

The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

Progress report on fast breeder reactor development in Japan

International Nuclear Information System (INIS)

1980-01-01

The performance test on the reactor power increase to 75 MW was started on July 3, and the target of 75 MW was reached on July 16, in the experimental fast reactor Joyo. The tests on the heat transport characteristics, power coefficient, the response to the change of outlet temperature, the loss of external power supply and so on were carried out, and the performance test was finished on August 23, except the test of 75 MW continuous operation. The annual inspection of the systems is being carried out in parallel with the regular inspection. The design to prepare for the manufacture of the prototype fast reactor Monju is being prepared. The analysis of decay heat removal is being carried out, and various calculation codes were developed. The technological survey on overseas LMFBRs is being made. The conceptual design of the demonstration reactor is being prepared. The research and development of reactor physics, structural components for Joyo and Monju, instrumentation and control, sodium technology, fuel materials, structural materials, safety problems and steam generators are reported. The tests on the transient boiling of sodium, fuel failure propagation, heat transfer between molten materials, post-accident decay heat removal and so on have been carried out. (Kako, I.)

Operation, test, research and development of the high temperature engineering test reactor (HTTR). (FY2005)

International Nuclear Information System (INIS)

2007-03-01

The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power. The full power operation of 30 MW was attained in December, 2001, and then JAERI (JAEA) received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. In fiscal 2005 year, periodical inspection and overhaul of reactivity control system were conducted, and safety demonstration tests were promoted. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2005. (author)

A state-of-the art on decommissioning of nuclear facilities in Japan

International Nuclear Information System (INIS)

Park, Seung Kook; Kim, Hee Reyoung; Chung, Un Soo; Jung, Ki Jung

2002-05-01

While proceeding the KRR-1 and 2 decommissioning project, we are carried out study for the state of the art on decommissioning of nuclear facilities in Japan. Also, we are studied for the research reactors and commercial power plant that has the object of decommissioning, and for the government and the organization related on decommissioning operation. We are investigated for decommissioning activities of nuclear facilities achieved by JAERI, and collected the information and data for decommissioning techniques and computational system through the JPDR(Japan Power Demonstration Reactor) decommissioning activities. Such techniques are applying for Tokai Power Station began the decommissioning project from last year, and for Fugen Nuclear Power Station to be planned the decommissioning from 2003. Recent techniques for decommissioning was acquired by direct contact. The status of the treatment for decommissioning waste and the disposal facility for the very low-level radioactive concrete wastes was grasped

Completion of reconstruction for Japan Research Reactor No.3

International Nuclear Information System (INIS)

Kakefuda, K.; Tani, M.; Isshiki, M.

1992-01-01

The works of the reconstruction for the Japan Research Reactor No.3 (JRR-3) started in 1985 and initial criticality of the new reactor achieved in March, 1990. After commissioning test, the new JRR-3 has been operated some operational cycles since November, 1990. This paper presents outline of the removal work on the old JRR-3 and the new JRR-3. (author)

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Progress Report on Sodium Cooled Fast Breeder Reactor Development in Japan, April 1975

International Nuclear Information System (INIS)

Tomabechi, K.

1975-01-01

The progress of the sodium cooled fast Breeder Reactor development in Japan in the past 12 months can be summarized as follows. Installation of all the components of the Experimental Fast Reactor, ''JOYO'', was completed in the end of the last year and various commissioning tests of the reactor began in January 1975. It is planned to charge sodium into the reactor in coming fall and the first criticality experiment is currently planned in the summer 1976. Most of the research and development works for ''JOYO'' are nearing completion. These include an endurance test of 3 prototype primary sodium pump for 12,000 hours. 86 core fuel subassemblies and 220 blanket subassemblies, a sufficient number for composing the initial core, have already been fabricated. Concerning the Prototype Fast Breeder Reactor, ''MONJU'', design activity as well as relevant research and development works are continued. A siting problem exists and it is hoped to be resolved soon. Of the research and development works, a significant achievement in the past 12 months can be a successful operation at full power of the 50 MW Steam Generator Test Facility. This facility was put into operation at full power in June 1974. No leak of water into sodium has been experienced with operation of the steam generator tested. The steam generator is being dismantled for a detailed inspection originally planned

Past and present role of fast breeder reactors in Japan

International Nuclear Information System (INIS)

Tomabechi, K.

1978-01-01

To meet increasing future energy demand in Japan utilization of fission energy should be promoted. In particular it is of vital importance to develop and utilize FBRs as soon as possible in order to save the natural uranium needed. If one considers the commercial introduction of FBRs in the mid-1990s in Japan, a delay of only one year will eventually result in an additional demand for natural uranium of more than 20,000 tonnes, because several LWRs will have to be installed instead. Ten years have passed since the development of FBRs in Japan was initiated as a national project with the highest priority and now the experimental fast reactor JOYO is successfully being commissioned and the prototype fast breeder reactor MONJU has reached the stage of proceeding to construction. Experience gained from past development work encourages further strong effort towards the commercialization of FBRs. This paper briefly reviews the important role to be played by FBRs, the development programme for FBRs, and the experience gained so far from the programme in Japan. (author)

An approach of raising the low power reactor trip block (P-7) in Maanshan Power Plant

International Nuclear Information System (INIS)

Wang, L.C.

1984-01-01

The technical specification for the Maanshan Nuclear Power Station (FSAR Table 16.2.2-3) requires that with an increasing reactor power level above the setpoint of low power reactor trip block (P-7), a turbine trip shall initiate a reactor trip. This anticipatory reactor trip on turbine trip prevents the pressurizer PORV from openning during turbine trip event. In order to reduce unnecessary reactor trip due to turbine trip on low reactor power level during Maanshan start-up stage, Taiwan Power Company performed a transient analysis for turbine trip event by using RETRAN code. The highest reactor power level at which a turbine trip will not open the pressurizer PORV is searched. The results demonstrated that this power level can be increased from the original value-10% of the rated thermal power-to about 48% of the rated thermal power

Reduction of releases of radioactive effluents from light-water-power-reactors in Japan

International Nuclear Information System (INIS)

Yoshida, Y.; Itakura, T.; Kanai, T.

1977-01-01

Japan Atomic Energy Commission established the dose objectives to the population around the light-water-reactors in May, 1975, based on the ''ALAP'' concept. These values are respectively, 5 mrems per year for total body and 15 mrems per year for thyroid of an individual in the critical group in the environs, due to both gaseous and liquid effluents from LWRs in one site. The present paper describes the implications of the dose objective values, control measures which have been adopted to reduce releases of radioactive materials and related technical developments in Japan. The main control measures for reduction of radioactive gaseous effluents are an installation of a charcoal gas holdup system for decay of noble gases and a supply of clean steam for the gland seal of a turbine in BWR, and a storage tank system allowing decay of noble gases in PWR. For liquid effluents are taken measures to re-use them as the primary coolant. Consequently, the amounts of radioactivity released to the environment from any LWR during normal operation have been maintained under the level to meet the above dose objective values. For research reactors, reduction of release of effluents has also been carried out in a similar way to LWRs. In order to establish the techniques applicable for further reduction, studies are being made on the control measures to reduce leakage of radioiodine, an apparatus for removal of krypton, the treatment of laundry waste and measures to remove the crud in the primary coolant. Presentation is also made on the energy-integrated gas monitor for gaseous effluent and systems of measuring γ dose from radioactive cloud descriminating from natural background, which have been developed for effective monitoring thus reduced environmental dose

Summary of Great East Japan Earthquake response at Onagawa Nuclear Power Station and further safety improvement measures

International Nuclear Information System (INIS)

Sato, Toru

2013-01-01

A large earthquake occurred on March 11, 2011 and tsunami was generated following it. The East Japan suffered serious damage by the earthquake and tsunami. This is called the Great East Japan Earthquake. Onagawa Nuclear Power Station (NPS) is located closest to the epicenter of Great East Japan Earthquake. We experienced intense shake by the earthquake and some flooding from the tsunami, however, we have succeeded safely cold shutdown of the reactors. In this paper, we introduce summary of Great East Japan Earthquake response a Onagawa NPS and safety improvement measures which are based on both experience of Onagawa NPS and lesson from Fukushima Daiichi NPS accident. (author)

Conceptual design of the JAERI demonstration fusion reactor

International Nuclear Information System (INIS)

Sako, K.; Tone, T.; Seki, Y.

1976-01-01

Conceptual design of a tokamak demonstration fusion reactor is carried out. This design is an extended and improved version of the previous design which was presented at the 5th IAEA Conference. The main design parameters are as follows: the reactor thermal power 2000 MW, torus radius 10.5 m, plasma radius 2.7 m, first wall radius 3.0 m, toroidal magnetic field on axis 6T, blanket fertile material Li 2 O, coolant He, structural material Mo-alloy and tritium breeding ratio 1.2

Evolution of on-power fuelling machines on Canadian natural uranium power reactors

International Nuclear Information System (INIS)

Isaac, P.

1984-10-01

The evolution of the on-power fuel changing process and fuelling machines on CANDU heavy-water pressure tube power reactors from the first nuclear power demonstration plant, 22 MWe NPD, to the latest plants now in design and development is described. The high availability of CANDU's is largely dependent on on-power fuelling. The on-power fuelling performance record of the 16 operating CANDU reactors, covering a 22 year period since the first plant became operational, is given. This shows that on-power fuel changing with light (unshielded), highly mobile and readily maintainable fuelling machines has been a success. The fuelling machines have contributed very little to the incapabilities of the plants and have been a key factor in placing CANDUs in the top ten list of world performance. Although fuel handling technology has reached a degree of maturity, refinements are continuing. A new single-ended fuel changing concept for horizontal reactors under development is described. This has the potential for reducing capital and operating costs for small reactors and increasing the fuelling capability of possible large reactors of the future

Reactor experiments, workshops, and human resource development education simulating the Great East Japan Earthquake

International Nuclear Information System (INIS)

Horiguchi, Tetsuo; Yamamoto, Tomosada

2012-01-01

Kinki University Atomic Energy Research Institute has been implementing a social education program such as reactor experiments and training sessions for junior and senior high school teachers since 1987, and in recent years, it has been implementing an education program for common citizens. However, the Great East Japan Earthquake has made it necessary to consider not only the dissemination of accurate knowledge, but also responding to the anxiety on nuclear power. This paper explains the contents of the social contribution activities and workshops conducted at Kinki University Atomic Energy Research Institute, after the Great East Japan Earthquake and the Fukushima Daiichi Nuclear Power Station accident. As the activities that are carried out in addition to training sessions, it introduces the implementation state of telephone consultation about nuclear power, and earthquake reconstruction assistance advisory at Kawamata Town, Date-gun, Fukushima Prefecture. As workshop support, it reports human resource development education in the nuclear field at the university, activities at the workshops for junior/senior high school teachers and general public, and questionnaire survey at the time of the workshops. (A.O.)

Statement to International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, 7 December 2009, Kyoto, Japan

International Nuclear Information System (INIS)

Amano, Yukiya

2009-01-01

Full text: Distinguished Guests, Ladies and Gentlemen, It is my honour to address participants at this opening session of the International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, organized by the IAEA and hosted by the Japan Atomic Energy Agency. Fast reactor technology has the potential to ensure that energy resources which would last hundreds of years with the technology we are using today will actually last several thousand years. In other words it can withstand enormous increases in demand. This innovative technology also reduces the risk to the environment and helps to limit the burden that will be placed on future generations in the form of waste products. The coming year will be an exciting one for the development of fast-spectrum nuclear reactors. We expect to reach many important milestones: - the first criticality of the China Experimental Fast Reactor; - the restart of the Monju prototype fast reactor in Japan; and - the new insights we will gain through the end-of-life studies at the Phenix reactor in France. In the near future, new fast reactors will be commissioned: the 500MW(e) Prototype Fast Breeder Reactor in India, the first in a series of five of the same type, and the BN-800 reactor in the Russian Federation. Moreover, France, Japan, India, China and the Republic of Korea are preparing advanced prototypes, demonstration or commercial reactors for the 2020-2030 period. Nuclear power is set to be an increasingly important part of the global energy mix in the coming decades as demand for energy grows. Scores of countries in both the developed and developing world have told the IAEA that they are interested in introducing nuclear power. The 30 countries which already have nuclear power reactors are set to build more. This trend is likely to be accompanied by accelerated deployment of fast reactors. Continued advances in research and technology development are necessary to ensure improved economics and

Current status and future perspective of upgrading on light water reactors in Japan

International Nuclear Information System (INIS)

Tateki, Hoshion

1991-01-01

Energy demands in the world would be expected to increase constantly furthermore. In this situation the role of nuclear energy has become more important than it used to be from the view of the countermeasure for the global warming issue resulting from the exhausted gases causing the greenhouse effect in addition to the view of the energy securities. In also Japan, the energy demand-supply perspective plan issued on October,1990 which aims to expand the capacity of nuclear power plants to 72,500MWe as a target. by 2010. (see table.l) Present Light Water Reactors(LWRs) have been experienced and this type of reactor will remain as major stream of nuclear power generation for a long time all over the world. Considering above perspective, Advisory Committee for Energy of MITI(Ministry of International Trade and Industry) has investigated the technology and technical development of coming LWRs and concluded the report on June of this year. In this paper the contents of above report are introduced

An inherently safe power reactor module

International Nuclear Information System (INIS)

Salerno, L.N.

1985-01-01

General Electric's long participation in liquid metal reactor technology has led to a Power Reactor Inherently Safe Module (PRISM) concept supported by DOE contract DE-AC06-85NE37937. The reactor module is sized to maximize inherent safety features. The small size allows factory fabrication, reducing field construction and field QA/QC labor, and allows safety to be demonstrated in full scale, to support a pre-licensed standard commercial product. The module is small enough to be placed underground, and can be combined with steam and electrical generating equipment to provide a complete electrical power producing plant in the range of 400-1200 MWe. Initial assessments are that the concept has the potential to be economically competitive with existing methods of power production used by the utility industry

Development of nuclear powered ship in Japan

International Nuclear Information System (INIS)

Sato, Hiroshi

1976-01-01

The development of nuclear merchant ship in Japan was started in 1955 by the establishment of Nuclear Ship Study Group, and since then, the investigation, test and research on nuclear ships have been continued. As a result, a nuclear ocean observation and supply ship was designed for trial. Researches were carried out also in JAERI and Institute for Technical Research of Ships. Meanwhile, the nuclear icebreaker Lenin was completed in Soviet Union in 1959, the nuclear ship Savannah set out for maiden voyage in U.S. in 1962, and the construction of the nuclear ore carrier Otto Hahn was prepared in FRG. Japan Nuclear Ship Development Corp. was established in 1963, and started the design and construction of the first nuclear ship in Japan, Mutsu. The basic policy in the construction is the improvement of nuclear ship technology, the securing of safety, and the use of domestic technologies as far as possible. The progress of the design, construction and test of the Mutsu is described. Owing to the problem of radiation leak, the development of nuclear ships stagnated for a while, but the nuclear plant of the Mutsu demonstrated the expected performance in the functional test, land criticality test and zero output test, and it is expected that the bud of the independent development brought up so far can bear valuable fruit. The independent development of marine nuclear reactors should be continued by selecting the way most suitable to Japan. (Kako, I.)

Conceptual design for Japan sodium-cooled fast reactor. (1) Current status of system design for JSFR

International Nuclear Information System (INIS)

Uto, Nariki; Sakai, Takaaki; Mihara, Takatsugu; Kotake, Shoji; Aoto, Kazumi; Toda, Mikio

2009-01-01

Japan Atomic Energy Agency is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project. In the FaCT project, the system design for JSFR has been carried out along the design categories such as safety design, reactor system, heat transport system, etc., together with research and developments (R and Ds) on innovative technologies to be adopted to JSFR for achieving economic competitiveness, enhanced safety and reliability. This paper describes the system design features of JSFR and a summary of the progresses of the design and R and Ds concerned with a compact reactor vessel, an innovative containment vessel, etc. The approach for the commercialization of fast reactors including discussion on a demonstration reactor for JSFR is also briefly described. (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.)

Rise-to-power test in High Temperature Engineering Test Reactor. Test progress and summary of test results up to 30 MW of reactor thermal power

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

2002-08-01

The High Temperature Engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30 MW and the reactor outlet coolant temperature of 850degC/950degC. Rise-to-power test in the HTTR was performed from April 23rd to June 6th in 2000 as phase 1 test up to 10 MW in the rated operation mode, from January 29th to March 1st in 2001 as phase 2 test up to 20 MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20 MW in the high temperature test the mechanism of the reactor outlet coolant temperature becomes 850degC at 30 MW in the rated operation mode and 950degC in the high temperature test operation mode. Phase 4 rise-to-power test to achieve the thermal reactor power of 30 MW started on October 23rd in 2001. On December 7th in 2001 it was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30 MW and 850degC respectively in the single loaded operation mode in which only the primary pressurized water cooler is operating. Phase 4 test was performed until March 6th in 2002. JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests by MEXT were passed successfully with the reactor transient test at an abnormal event as a final pre-operation test. From the test results of the rise-up-power test up to 30 MW in the rated operation mode, performance of the reactor and cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely. Some problems to be solved were found through the tests. By solving them, the reactor operation with the reactor outlet coolant temperature of 950degC will be achievable. (author)

White paper on atomic energy, for 1974 and 1975. [Japan

Energy Technology Data Exchange (ETDEWEB)

1975-09-01

Nearly 20 years have passed since the initiation of peaceful uses of atomic energy in Japan. Close to the end of this period, there occurred the so-called oil crisis, which emphasized the need for nuclear power development. Meanwhile, voices of the people in Japan are varied concerning nuclear power, as in siting of the power plants and the n.s. (nuclear ship) Mutsu. The paper describes the following: safety, environmental protection, nuclear power generation, nuclear fuel cycle, fission reactor and fusion reactor development, nuclear-powered ships, and radiation utilization.

Development, operational experience and implications for future design of FBRS in Japan

International Nuclear Information System (INIS)

Sawai, S.; Hori, M.

1990-01-01

Joyo, the 100 MW t experimental reactor, has been successfully operated since 1977, and Monju, the 280 MW e prototype FBR, is under construction, with the first criticality planned for 1991. To promote FBR research and development efficiently - including the demonstration FBR (DFBR) programme - a steering committee for R and D was organized in 1986 by the Japan Atomic Power Company, the Power Reactor and Nuclear Fuel Development Corporation, the Japan Atomic Energy Research Institute and the Central Research Institute of Electric Power Industry. A design study of the DFBR is now underway to define its basic specifications by 1990. R and D for Monju, DFBR and future commercial FBRs has been done (1) to improve key technologies developed through the Joyo and Monju programmes; (2) to develop innovative technologies to make FBRs commercial; (3) to promote FBR development in conjunction with the development of the FBR fuel cycle. (author)

Demonstration tokamak power plant

International Nuclear Information System (INIS)

Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

1983-01-01

A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System

Conceptual design report on advanced marine reactor MRX of Japan

International Nuclear Information System (INIS)

Wang Shengguo

1995-01-01

Design studies on the advanced marine reactors have been done continuously since 1983 at Japan Atomic Energy Institute (JAERI) in order to develop attractive marine reactors for the next generation. At present, two concepts of marine reactor are being formulated. One is 100 MWt MRX (marine Reactor X) for the marine reactor and the other is 150 kWe DRX (Deep Sea-Reactor X) for a deep-sea research vessel. They are characterized by an integral type PWR, built-type control rod drive mechanisms, a water-filled container and a passive decay heat removal system, which realize highly passive safe and compact reactors. The paper is a report about all major results of the MRX design study

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

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

Present status of nuclear education and training in Japan

International Nuclear Information System (INIS)

Kiyose, R.; Sumita, K.; Moriya, F.

1994-01-01

In Japan, where about 30% of electricity is supplied by nuclear actives require a good number of able and ambitious young scientists and engineers especially in the future. On the other hand, almost all Japanese electric power companies, which operate nuclear power plants, are striving to keep expertise of reactor operators as high as possible. Present status in Japan of education at universities, research and training reactors, training courses at governmental institutions and nonprofit organizations, and operator training centers of electric power companies, are reviewed. 3 tabs

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

Trends on R and D of the innovative nuclear reactors in Japan

International Nuclear Information System (INIS)

Kinoshita, Izumi

2002-01-01

In Japan, since LWRs introduced from U.S.A. began their business operations one by one from 1970 and 1971, their scale-up were carried out, to reach, at present, a condition on developments of ABWR-2 of 1700 MW class in output and APWR+. They are on a line of large scale LWR development aiming at further upgrading of their economical efficiency, safety, operability and maintenance by improving and developing conventional reactors. On the other hand, an innovative small scale reactor capable of siting at proximity of its markets and flexibly responsible to needs, a low decelerated spectrum reactor intending to effectively use the resources, an super-critical pressure reactor aiming at upgrading of thermal efficiency, a high temperature gas reactor aiming at hydrogen production using nuclear heat , and so on, and so forth, are investigated at a number of institutes. And, on the fast breeder reactor, some innovative investigations such as small-scale reactor, reactor using coolant except metal sodium, and so on, in addition to development of sodium cooling large-scale reactor, under the 'Actual use strategy survey research' progressed at a center of the Japan Nuclear Cycle Development Institute, are promoted. Here were outlined on trends of R and D on various innovative reactors under classification of water cooling reactor, gas cooling reactor, and liquid metal cooling reactor. (G.K.)

Early hydrogen water chemistry in the boiling water reactor: industry-first demonstration

International Nuclear Information System (INIS)

Garcia, Susan E.; Odell, Andrew D.; Giannelli, Joseph F.

2012-09-01

Hydrogen injection into the BWR feedwater during power operation has resulted in significant IGSCC reductions. Further, noble metal application (NMCA) during shutdown or On-line NobleChem TM (OLNC) during power operation has greatly reduced the required hydrogen injection rate by catalyzing the hydrogen-oxygen reaction on the metal surfaces, reducing the electrochemical corrosion potential (ECP) at operating temperature to well below the mitigation ECP of -230 mV (SHE) at reactor water hydrogen to oxidant (O 2 + H 2 O 2 ) molar ratios of ≥2. Since IGSCC rates increase markedly at reduced temperature, and the potential for crack initiation exists, additional crack mitigation was desired. To close this gap in mitigation, the EPRI BWR Startup ECP Reduction research and development program commenced in 2008 to undertake laboratory and feasibility studies for adding a reductant to the reactor water system during start-ups. Under this program, ECP reductions of noble metal treated stainless steel sufficient to mitigate IGSCC at startup temperatures were achieved in the laboratory in the absence of radiation at hydrogen, hydrazine and carbohydrazide to oxygen molar ratios of ≥ 2, ≥1.5 and ≥0.7, respectively. Based on the familiarity of operating BWRs with using hydrogen, a demonstration of hydrogen injection during the startup of an actual BWR using noble metals was planned. This process, named EHWC (Early Hydrogen Water Chemistry), differs from the HDS (Hydrogen During Startup) approach that has been successful in Japan in that HDS injects sufficient hydrogen for bulk oxidant reduction whereas EHWC injects a smaller amount of hydrogen, sufficient to achieve a hydrogen:oxidant molar ratio of at least two at noble metal treated surfaces. The industry-first EHWC demonstration was performed at Exelon's Peach Bottom 3 nuclear power plant in October 2011. Prior to EHWC, Peach Bottom 3 had one NMCA (October 1999) and five annual OLNC applications (starting in 2007

Urgent Safety Measures in Japan after Great East Japan Earthquake

International Nuclear Information System (INIS)

Taniura, W.; Otani, H.

2012-01-01

Due to tsunami triggered by the Great East Japan Earthquake on March 11, 2011, the operating and refueling reactor facilities at Fukushima Dai-ichi and Dai-ni Nuclear Power Plants of Tokyo Electric Power Co. caused a nuclear hazard. Japanese electric power companies voluntarily began to compile various urgent measures against tsunami within the week the hazard was caused. As for the urgent safety measures of each licensee, it is clarified that effective measures have been appropriately implemented as a result of the inspection of the national government, the verification based on the guideline of the Japan Society of Maintenology and the stress test. (author)

Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

International Nuclear Information System (INIS)

Murata, Takashi; Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

2001-01-01

The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

Energy Technology Data Exchange (ETDEWEB)

Murata, Takashi [Kyoto Univ., Graduate School of Energy Science, Kyoto (Japan); Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

2001-12-01

The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

Change in plan for installation of nuclear reactor in No.1 atomic powered vessel of Japan Atomic Energy Research Institute (change in purpose of use and in method for nuclear reactor installation and spent fuel disposal) (report)

International Nuclear Information System (INIS)

1987-01-01

This report, compiled by the Nuclear Safety Commission to be submitted to the Prime Minister, deals with studies concerning some changes in the plan for the installation of a nuclear reactor in the No.1 atomic powered vessel to be constructed under the Japan Atomic Energy Research Institute (changes in the purpose of its use and in the methods for the nuclear reactor installation and spent fuel disposal). The conclusions of and procedures for the examination and evaluation are presented and then detailes of the studies are described. The study on the location requirements for the incidental land facilities at Sekinehama covers various conditions concerning the location, geology, earthquakes, meteorology, hydrology and social environment. The study on the safety design of the nuclear reactor facilities deals with the reactor, fuel handling facilities and other auxiliary facilities, as well as various land facilities to be constructed at Sekinehama including the reactor facilities and other facilities for fuel handling, waste disposal and protection and management of radioactive rays. Evaluation of possible radiation emission is shown and the accident analysis is also addressed. (Nogami, K.)

Thorium research activities in Japan

International Nuclear Information System (INIS)

Sasa, Toshinobu

2015-01-01

The nuclear energy policy in Japan is based on the Uranium-Plutonium fuel cycle with Light Water Reactors (LWR) and Fast Breeder Reactors (FBR). After the accident at Fukushima-Daiichi Nuclear Power Plant, the Japanese government recognizes the importance to ensure the flexibility for future nuclear power generation and then, it was specified in the latest Japanese strategic energy plan. Two research groups related to thorium fuelled nuclear systems and fuel cycle was set up in the Atomic Energy Society of Japan in 2013. One is a 'Research Committee on Nuclear Applications of Molten Salt'. The committee was established to discuss the current molten-salt technology including molten-salt cooled reactor, molten-salt fuelled reactor, accelerator driven system, fusion reactor blankets and dry reprocessing processes. Throughout two years discussion, the committee summarizes a current state of the art and issues of molten-salt application systems. Committee also discussed the handling technologies for molten-salt reactors especially in China and United Kingdom, issues of molten-salt application to fusion reactor, dry reprocessing of spent nuclear fuel, and non-nuclear application of molten-salt. Term of the committee will be extended for further research activities

Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

International Nuclear Information System (INIS)

Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

2011-01-01

Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

Public acceptance of nuclear power development in Japan

International Nuclear Information System (INIS)

Ohori, H.

1977-01-01

Although the Japanese set out to achieve the peaceful uses of atomic energy in 1956, the question of public acceptance took on serious proportions only as the development of nuclear power moved toward commercial application. A string of reactor troubles over the past few years complicated the question apparently to the point where it could scarcely be worse. It is not possible to deal with opposition movements in Japan without taking into account the background of the special national sentiment born of the people's experience of the atomic bombings, but it is also true that the people's deep-going fears of atomic energy have been increased by sensational newspaper reports, as well as internetional attacks by the opponents of nuclear development. Added to this, the ''Mutsu'' incident and other troubles have given the people distrust of the whole nuclear administration and those responsible for nuclear regulation. but, at the same time, the oil crisis of 1973 brought about an awakening of the people to the need for the development of nuclear power to solve Japan's energy problems, for Japan is seriously lacking in natural resources. An influential newspaper took samplings of public opinion in 1975 which revealed that, while 48 percent of the people expressed fears of atomic energy, 70 percent, including those who had some misgivings but still took the need for granted, said that Japan has no choice but to depend on nuclear power. The Government and industry have made long-range projections on nuclear power generation, forecasting that it will expand to 25 percent of all power generating plants by 1985, and to 35 percent by 1990. The gravest problem to be solved if this projected scale of nuclear development is to be achieved is the shortage of adequate plant sites. This can not be solved unless every effort is made to dispel the general feeling of mistrust mentioned, and to make sure that the development of nuclear power is socially accepted. It is hoped that the

Cermet-fueled reactors for multimegawatt space power applications

International Nuclear Information System (INIS)

Cowan, C.L.; Armijo, J.S.; Kruger, G.B.; Palmer, R.S.; Van Hoomisson, J.E.

1988-01-01

The cermet-fueled reactor has evolved as a potential power source for a broad range of multimegawatt space applications. In particular, the fast spectrum reactor concept can be used to deliver 10s of megawatts of electric power for continuous, long term, unattended operation, and 100s of megawatts of electric power for times exceeding several hundred seconds. The system can also be utilized with either a gas coolant in a Brayton power conversion cycle, or a liquid metal coolant in a Rankine power conversion cycle. Extensive testing of the cermet fuel element has demonstrated that the fuel is capable of operating at very high temperatures under repeated thermal cycling conditions, including transient conditions which approach the multimegawatt burst power requirements. The cermet fuel test performance is reviewed and an advanced cermet-fueled multimegawatt nuclear reactor is described in this paper

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2014

International Nuclear Information System (INIS)

2016-02-01

The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30 MW in December 2001 and achieved the 950degC of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2014, we started to apply the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 by the Pacific coast of Tohoku Earthquake. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2014. (author)

Status of reactor pressure vessel embrittlement study in Japan

International Nuclear Information System (INIS)

Sasajima, H.

1997-01-01

Since the construction of Japanese first commercial nuclear power plant in 1966, 52 nuclear power plants have been commissioned in Japan to commercial operation. Japanese first nuclear power plant has now been service for 30 years and the aging of nuclear power plants is steadily progressing in general. Under these circumstances, the Japan Power Engineering and Inspection Corporation (JAPEIC) is executing, under consignment by the Ministry of International Trade and Industry (MITI), the development and verification test programs for plant integrity evaluation technology by which nuclear power plant aging can be appropriately handled. This paper shows the outline of study dealing with embrittlement of RPV caused by neutron irradiation, as one of the activity of JAPEIC. The embrittlement of RPV caused by neutron irradiation is manifested as a shift of transition temperature and as a reduction in Upper Shelf Energy (USE). In JAPEIC, the study dealing with a shift of transition temperature was conducted in the ''Reactor Pressure Vessel Pressurized Thermal Shock Test Project (the PTS Project)'', and the study dealing with a reduction in USE has been conducted in the ''Nuclear Power Plant Life Management Technology (the PLIM Project)''. And the reconstitution technology of surveillance test specimen has been conducted in PLIM Project as one of the measures to improve monitoring above material characteristic changes. The integrity evaluation under the Pressurized Thermal Shock (PTS) events including the effect of neutron irradiation embrittlement was initiated in 1983 FY as the PTS Project and was completed in the 1991 FY. The study verified that plant integrity could be assured at not only the end of design life, but also an extended service life even when the severest PTS events were postulated. The PLIM Project, designed to develop and verify the integrity evaluation technology dealing with reduction of USE by neutron irradiation, was started in the 1996 FY as a 10

Situation of nuclear industry in Japan

International Nuclear Information System (INIS)

2004-03-01

This document presents the situation of nuclear industry in Japan: cooperation with France in the domain of the fuel cycle (in particular the back-end) and of for the industrial R and D about fast reactors and nuclear safety; present day situation characterized by a series of incidents in the domain of nuclear safety and by an administrative reorganization of the research and safety organizations; power of local representatives, results of April 2003 elections, liberalization of the electric power sector, impact of the TEPCO affair (falsification of safety reports) on the nuclear credibility, re-start up of the Monju reactor delayed by judicial procedures, stopping of the program of MOX fuel loading in Tepco's reactors, discovery of weld defects in the newly built Rokkasho-mura reprocessing plant, an ambitious program of reactors construction, the opportunity of Russian weapons dismantling for the re-launching of sodium-cooled fast reactors; the competition between France and Japan for the setting up of ITER reactor and its impact of the French/Japanese partnership. (J.S.)

Configuration and layout of the tandem mirror Fusion Power Demonstrator

International Nuclear Information System (INIS)

Clarkson, I.R.; Neef, W.S.

1983-01-01

Studies have been performed during the past year to determine the configuration of a tandem mirror Fusion Power Demonstrator (FPD) machine capable of producing 1750 MW of fusion power. The FPD is seen as the next logical step after the Mirror Fusion Test Facility-B (MFTF-B) toward operation of a power reactor. The design of the FPD machine allows a phased construction: Phase I, a hydrogen or deuterium checkout machine; Phase 2, a DT breakeven machine; Phase 3, development of the Phase 2 machine to provide net power and act as a reactor demonstrator. These phases are essential to the development of remote handling equipment and the design of components that will ultimately be remotely handled. Phasing also permits more modes funding early in the program with some costs committed only after reaching major milestones

Reactor power control device

International Nuclear Information System (INIS)

Doi, Kazuyori.

1981-01-01

Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

FY 2000 report on the demonstrative research for photovoltaic power generation system in Thailand. Demonstrative study on photovoltaic power generation grid-connected system

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-09-01

In relation to the demonstrative study of the photovoltaic power system that is planned in Libong island, Thailand, the FY 2000 results were reported. In this R and D, construction/demonstrative operation were planned for a photovoltaic power station with a generation output of 100kW, photovoltaic power system in school facilities, and system for transmitting/distributing power to houses by connecting the power station and power system. In this fiscal year, the field survey was conducted together with the alteration from Yao Yai island, for which the demonstrative study was planned at first, to Libong island. The electric equipment was selected which met the requests from Thailand and the results of the field survey, and the basic design of the photovoltaic power generation/transmission/distribution system was completed. Based on this, the design/manufacture of photovoltaic power generation modules, power control equipment, measuring equipment, etc. were made. At the construction site of photovoltaic power station, construction work such as land formation was conducted. Further, Thai engineers who visited Japan did the following: discussions about power system, presence at test/inspection of photovoltaic power generation modules, visits to photovoltaic power stations, wind power stations, etc. (NEDO)

Fusion neutronics plan in the development of fusion reactor. With the aim of realizing electric power

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Hiroo; Morimoto, Yuichi; Ochiai, Kentarou; Sugimoto, Masayoshi; Nishitani, Takeo; Takeuchi, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-10-01

On June 1992, Atomic Energy Commission in Japan has settled Third Phase Program of Fusion Research and Development to achieve self-ignition condition, to realize long pulse burning plasma and to establish basis of fusion engineering for demonstration reactor. This report describes research plan of Fusion Neutron Laboratory in JAERI toward a development of fusion reactor with an aim of realizing electric power. The fusion neutron laboratory has a fusion neutronics facility (FNS), intense fusion neutron source. The plan includes research items in the FNS; characteristics of shielding and breeding materials, nuclear characteristics of materials, fundamental irradiation process of insulator, diagnostics materials and structural materials, and development of in-vessel diagnostic technology. Upgrade of the FNS is also described. Also, the International Fusion Material Irradiation Facility (IFMIF) for intense neutron source to develop fusion materials is described. (author)

Nuclear power development in Japan

International Nuclear Information System (INIS)

Mishiro, M.

2000-01-01

This article describes the advantages of nuclear energy for Japan. In 1997 the composition of the total primary energy supply (TPES) was oil 52.7%, coal 16.5%, nuclear 16.1% and natural gas 10.7%. Nuclear power has a significant role to play in contributing to 3 national interests: i) energy security, ii) economic growth and iii) environmental protection. Energy security is assured because a stable supply of uranium fuel can be reasonably expected in spite of dependence on import from abroad. Economic growth implies the reduction of energy costs. As nuclear power is capital intensive, the power generation cost is less affected by the fuel cost, therefore nuclear power can realize low cost by favoring high capacity utilization factor. Fossil fuels have substantial impacts on environment such as global warming and acid rain by releasing massive quantities of CO 2 , so nuclear power is a major option for meeting the Kyoto limitations. In Japan, in 2010 nuclear power is expected to reach 17% of TPES and 45% of electricity generated. (A.C.)

US nuclear policy and business trend of Japan's nuclear industries

International Nuclear Information System (INIS)

Matsuo, Yuji

2010-01-01

As several countries in the east-Asia and middle-east area have been taking an increasing interest in the deployment of nuclear power generation, Japan's nuclear industries have promoted international business activities including the success in the bid of second nuclear power plants in Vietnam. While there are plans for more than thirty of new reactors in the US, the lifetime extension of existing aged reactors, development of non-existing natural gas and trend of greenhouse gases reduction measures have dampened these plans and probably most of new units will not start construction by 2030. This article reviewed the details of US's new nuclear power introduction, trend of recent government's policies, future perspective of nuclear power construction and business trend of Japan's nuclear industries. Japan's industries should be flexible regarding nuclear power as one option to realize low-carbon society. (T. Tanaka)

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

Japan and atomic co-operation

International Nuclear Information System (INIS)

1965-01-01

Japan, which is host country for the Ninth Regular Session of the Agency General Conference, has an important programme of nuclear power development to meet future needs. In addition, Japan is active in other applications of atomic energy and is building up a domestic nuclear engineering industry. Japan has profited by the Agency as a channel of international cooperation, and was a party to the first bilateral agreement in which the responsibility for administering safeguards against the diversion of materials to military purposes, was transferred to the Agency. Japan has also lent support to Agency programmes by gifts, training courses, research, and the loan of experts. In 1961, the Japan Atomic Energy Commission (AEC) formulated the 'Long-Range Programme for Development and Utilization of Atomic Energy', on the basis of the economic prospects of nuclear power generation, and the conditions necessary to meet the ever-increasing domestic energy demands. According to this programme, in the light of power reactor development trends overseas, it is expected that nuclear power costs will compete with those of oil burning stations by 1970. On this basis, total nuclear power generating capacity of 1000 MW(e) will be attained by 1970, and 7000 - 9 500M(e) by 1980. As a prelude to the above programme the Japan Atomic Power Company (JAPCO) began construction in 1959 of a graphite-moderated gas-cooled nuclear power station (Improved Calder Hall type) of 165 MW(e) gross capacity. This is now progressing smoothly, and reached criticality in May 1965; it is expected to supply commercial power by the end of this year. The second nuclear power station will be built by the same company on the coast of the Japan Sea, with a light water-moderated reactor of 250 - 300 MW(e) capacity. The construction plan i s currently being pushed forward for completion in 1970. Thereafter three private utility companies - Tokyo, Kansai and Chubu Electric Companies - are doing preparatory work for

Conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR), September 1978

International Nuclear Information System (INIS)

Kelley, J.L.

1978-12-01

The flexibility of the fusion hybrid reactor to function as a fuel production facility, power plant, waste disposal burner or combinations of all of these, as well as the reactor's ability to use proliferation resistant fuel cycles, has provided the incentive to assess the feasibility of a near-term demonstration plant. The goals for a Demonstration Tokamak Hybrid Reactor (DTHR) were established and an initial conceptual design was selected. Reactor performance and economics were evaluated and key developmental issues were assessed. The study has shown that a DTHR is feasible in the late 1980's, a significant quantity of fissile fuel could be produced from fertile thorium using present day fission reactor blanket technology, and a large number of commercially prototypical components and systems could be developed and operationally verified. The DTHR concept would not only serve as proof-of-principle for hybrid technology, but could be operated in the ignited mode and provide major advancements for pure fusion technology

Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY2003

International Nuclear Information System (INIS)

2005-03-01

The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research Establishment of The Japan Atomic Energy Research Institute (JAERI) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power. Coolant of helium-gas circulates under the pressure of about 4Mpa, and the reactor inlet and outlet temperature are 395degC and 950degC (maximum), respectively coated particle fuel is used as fuel, and the HTTR core is composed of graphite prismatic blocks. The full power operation of 30MW was attained in December, 2001, and then JAERI received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2003 before the high temperature test operation of 950degC. (author)

State of opening the cover and carrying out the checkup of the reactor vessel of the nuclear-powered ship 'Mutsu' by Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

1989-01-01

In the checkup by opening the cover of the reactor vessel of the nuclear-powered ship 'Mutsu', Japan Atomic Energy Research Institute carried out the checkup and maintenance for the reactor proper, control system and primary coolant facilities including the secondary side of steam generators and the pressure balancing valve of the containment vessel. The works were classified into the opening of the reactor, checkup, maintenance and restoration. The opening was begun on August 4, 1988, and finished on December 5. The checkup and maintenance were begun on September 22, and are still continued now. The maximum radiation dose rate on the surfaces of fuel assemblies and control rods and at the positions 1 m distant from them was measured. The results of the checkup of various components are reported. In 290 absorbent rods of control rods, spot corrosion and discoloration were observed, of which the spot corrosion penetrated the walls of 4 rods. Also in 12 fuel rods, spot corrosion was observed near the welded end plugs, but leak was not observed. (K.I.)

The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1979-01-01

The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

Radioactive waste management and disposal scenario for fusion power reactors

Energy Technology Data Exchange (ETDEWEB)

Tabara, Takashi; Yamano, Naoki [Sumitomo Atomic Energy Industries Ltd., Tokyo (Japan); Seki, Yasushi; Aoki, Isao

1997-10-01

The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a light water reactor (LWR) have been evaluated and compared. At first, the amount and the radioactive level of the radwaste generated in five fusion reactors ware evaluated by an activation calculation code. Next, a possible radwaste disposal scenario applicable to fusion radwaste in Japan is considered and the disposal cost evaluated under certain assumptions. The exposure doses are evaluated for the skyshine of gamma-rays during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical LWR was estimated based on a literature survey and the disposal cost was evaluated using the same assumptions as for the fusion reactors. It is found that the relative cost of disposal is strongly dependent on the cost for interim storage of medium level waste of fusion reactors and the cost of high level waste for the LWR. (author)

Calibration of RB reactor power

International Nuclear Information System (INIS)

Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

1976-09-01

The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

Water electrolysis plants for hydrogen and oxygen production. Shipped to Tsuruga Power Station Unit No.1, and Tokai No.2 power station, the Japan Atomic Power Co

International Nuclear Information System (INIS)

Ueno, Syuichi; Sato, Takao; Ishikawa, Nobuhide

1997-01-01

Ebara's water electrolysis plants have been shipped to Tsuruga Power Station Unit No.1, (H 2 generation rate: 11 Nm 3 /h), and Tokai No.2 Power Station (H 2 generation rate: 36 Nm 3 /h), Japan Atomic Power Co. An outcome of a business agreement between Nissho Iwai Corporation and Norsk Hydro Electrolysers (Norway), this was the first time that such water electrolysis plants were equipped in Japanese boiling water reactor power stations. Each plant included an electrolyser (for generating hydrogen and oxygen), an electric power supply, a gas compression system, a dehumidifier system, an instrumentation and control system, and an auxiliary system. The plant has been operating almost continuously, with excellent feedback, since March 1997. (author)

Review of the general atomic experimental fusion power reactor initial conceptual design

International Nuclear Information System (INIS)

Baker, C.C.; Sager, P.H. Jr.; Harder, C.R.

1976-01-01

The primary objective of the Experimental Power Reactor (EPR) is to provide the necessary interface between physics experiments and the first demonstration power plants. Since economically viable tokamak-type reactors may well have to be very high Q devices (ratio of fusion power out to power into the plasma), it will be essential for a tokamak demonstration reactor to operate at or near ignition conditions. Thus, it is believed that one of the primary objectives of the EPR must be to fully model the behavior of a D-T burning plasma required in the reactor of a demonstration plant. Therefore, a major objective of the EPR should be to achieve ignition conditions. In addition to demonstrating the ability to ignite and control a D-T plasma, it is also desirable that the EPR should produce, or at least demonstrate the ability to produce, a small amount of net electrical power. These objectives should be accomplished at a reasonable cost; this implies achieving a sufficiently high β (ratio of plasma pressure to magnetic field pressure). It is believed that noncircular cross section tokamaks offer the best chance of realizing these objectives. Consequently, noncircular cross sections are a major design feature of the General Atomic EPR

Study of lower hybrid current drive for the demonstration reactor

Energy Technology Data Exchange (ETDEWEB)

Molavi-Choobini, Ali Asghar [Dept. of Physics, Faculty of Engineering, Islamic Azad University, Shahr-e-kord Branch, Shahr-e-kord (Iran, Islamic Republic of); Naghidokht, Ahmed [Dept. of Physics, Urmia University, Urmia (Iran, Islamic Republic of); Karami, Zahra [Dept. of Engineering, Islamic Azad University, Zanjan Branch, Zanjan (Iran, Islamic Republic of)

2016-06-15

Steady-state operation of a fusion power plant requires external current drive to minimize the power requirements, and a high fraction of bootstrap current is required. One of the external sources for current drive is lower hybrid current drive, which has been widely applied in many tokamaks. Here, using lower hybrid simulation code, we calculate electron distribution function, electron currents and phase velocity changes for two options of demonstration reactor at the launched lower hybrid wave frequency 5 GHz. Two plasma scenarios pertaining to two different demonstration reactor options, known as pulsed (Option 1) and steady-state (Option 2) models, have been analyzed. We perceive that electron currents have major peaks near the edge of plasma for both options but with higher efficiency for Option 1, although we have access to wider, more peripheral regions for Option 2. Regarding the electron distribution function, major perturbations are at positive velocities for both options for flux surface 16 and at negative velocities for both options for flux surface 64.

Role of fast breeders in Japan

International Nuclear Information System (INIS)

Oyama, A.; Tomabechi, K.

1978-09-01

To meet increasing future energy demand in Japan utilization of fission energy should be promoted. In particular it is of vital importance to develop and utilize FBRs as soon as possible in order to save the natural uranium needed. If one considers the commercial introduction of FBRs in the mid-1990s in Japan, a delay of only one year will eventually result in an additional demand for natural uranium of more than 20,000 tons, because several LWRs will have to be installed instead. Ten years have passed since the development of FBRs in Japan was initiated as a national project with the highest priority and now the experimental fast reactor JOYO is successfully being operated at 50MW and the prototype fast breeder reactor MONJU has reached the stage of proceeding to construction with a schedule of operation in the mid-1980s. Following operation of MONJU, construction of a large demonstration reactor of 1000 - 15000 MW(e) will be undertaken. Some 2 - 3 years after the construction of the demonstration reactor, a series of reactors will be constructed similar in size and design to promote commercialization of LMFBRs. Strong efforts will be made to put this programme into practice. It is expected that LMFBRs will play an important role in mitigating the serious problem of energy supply in Japan foreseeable around the turn of the century

Conceptual design of a demonstration reactor for electric power generation

International Nuclear Information System (INIS)

Asaoka, Y.; Hiwatari, R.; Okano, K.; Ogawa, Y.; Ise, H.; Nomoto, Y.; Kuroda, T.; Mori, S.; Shinya, K.

2005-01-01

Conceptual study on a demonstration plant for electric power generation, named Demo-CREST, was conducted based on the consideration that a demo-plant should have capacities both (1) to demonstrate electric power generation in a plant scale with moderate plasma performance, which will be achieved in the early stage of the ITER operation, and foreseeable technologies and materials and (2) to have a possibility to show an economical competitiveness with advanced plasma performance and high performance blanket systems. The plasma core was optimized to be a minimum size for both net electric power generation with the ITER basic plasma parameters and commercial-scale generation with advance plasma parameters, which would be attained by the end of ITER operation. The engineering concept, especially the breeding blanket structure and its maintenance scheme, is also optimized to demonstrate the tritium self-sustainability and maintainability of in-vessel components. Within the plasma performance as planned in the present ITER program, the net electric power from 0 MW to 500 MW is possible with the basic blanket system under the engineering conditions of maximum magnetic field 16 T, NBI system efficiency 50%, and NBI current drive power restricted to 200 MW. Capacities of stabilization of reversed shear plasma and the high thermal efficiency are additional factors for optimization of the advanced blanket. By replacing the blanket system with the advanced one of higher thermal efficiency, the net electric power of about 1000 MW is also possible so that the economic performance toward the commercial plant can be also examined with Demo-CREST. (author)

IAEA Expert Team Completes Mission to Review Japan's Nuclear Power Plant Safety Assessment Process, 31 January 2012, Tokyo, Japan

International Nuclear Information System (INIS)

2012-01-01

Full text: A team of international nuclear safety experts today completed a review of Japan's two-stage process for assessing nuclear safety at the nation's nuclear power plants. The team began its work on 23 January and delivered a Preliminary Summary Report to Japanese officials today and plans to finish the final report by the end of February. National safety assessments and their peer review by the IAEA are a key component of the IAEA's Action Plan on Nuclear Safety, which was approved by the Agency's 152 Member States following last year's nuclear accident at Fukushima Daiichi Nuclear Power Station. At the request of the Government of Japan, the International Atomic Energy Agency (IAEA) organized a 10-person team to review the Japanese Nuclear and Industrial Safety Agency's (NISA) approach to the Comprehensive Assessments for the Safety of Existing Power Reactor Facilities and how NISA examines the results submitted by nuclear operators. The IAEA safety review mission consisted of five IAEA and three international nuclear safety experts. To help its review, the team held meetings in Tokyo with officials from NISA, the Japanese Nuclear Energy Safety (JNES) Organization, and the Kansai Electric Power Company (KEPCO), and the team visited the Ohi Nuclear Power Station to see an example of how Japan's Comprehensive Safety Assessment is being implemented by nuclear operators. 'We concluded that NISA's instructions to power plants and its review process for the Comprehensive Safety Assessments are generally consistent with IAEA Safety Standards', said team leader James Lyons, director of the IAEA's Nuclear Installation Safety Division. In its Preliminary Summary Report delivered today, the team highlighted a number of good practices and identified some improvements that would enhance the overall effectiveness of the Comprehensive Safety Assessment process. Good practices identified by the mission team include: Based on NISA instructions and commitments of the

Inspection, evaluation and maintenance guidelines for reactor vessel internals in JAPAN

International Nuclear Information System (INIS)

Sakashita, Akihiro; Goto, Tomoya; Hirano, Shinro; Dozaki, Koji

2010-01-01

Inspection and Evaluation Guidelines for reactor internals has been taken into the Rules on Fitness-for–Service for Nuclear Power Plants of The Japan Society of Mechanical Engineers. It is a base of the maintenance plan of each Nuclear Power Plant. A plant maintenance methodology will have more importance to maintain the plant safety and stable plant operation. This paper introduces the systematization of the maintenance such as repair, replacement, preventive maintenance in these guidelines. Maintenance methodologies are classified follows. Repair: methodology to reinforce degraded parts by some methods or prevent progress of degradation of without replacement of the existing structure when the degradation of structure is actualized. Replacement: methodology to replace the existing structure with new one when the degradation of structure is actualized. Preventive maintenance : methodology to mitigate the damaged condition. When the maintenance methodologies are implemented in the actual plant, we have to consider the feedback of the inspection program and plant life management. (author)

Spent fuel management in Japan

International Nuclear Information System (INIS)

Shirahashi, K.; Maeda, M.; Nakai, T.

1996-01-01

Japan has scarce energy resources and depends on foreign resources for 84% of its energy needs. Therefore, Japan has made efforts to utilize nuclear power as a key energy source since mid-1950's. Today, the nuclear energy produced from 49 nuclear power plants is responsible for about 31% of Japan's total electricity supply. The cumulative amount of spent fuel generated as of March 1995 was about 11,600 Mg U. Japan's policy of spent fuel management is to reprocess spent nuclear fuel and recycle recovered plutonium and uranium as nuclear fuel. The Tokai reprocessing plant continues stable operation keeping the annual treatment capacity or around 90 Mg U. A commercial reprocessing plant is under construction at Rokkasho, northern part of Japan. Although FBR is the principal reactor to use plutonium, LWR will be a major power source for some time and recycling of the fuel in LWRs will be prompted. (author). 3 figs

Reactor power monitoring device

International Nuclear Information System (INIS)

Dogen, Ayumi; Ozawa, Michihiro.

1983-01-01

Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

Status of LMFBR development project in Japan

International Nuclear Information System (INIS)

Nagane, G.; Akebi, M.; Matsuno, Y.

1987-01-01

Initiation of the LMFBR development project in Japan was decided by the Atomic Energy Commission of Japan in 1966. In 1967, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established to realize the project as a part of its tasks of a wide scope covering all the reseatch and development activities concerning fuel cycle. In the present paper the status of experimental fast reactor (Joyo), which is the first milestone of the LMFBR project, prototype fast reactor (Monju) and R and D activities supporting the project including that for larger LMFBRs in the future is described. (author)

New reactor technology: safety improvements in nuclear power systems.

Science.gov (United States)

Corradini, M L

2007-11-01

Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

Engineering experiences through nuclear power development in Japan

International Nuclear Information System (INIS)

Uchida, Hideo

2004-01-01

This keynote paper deals with: energy issues and nuclear power development in Japan, problems of radiation protection, licensing and safety regulations, research on LOCA and ECCS, stress corrosion cracks related to pressure vessels, nuclear fuel failures, steam generators, incidents, waste management and fuel cycle facilities. In conclusion it is stated that: on order to cope with global matters vitally affecting the electricity generation, taking into consideration Japanese specific energy issues, the nuclear power development has been an indispensable policy of Japan. In order to proceed with further development of nuclear power plants, it is necessary to obtain proper understanding by the public, showing assurance of the safety and reliable operation of nuclear power plants through daily plant operation. The nuclear safety issues should be considered from a global point of view. It is necessary to establish common safety standards which could harmonize the safety level of nuclear power plants in the world. The safety goal concerning severe accidents should be established as an internationally agreeable one. Japan has accumulated highly technological experience in maintenance of nuclear power plants. It is believed that the cumulative experiences in Japan can contribute to the further improvement of safety of nuclear power plants throughout the world, and for this aim a mutual information exchange should be encouraged

Present state and future prospect of development of high temperature gas-cooled reactors in Japan

International Nuclear Information System (INIS)

Sanokawa, Konomo

1994-01-01

High temperature gas-cooled reactors can supply the heat of about 1000degC, and the high efficiency and the high rate of heat utilization can be attained. Also they have the features of excellent inherent safety, the easiness of operation, the high burnup of fuel and so on. The heat utilization of atomic energy in addition to electric power generation is very important in view of the protection of global environment and the diversification of energy supply. Japan Atomic Energy Research Institute has advanced the construction of the high temperature engineering test and research reactor (HTTR) of 30 MW thermal output, aiming at attaining the criticality in 1998. The progress of the development of a high temperature gas-cooled reactor is described. For 18 years, the design study of the reactor was advanced together with the research and development of the reactor physics, fuel and materials, high temperature machinery and equipment and others, and the decision of the design standard and the development of computation codes. The main specification and the construction schedule are shown. The reactor building was almost completed, and the reactor containment vessel was installed. The plan of the research and development by using the HTTR is investigated. (K.I.)

Revision of construction plan for advanced thermal demonstration reactor

International Nuclear Information System (INIS)

1996-01-01

The Federation of Electric Power Companies demanded the revision of the construction plan for the advanced thermal demonstration reactor, which is included in the 'Long term plan on the research, development and utilization of atomic energy' decided by the Atomic Energy Commission in 1994, for economical reason. The Atomic Energy Commission carried out the deliberation on this demand. It was found that the cost of construction increases to 580 billion yen, and the cost of electric power generation increases three times as high as that of LWRs. The role as the reactor that utilizes MOX fuel can be substituted by LWRs. The relation of trust with the local town must be considered. In view of these circumstances, it is judged that the stoppage of the construction plan is appropriate. It is necessary to investigate the substitute plan for the stoppage, and the viewpoints of investigating the substitute plan, the examination of the advanced BWR with all MOX fuel core and the method of advancing its construction are considered. On the research and development related to advanced thermal reactors, the research and development contributing to the advance of nuclear fuel recycling are advanced, and the prototype reactor 'Fugen' is utilized. (K.I.)

Time-optimal control of reactor power

International Nuclear Information System (INIS)

Bernard, J.A.

1987-01-01

Control laws that permit adjustments in reactor power to be made in minimum time and without overshoot have been formulated and demonstrated. These control laws which are derived from the standard and alternate dynamic period equations, are closed-form expressions of general applicability. These laws were deduced by noting that if a system is subject to one or more operating constraints, then the time-optimal response is to move the system along these constraints. Given that nuclear reactors are subject to limitations on the allowed reactor period, a time-optimal control law would step the period from infinity to the minimum allowed value, hold the period at that value for the duration of the transient, and then step the period back to infinity. The change in reactor would therefore be accomplished in minimum time. The resulting control laws are superior to other forms of time-optimal control because they are general-purpose, closed-form expressions that are both mathematically tractable and readily implanted. Moreover, these laws include provisions for the use of feedback. The results of simulation studies and actual experiments on the 5 MWt MIT Research Reactor in which these time-optimal control laws were used successfully to adjust the reactor power are presented

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

International Nuclear Information System (INIS)

2014-12-01

The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30MW in December 2001 and achieved the 950degC of outlet coolant temperature at the outside the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2013, we started to prepare the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 when the Pacific coast of Tohoku Earthquake (2011.3.11) occurred. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2013. (author)

Outline of examination guides of water-cooled research reactors in Japan

International Nuclear Information System (INIS)

Yoshino, F.; Kimura, R.

1992-01-01

The Nuclear Safety Commission of Japan published two examination guides of water-cooled research reactors on July 18, 1991; one is for safety design, and another is for safety evaluation. In these guides, careful consideration is taken into account on the basic safety characteristic features of research reactors in order to be reasonable regulative requirements. This paper describes the fundamental philosophy and outline of the guides. (author)

Reactor power control method upon accidents of electrical power system

International Nuclear Information System (INIS)

Hirose, Masao.

1983-01-01

Purpose: To enable to continue the operation of a BWR type reactor by avoiding the scram while suppressing the reactor power, just after the external disturbance such as earth-trouble in power-transmission network. Method: Steep power drop of an electrical generator is to be detected not only by a current-type power-load-unbalance relay but also with a power-type power-load-unbalance-relay. If steep power-drop was detected by the latter relay, a previously selected control rod is rapidly inserted into the reactor. In this way, in the case where there is a possibility of the reactor scram, the scram can be avoided by suppressing the reactor power, thus the reactor operation can be continued. (Kamimura, M.)

Power reactors in member states

International Nuclear Information System (INIS)

1975-01-01

This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-09-01

A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

Hybrid reactors: recent progress of a demonstration pilot

International Nuclear Information System (INIS)

Billebaud, Annick

2006-12-01

Accelerator driven sub-critical reactors are subject of many research programmes since more than ten years, with the aim of testing the feasibility of the concept as well as their efficiency as a transmutation tool. Several key points like the accelerator, the spallation target, or neutronics in a subcritical medium were investigated extensively these last years, allowing for technological choices and the design of a low power European demonstration ADS (a few tens of MWth). Programmes dedicated to subcritical reactor piloting proposed a monitoring procedure to be validated in forthcoming experiments. Accelerator R and D provided the design of a LINAC for an ADS and research work on accelerator reliability is going on. A spallation target was operated at PSI and the design of a windowless target is in progress. All this research work converges to the design of a European demonstration ADS, the ETD/XT-ADS, which could be the Belgian MYRRHA project. (author)

Reactor power reduction system and method

International Nuclear Information System (INIS)

Bruno, S.J.; Dunn, S.A.; Raber, M.

1978-01-01

A method of operating a nuclear power reactor is disclosed which enables an accelerated power reduction of the reactor without completely shutting the reactor down. The method includes monitoring the incidents which, upon their occurrence, would require an accelerated power reduction in order to maintain the reactor in a safe operation mode; calculating the power reduction required on the occurrence of such an incident; determining a control rod insertion sequence for the normal operation of the reactor, said sequence being chosen to optimize reactor power capability; selecting the number of control rods necessary to respond to the accelerated power reduction demand, said selection being made according to a priority determined by said control rod insertion sequence; and inserting said selected control rods into the reactor core. 11 claims, 13 figures

Analysis of accidents and troubles of nuclear power plants in Japan

International Nuclear Information System (INIS)

Kobayashi, Kunio

1980-01-01

In Japan, electric power companies are obliged to report the accidents and troubles occurred in nuclear power stations to the MITI according to the relevant laws, and 166 cases in total have been reported as of the end of March, 1980. These accidents and troubles are all trivial, and do not cause problems from the viewpoint of the safety nuclear power stations. Regarding respective accidents and troubles, the causes have been sought thoroughly, and the sufficient countermeasures have been taken on all occasions. But in order to improve the reliability of nuclear power stations further, it is important to treat the accidents and troubles occurred so far statistically and grasp the general trend. Thereupon, 152 accidents and troubles occurred till September, 1979, were analyzed quantitatively, and the results are reported in this paper. From the results, the prospect hereafter is discussed. The number of the reported cases of accidents and troubles in each nuclear power plant in operation every year is tabulated. The accidents and troubles were relatively frequent in the initial two or three years of operation of respective new reactor types, but decreased thereafter. The systems to which troubled equipments belong and the troubled equipments are shown. Most troubles have occurred in reactor cooling systems and valves. The situations and causes of troubles, the operational conditions at the time of the accidents and troubles and the effects and others are reported. (Kako, I.)

SCWR Concepts in Japan

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-08-15

Two SCWR concepts are being developed in Japan, one corresponding to the thermal spectrum reactor and the other to the fast spectrum reactor. Yamada et al. described the thermal-spectrum reactor concept referred to as the Japan SCWR (or JSCWR). This concept was developed under the financial support of the Ministry of Economy, Trade and Industry (METI). The basic philosophy of the JSCWR development is to utilize proven light water reactor and supercritical fossil-fired power plant technologies as much as possible to minimize the R&D cost, time and risks. Therefore, the JSCWR is designed as a thermal neutron spectrum reactor using light water as moderator and reactor coolant. The JSCWR plant consists of a pressure-vessel type, once-through reactor and a direct Rankine cycle system. Reactor coolant fed through inlet nozzles is heated up in the core and flows through outlet nozzles with no recirculation in the vessel. Other options to the JSCWR core design are being investigated at the University of Tokyo. The electric output of the JSCWR is assumed to range from 600 MWe to 1700 MWe class to fulfill user’s requirements as much as possible. In this section, the reference value is selected to 1725 MWe, which corresponds to a reactor thermal output of 4039 MWth. Nakatsuka et al. described the core design for the fast-spectrum reactor, which is based on a similar plant system compared to that of the thermal-spectrum reactor. The fast-spectrum reactor, however, would produce higher power rating than the thermal-spectrum one of the same reactor pressure-vessel size. Since the fast-spectrum reactor does not require the moderator, its unit capital cost would be lower than the thermal-spectrum reactor.

Demonstrated operational and inherent safety of the prototype fast reactor (PFR)

International Nuclear Information System (INIS)

Smedley, J.A.; Gregory, C.V.; Judd, A.M.

1983-01-01

The Prototype Fast Reactor (PFR) is sited at Dounreay, on the north coast of Scotland in the United Kingdom, and has been in operation since 1974. Three aspects of the safety of the reactor are described, including the all-important practical consideration of operational safety, a demonstration of the limited consequences of a sodium/water reaction in a steam generator and the ability of the reactor to protect itself against highly improbable incidents. Attention is drawn to the low radiation levels in the plant and the correspondingly low dose rate to personnel. A feature of PFR operation has been the stable and predictable behaviour of its core together with the high degree of reliability exhibited by the engineered safety system. No failures have occurred within the standard driver charge but two experimental fuel pins suffered cladding failure, which was detected easily by the fission gas and delayed neutron detection systems. In the steam generating units sodium and water are separated by the single steel wall of the steam tubes. Although no under-sodium leak has occurred, an experimental programme is continuing and demonstrates that were any such leak to occur its consequences would be containable and would not result in the release of sodium to the environment or any breach of the reactor containment. The final section describes the inherent safety features of the reactor which enable it to survive a range of very improbable incidents even when the engineered safeguards fail. The features considered are natural circulation, which has been demonstrated by reactor experiment; the reactor's negative power coefficient, which, for example, enables the reactor to survive a complete loss of heat sink; and the durability of the fuel pins, demonstrated by a series of boiling experiments in the Dounreay Fast Reactor (DFR). (author)

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

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

International Nuclear Information System (INIS)

Schreiner, L.A.

1991-01-01

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

Reactor power distribution monitor

International Nuclear Information System (INIS)

Hoizumi, Atsushi.

1986-01-01

Purpose: To grasp the margin for the limit value of the power distribution peaking factor inside the reactor under operation by using the reactor power distribution monitor. Constitution: The monitor is composed of the 'constant' file, (to store in-reactor power distributions obtained from analysis), TIP and thermocouple, lateral output distribution calibrating apparatus, axial output distribution synthesizer and peaking factor synthesizer. The lateral output distribution calibrating apparatus is used to make calibration by comparing the power distribution obtained from the thermocouples to the power distribution obtained from the TIP, and then to provide the power distribution lateral peaking factors. The axial output distribution synthesizer provides the power distribution axial peaking factors in accordance with the signals from the out-pile neutron flux detector. These axial and lateral power peaking factors are synthesized with high precision in the three-dimensional format and can be monitored at any time. (Kamimura, M.)

Fuel combustion in thermal power plants in Japan

Energy Technology Data Exchange (ETDEWEB)

Kotler, V.R.

1983-11-01

The position of black coal in the energy balance of Japan is discussed. About 75% of electric energy is produced by thermal power plants. Eighty-five per cent of electricity is produced by power plants fired with liquid fuels and 3% by coal fired plants. Coal production in Japan, the forecast coal import to the country by 1990 (132 Mt/year), proportion of coal imported from various countries, chemical and physical properties of coal from Australia, China and Japan are discussed. Coal classification used in Japan is evaluated. The following topics associated with coal combustion in fossil-fuel power plants in Japan are discussed: coal grindability, types of pulverizing systems, slagging properties of boiler fuel in Japan, systems for slag removal, main types of steam boilers and coal fired furnaces, burner arrangement and design, air pollution control from fly ash, sulfur oxides and nitrogen oxides, utilization of fly ash for cement production, methods for removal of nitrogen oxides from flue gas using ammonia and catalysts or ammonia without catalysts, efficiency of nitrogen oxide control, abatement of nitrogen oxide emission from boilers by flue gas recirculation and reducing combustion temperatures. The results of research into air pollution control carried out by the Nagasaki Technical Institute are reviewed.

Possibilities of TWR and long life reactor

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Shimazu, Yoichiro; Handa, Norihiko

2010-01-01

Bill Gates identified the need to switch to zero-emission energy and clarified investing in Terra Power developing the TWR (Traveling Wave Reactor) in February 2010. He also visited Toshiba developing small reactor 4S (Super Safe Small and Simple). In Japan design studies of the TWR have been conducted on the CANDLE reactor without refueling and the 4S long life reactor with maintenance free. In this feature article, the state of R and D on the TWR in Japan and IAEA's activities on small reactors without online refueling were reviewed in addition to articles on impacts of Bill Gates' investment in the TWR and state of the TWR development from an interview with John Gilleland of Terra Power. (T. Tanaka)

Scoping and sensitivity analyses for the Demonstration Tokamak Hybrid Reactor (DTHR)

International Nuclear Information System (INIS)

Sink, D.A.; Gibson, G.

1979-03-01

The results of an extensive set of parametric studies are presented which provide analytical data of the effects of various tokamak parameters on the performance and cost of the DTHR (Demonstration Tokamak Hybrid Reactor). The studies were centered on a point design which is described in detail. Variations in the device size, neutron wall loading, and plasma aspect ratio are presented, and the effects on direct hardware costs, fissile fuel production (breeding), fusion power production, electrical power consumption, and thermal power production are shown graphically. The studies considered both ignition and beam-driven operations of DTHR and yielded results based on two empirical scaling laws presently used in reactor studies. Sensitivity studies were also made for variations in the following key parameters: the plasma elongation, the minor radius, the TF coil peak field, the neutral beam injection power, and the Z/sub eff/ of the plasma

Selected power reactor projects in Canada and the United States of America

Energy Technology Data Exchange (ETDEWEB)

NONE

1964-11-01

As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power.

Selected power reactor projects in Canada and the United States of America

International Nuclear Information System (INIS)

1964-01-01

As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power

Current status of radioactive waste management in Japan

International Nuclear Information System (INIS)

Amanuma, Tsuyoshi

1985-01-01

In Japan the nuclear power generation capacity now exceeds the level of 20,000 MW, 24.3 % of the total power generation. It constitutes the major position of energy source, a substitute for a petroleum. In the nuclear power, chemical engineering contributes significantly to treatment and disposal of the radioactive wastes. In the interim report by an ad hoc committee in the Atomic Energy Commission, for the future, rational grouping of the wastes and the direction of land disposal are stated. Contents are the following: basic ideas for the radioactive wastes, radioactive wastes countermeasures in Japan (wastes classification, low and high level and transuranic wastes), radioactive wastes in the nuclear fuel cycle (reactor and fuel reprocessing and reactor dismantling wastes). (Mori, K.)

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Forced vibration tests on the reactor building of a nuclear power station, 1

International Nuclear Information System (INIS)

Takeda, Toshikazu; Tsunoda, Tomohiko; Wakamatsu, Kunio; Kaneko, Masataka; Nakamura, Mitsuru; Kunoh, Toshio; Murahashi, Hisahiro

1988-01-01

Tsuruga Unit No.2 Nuclear Power Station of the Japan Atomic Power Company is the first PWR-type 4-loop plant constructed in Japan with a prestressed concrete containment vessel (PCCV). This report describes forced vibration tests carried out on the reactor building of this plant. The following were obtained as results: (1) The results of the forced vibration tests corresponded well on the whole with design values. (2) The vibration characteristics of the PCCV observed in the tests after prestressing are no different from the ones before prestressing. This shows that the vibration properties of the PCCV are practically independent of prestressing loads. (3) A seismic response analysis of the design basis earthquake was made on the design model reflecting the test results. The seismic safety of the plant was confirmed by this analysis. (author)

PA activity by using nuclear power plant safety demonstration and analysis

International Nuclear Information System (INIS)

Tsuchiya, Mitsuo; Kamimae, Rie

1999-01-01

INS/NUPEC presents one of Public acceptance (PA) methods for nuclear power in Japan, 'PA activity by using Nuclear Power Plant Safety Demonstration and Analysis', by using one of videos which is explained and analyzed accident events (Loss of Coolant Accident). Safety regulations of The National Government are strictly implemented in licensing at each of basic design and detailed design. To support safety regulation activities conducted by the National Government, INS/NLTPEC continuously implement Safety demonstration and analysis. With safety demonstration and analysis, made by assuming some abnormal conditions, what impacts could be produced by the assumed conditions are forecast based on specific design data on a given nuclear power plants. When analysis results compared with relevant decision criteria, the safety of nuclear power plants is confirmed. The decision criteria are designed to help judge if or not safety design of nuclear power plants is properly made. The decision criteria are set in the safety examination guidelines by taking sufficient safety allowance based on the latest technical knowledge obtained from a wide range of tests and safety studies. Safety demonstration and analysis is made by taking the procedure which are summarized in this presentation. In Japan, various PA (Public Acceptance) pamphlets and videos on nuclear energy have been published. But many of them focused on such topics as necessity or importance of nuclear energy, basic principles of nuclear power generation, etc., and a few described safety evaluation particularly of abnormal and accident events in accordance with the regulatory requirements. In this background, INS/NUPEC has been making efforts to prepare PA pamphlets and videos to explain the safety of nuclear power plants, to be simple and concrete enough, using various analytical computations for abnormal and accident events. In results, PA activity of INS/NUPEC is evaluated highly by the people

Progress report on fast breeder reactor development in Japan

International Nuclear Information System (INIS)

1979-01-01

The experimental fast reactor ''Joyo'' will be tested at 75 MW output, starting in April, 1980. In connection with the accident in the Three Mile Island plant, the reexamination of the plant safety and the rechecking-up of the maintenance control system were carried out, and the special inspection by the Science and Technology Agency was executed from May 21 to 23, 1979. Thereafter, the preparation for raising the power output was completed. The periodical inspection after the completion of 50 MW operation is being carried out. The state of progress of various equipments and the codes for core characteristic analysis is reported. The construction preliminary design (2) of the prototype reactor ''Monju'' is examined, and the same design (3) is prepared. The analysis of the decay heat in the prototype reactor is carried on for the safety licensing. The technological investigation of LMFBRs in foreign countries is under way. The preliminary design (4) of the demonstration reactor is under examination, and the technical specifications of the conceptual design (1) are prepared. The researches and developments of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structures and materials, safety and steam generators are reported. (Kako, I.)

The japan a nuclear power?

International Nuclear Information System (INIS)

Cumin, D.; Joubert, J.P.

2003-01-01

This work analyzes the Japan nuclear policy, in the frame of its foreign and safety policy in Pacific Asia, since the end of the cold war, especially the relations with the Usa and China. The Japan is a civil power because it has submitted the military institution to juridical restrictions and because it does not rely on the armed force to promote its national interests. The anti nuclear speech is joined with the acknowledgement of the dissuasion necessity, of the control of industrial processes and energy channels susceptible of military applications. Cultivating the ambiguity, the Japanese government can send a dissuasive message, perfectly legible, kind of communication of latent intimidation constituted by the virtual nuclear power of a state that takes part to the non proliferation treaty. (N.C.)

Project WAGR: The UK demonstration project for power reactor decommissioning - removing the core and looking to completion

International Nuclear Information System (INIS)

Benest, T. G.

2003-01-01

The United Kingdom Atomic Energy Authority (UKAEA) has built and operated a wide range of nuclear facilities since the late 1940's. UKAEA's present mission is to restore the environment of these facilities in a safe and environmentally responsible manner. This restoration includes the decommissioning of a number of redundant research and power reactors, one of which is the Windscale Advanced Gas-cooled Reactor (WAGR). Following shut down, UKAEA decided to continue the prototype function of the reactor into the decommissioning phase to develop dismantling techniques and establish waste routes. The reactor core and pressure vessel are now being dismantled in a programme of 10 campaigns, seven of which have been completed since 1998. It is anticipated that the current programme will be completed by summer 2005. This paper outlines the history of the reactor, the operation of the waste-processing route, the installed dismantling equipment and the successful completion of the first seven campaigns. This earlier work has been described in a number of publications and conferences, so this paper concentrates on recent work to select and develop cutting equipment to dismantle the core support structures and the pressure vessel. The decommissioning of the Windscale Advance Gas-cooled reactor is being undertaken to demonstrate that a power reactor can be decommissioned shortly after shutdown. The removal of the core and pressure vessel has been broken down into a series of 10 campaigns associated with particular core components. The first 7 campaigns have been successfully completed and the 8., is expected to commence in September 2003 17 months earlier than planned. Dismantling methodologies and tools have been developed specifically for each of these campaigns. Full-scale mock-ups have been used to test the tools, train the operators and assess the duration of operations. However, despite successful trials, operational experience has shown that some of these tools have not

Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-02-01

concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.

Development of multi-functional telerobotic systems for reactor dismantlement

International Nuclear Information System (INIS)

Fujii, Yoshio; Usui, Hozumi; Shinohara, Yoshikuni

1992-01-01

This report summarizes technological features of advanced telerobotic systems for reactor dismantling application developed at the Japan Atomic Energy Research Institute. Taking into consideration the special environmental conditions in reactor dismantling, major effort was made to develop multifunctional telerobotic system of high reliability which can be used to perform various complex tasks in an unstructured environment and operated in an easy and flexible manner. The system development was carried out through constructing three systems in seccession; a light-duty and a heavy-duty system as a prototype system for engineering test in cold environment, and a demonstration system for practical on-site application to dismantling highly radioactive reactor internals of an experimental boiling water reactor JPDR (Japan Power Demonstration Reactor). Each system was equipped with one or two amphibious manipulators which can be operated in either a push-button manual, a bilateral master-slave, a teach-and-playback or a programmed control mode. Different scheme was adopted in each system at designing the manipulator, transporter and man-machine interface so as to compare their advantages and disadvantages. According to the JPDR decommissioning program, the demonstration system was successfully operated to dismantle a portion of the radioactive reactor internals of the JPDR, which used underwater plasma arc cutting method and proved the usefulness of the multi-functional telerobotic system for reducing the occupational hazards and enhancing the work efficiency in the course of dismantling highly radioactive reactor components. (author)

Reactor power control system

International Nuclear Information System (INIS)

Tomisawa, Teruaki.

1981-01-01

Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

Situation of nuclear industry in Japan

International Nuclear Information System (INIS)

2002-08-01

This document is a reprint of a note published by the nuclear service of the French embassy in Japan. It evokes the present day situation of nuclear facilities in Japan, the public acceptance and its attitude in front of accidents, the national energy program, the deregulation and competitiveness of nuclear power, the carrying out of the nuclear program, the future reactors, the fast neutron reactors, the dismantling activities, the fuel enrichment and reprocessing of spent fuels, the use of MOX fuel, the off-site storage, the vitrified and radiological wastes, the geological disposal of wastes, the prospects of the nuclear program, the companies involved in the Japan nuclear industry, the French-Japanese bilateral cooperation, and the ITER project in the domain of nuclear fusion. (J.S.)

Japan Atomic Energy Research Institute in the 21st century

International Nuclear Information System (INIS)

Sato, Y.

2001-01-01

Major nuclear research institutes in Japan are the Japan Atomic Energy Research Institute (JAERI), Nuclear Cycle Development Institute (JNC), National Research Institute of Radiological Science (NIRS), and the Institute of Physical and Chemical Research (RIKEN). In the 50s and 60s JAERI concentrated on the introduction of nuclear technology from overseas. Energy security issues led to the development of a strong nuclear power programme in the next two decades resulting in Japan having 50 light water cooled nuclear power plants in operation. Japan also worked on other reactor concepts. The current emphasis of JAERI is on advanced reactors and nuclear fusion. Its budget of 270 million US$ supports five research establishments. JAERI has strong collaboration with industry and university system on nuclear and other advanced research topics (neutron science, photon science). In many areas Japan has strong international links. JAERI has also been transferring know-how on radioisotope and radiation applications to the developing countries particularly through IAEA-RCA mechanisms. (author)

Demonstration-informative center based on research reactor IR-50 in heat regime

International Nuclear Information System (INIS)

Krupenina, Ph.

2000-01-01

Many problems exist in the nuclear field, but the most significant one is the public's mistrust of Nuclear Energy. Strong downfalls of the radiological culture affect public perception, the main paradox being the situation after Chernobyl. The task of creating a Demonstration-Informative Center (Minatom RF) on reactor IR-50 research is conducted by Research and Development Institute of Power Engineering (ENTEK). The IR-50 is situated on the grounds of the institute. It will be a unique event when the functional reactor is situated in the center of the city. The purposes of the Demonstration-Informative Center are discussed. (authors)

Application of seismic isolation technology to demonstration FBR

International Nuclear Information System (INIS)

Kato, Muneaki

1994-01-01

The Japanese demonstration FBR is loop type, the intermediate heat exchanger is installed between the reactor and the steam generator, and up to the intermediate heat exchanger is in the containment vessel, which is designed as a reinforced concrete vessel. In FBRs, the optimization in aseismatic design and high temperature structural design is important. The reactor building is buried in rock bed up to its center of gravity to minimize the amplifying earthquake response. If the seismic isolation structure for a reactor is realized, cost reduction can be expected by the rationalization of machinery and equipment and the standardization of buildings and facilities. The research on FBR seismic isolation design has been carried out by Central Research Institute of Electric Power Industry and Japan Atomic Power Co. The concept of FBR seismic isolation design, the basic condition for the design evaluation, the research on safety allowance and the conceptual design analysis are reported. (K.I.)

Radioisotope production by reactors and cyclotrons in Japan

International Nuclear Information System (INIS)

Murakami, Yukio

1978-01-01

Present status of radioisotope production in Japan and the increasing demand from various fields are generally reviewed. Future problems associated with the shortage of economical supply are also discussed. The first half of this report is devoted to general review of the increasing demand for various radioisotopes from increasing number of users. The present status and future trends of the distribution of users of specific radioisotopes and their demands are shown. The remaining half of this report reviews the production with reactors and cyclotrons. The Japanese reactors producing radioisotopes are limited to low flux (10 13 ) research reactors at JAERI. Some problems associated with the improvement of availability and with the organizational structure are discussed. As for the production with cyclotrons, available facilities and the method of production are explained in detail. For clinical use, especially for the production of short lived radioisotopes, the advantage of a small special purpose cyclotron at each medical organization is emphasized. (Aoki, K.)

Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

International Nuclear Information System (INIS)

Zado, V.

1998-01-01

The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

Review of tokamak power reactor and blanket designs in the United States

International Nuclear Information System (INIS)

Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

1986-01-01

The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

Science.gov (United States)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

Energy Technology Data Exchange (ETDEWEB)

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

2016-10-01

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

Low power modular power generating reactors or Small Modular Reactors (SMR)

International Nuclear Information System (INIS)

Chenais, Jacques

2016-01-01

Electronuclear reactors were small reactors at the beginning, and then tend to be always bigger and more powerful, but since some recent times, several countries specialized in reactor design and fabrication (USA, Russia, China, and South Korea) have been developing Small Modular Reactors (SMR) of less than 300 MW. As France has already produced feasibility studies and is about to launch a SMR development programme, the author comments some specific aspects of this new architecture of reactors, characterises the targeted markets, gives an overview of the various more or less advanced existing concepts: a floating barge in Russia, the SMART 100 MW project in South Korea, several concepts in the USA (the mPower 125 MW, the NuScale 45 MW, the Westinghouse 225 MW, and the HI-SMUR 160 MW projects), the ACP 100 MW in China, the CAREM 27 MW in Argentina. French projects developed by the CEA, EDF, Areva and DCNS are then presented

ASTRID, Generation IV advanced sodium technological reactor for industrial demonstration

International Nuclear Information System (INIS)

Gauche, F.

2013-01-01

ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) is an integrated technology demonstrator designed to demonstrate the operability of the innovative choices enabling fast neutron reactor technology to meet the Generation IV criteria. ASTRID is a sodium-cooled fast reactor with an electricity generating power of 600 MWe. In order to meet the generation IV goals, ASTRID will incorporate the following decisive innovations: -) an improved core with a very low, even negative void coefficient; -) the possible installation of additional safety devices in the core. For example, passive anti-reactivity insertion devices are explored; -) more core instrumentation; -) an energy conversion system with modular steam generators, to limit the effects of a possible sodium-water reaction, or sodium-nitrogen exchangers; -) considerable thermal inertia combined with natural convection to deal with decay heat; -)elimination of major sodium fires by bunkerization and/or inert atmosphere in the premises; -) to take into account off-site hazards (earthquake, airplane crash,...) right from the design stage; -) a complete rethink of the reactor architecture in order to limit the risk of proliferation. ASTRID will also include systems for reducing the length of refueling outages and increasing the burn-up and the duration of the cycle. In-service inspection, maintenance and repair are also taken into account right from the start of the project. The ASTRID prototype should be operational by about 2023. (A.C.)

Development of fresh fuel packaging for ATR demonstration reactor

International Nuclear Information System (INIS)

Kurakami, J.; Kurita, I.

1993-01-01

Related to development of the demonstration advanced thermal reactor, it is necessary and important to develop transport packaging which is used for transporting fresh fuel assemblies. Therefore, the packaging is now being developed in Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, PNC is fabricating two prototype packagings based on the final design, and land cruising and vibration tests, handling performance tests and prototype packaging tests will be executed with prototype packagings in order to experimentally confirm the soundness of packaging and its contents and the propriety of design technique. This paper describes the summary of general specifications and structures of this packaging and the summary of preliminary safety analysis of package. (J.P.N.)

Identification of fast power reactivity effect in nuclear power reactor

International Nuclear Information System (INIS)

Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

1987-01-01

A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

Power oscillations in BWR reactors

International Nuclear Information System (INIS)

Espinosa P, G.

2002-01-01

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

Power generator in BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1984-01-01

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

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Use of self powered neutron detectors in the IEA-R1 reactor

International Nuclear Information System (INIS)

Galo Rocha, F. del.

1989-01-01

A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

High power ring methods and accelerator driven subcritical reactor application

Energy Technology Data Exchange (ETDEWEB)

Tahar, Malek Haj [Univ. of Grenoble (France)

2016-08-07

High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g., PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field imperfections on the

Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

International Nuclear Information System (INIS)

Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2002-07-01

The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

Design approaches for enhancing the engineering feasibility of tokamak power reactors

International Nuclear Information System (INIS)

Shannon, T.E.; Steiner, D.

1977-01-01

The design approach developed in the ORNL Fusion Power Demonstration Study is reviewed. The design concepts having greatest impact on reactor feasibility by the application of current or near term technology are described briefly. These are: blanket structural material, blanket coolant, power conversion system, and pulsed electrical system. Concepts relative to the approach taken to simplify the overall reactor design are listed

The European Fusion Energy Research Programme towards the realization of a fusion demonstration reactor

International Nuclear Information System (INIS)

Gasparotto, M.; Laesser, R.

2006-01-01

Since its inception, the European Fusion Programme has been orientated towards the establishment of the knowledge base needed for the definition of a reactor to be used for power production. Its ultimate goal is then to demonstrate the scientific and the technological feasibility of fusion power while incorporating the assessment of the safety, environmental, social and economic features of this type of energy source. At present, the JET device, the largest tokamak in the world, and the other medium-sized experimental machines are contributing essentially to the basic scientific phase of this development path. Their successful operation greatly contributed to support the design basis of ITER, the next step in fusion, which will aim to demonstrate the scientific and technical feasibility of fusion power production by achieving extended D-T burning plasma operation. Following ITER, the conception and construction of the DEMO device is planned. DEMO will be a demonstration power plant which will be the first fusion device to generate a significant amount of electrical power from fusion. This paper describes the status of fusion research and the European strategy for achievement of the ultimate goal of construction of a prototype reactor. (author)

The Optimization of power reactor control system

International Nuclear Information System (INIS)

Danupoyo, S.D.

1997-01-01

A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system

Development status on hydrogen production technology using high-temperature gas-cooled reactor at JAEA, Japan

International Nuclear Information System (INIS)

Shiozawa, Shusaku; Ogawa, Masuro; Hino, Ryutaro

2006-01-01

The high-temperature gas-cooled reactor (HTGR), which is graphite-moderated and helium-cooled, is attractive due to its unique capability of producing high temperature helium gas and its fully inherent reactor safety. In particular, hydrogen production using the nuclear heat from HTGR (up to 900 deg. C) offers one of the most promising technological solutions to curb the rising level of CO 2 emission and resulting risk of climate change. The interests in HTGR as an advanced nuclear power source for the next generation reactor, therefore, continue to rise. This is represented by the Japanese HTTR (High-Temperature Engineering Test Reactor) Project and the Chinese HTR-10 Project, followed by the international Generation IV development program, US nuclear hydrogen initiative program, EU innovative HTR technology development program, etc. To enhance nuclear energy application to heat process industries, the Japan Atomic Energy Agency (JAEA) has continued extensive efforts for development of hydrogen production system using the nuclear heat from HTGR in the framework of the HTTR Project. The HTTR Project has the objectives of establishing both HTGR technology and heat utilization technology. Using the HTTR constructed at the Oarai Research and Development Center of JAEA, reactor performance and safety demonstration tests have been conducted as planned. The reactor outlet temperature of 950 deg. C was successfully achieved in April 2004. For hydrogen production as heat utilization technology, R and D on thermo-chemical water splitting by the 'Iodine-Sulfur process' (IS process) has been conducted step by step. Proof of the basic IS process was made in 1997 on a lab-scale of hydrogen production of 1 L/h. In 2004, one-week continuous operation of the IS process was successfully demonstrated using a bench-scale apparatus with hydrogen production rate of 31 L/h. Further test using a pilot scale facility with greater hydrogen production rate of 10 - 30 m 3 /h is planned as

TerraPower, Bill Gates' reactor

International Nuclear Information System (INIS)

Guidez, J.

2016-01-01

TerraPower is a traveling wave reactor, it means that the reactor gradually converts non fissile material into the fuel it needs and the active part of the core progressively moves through the core leaving spent fuel behind. The last design of the TerraPower shows that it will use depleted uranium as fuel and that its core will need reloading every 10 years. Re-arrangement of the nuclear fuel will have to be made every 18 months to keep the core reactive. Metallic nuclear fuels will be used as they allow the highest breeding rates. It appears that apart from the very specific configuration of the core, the TerraPower is a reactor very similar to sodium-cooled fast reactors. Neutron transport inside traveling wave reactor core is complex and simulations show that the piling-up of fission product tends to kill the chain reaction and a continuous neutron addition may be necessary to keep the reactor going. A large part of the TerraPower feasibility studies concerns neutron transport inside its core. (A.C.)

Power reactor noise

International Nuclear Information System (INIS)

Thie, J.A.

1981-01-01

This book concentrates on the different types of noise present in power reactors and how the analysis of this noise can be used as a tool for reactor monitoring and diagnostics. Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions thus preventing further complications by alerting operators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussion of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

Compact reactor/ORC power source

International Nuclear Information System (INIS)

Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

1986-01-01

A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs

Correlations between power and test reactor data bases

International Nuclear Information System (INIS)

Guthrie, G.L.; Simonen, E.P.

1989-02-01

Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

Various problems in establishment of fuel cycle business in Japan

International Nuclear Information System (INIS)

Murata, Hiroshi

1985-01-01

Since Japan instituted the Atomic Energy Act in 1956, and organized the Atomic Energy Commission, as the fundamental policy of the peaceful use of atomic energy, the industrialization and establishment of fuel cycle technology have been advanced as well as the development of power reactors. The consistent and harmonious industrialization of uranium enrichment, fuel fabrication, reprocessing, the utilization of recovered plutonium and uranium, and the storage, treatment and disposal of wastes has been the target. As the nuclear power generation in Japan grew, the enhancement of the various factors of nuclear fuel cycle as the base of supporting nuclear power generation has become necessary. The effort of technical development has been continued in the fields of uranium enrichment, fuel reprocessing, plutonium fuel and waste treatment by the Power Reactor and Nuclear Fuel Development Corp., Japan Atomic Energy Research Institute and related industries. The plan and present status of nuclear fuel cycle business in Japan, the problems such as the roles of the government and private enterprises, technology transfer, the economy of nuclear fuel cycle business, the industrialization of mixed oxide fuel fabrication, nuclear nonproliferation policy and location are discussed. (Kako, I.)

NRC assessment of the Department of Energy annealing demonstration project

International Nuclear Information System (INIS)

Jackson, D.A.; Malik, S.N.

1997-01-01

Thermal annealing is the only known method for mitigating the effects of neutron irradiation embrittlement in reactor pressure vessel (RPV) steels. In May 1996, the US Department of Energy (DOE) in conjunction with the American Society of Mechanical Engineers, Westinghouse, Cooperheat, Electric Power Research Institute (with participating utilities), Westinghouse Owner's Group, Consumers Power, Electricite' de France, Duquesne Light and the Central Research Institute of the Electric Power Industry (Japan) sponsored an annealing demonstration project (ADP) at Marble Hill. The Marble Hill Plant, located in Madison, Indiana, is a Westinghouse 4 loop design. The plant was nearly 70% completed when the project was canceled. Hence, the RPV was never irradiated. The paper will present highlights from the NRCs independent evaluation of the Marble Hill Annealing Demonstration Project

Reactor power distribution monitor

International Nuclear Information System (INIS)

Sekimizu, Koichi

1980-01-01

Purpose: To improve the performance and secure the safety of a nuclear reactor by rapidly computing and display the power density in the nuclear reactor by using a plurality of processors. Constitution: Plant data for a nuclear reactor containing the measured values from a local power monitor LPRM are sent and recorded in a magnetic disc. They are also sent to a core performance computer in which burn-up degree distribution and the like are computed, and the results are sent and recorded in the magnetic disc. A central processors loads programs to each of the processors and applies data recorded in the magnetic disc to each of the processors. Each of the processors computes the corresponding power distribution in four fuel assemblies surrounding the LPRM string by the above information. The central processor compiles the computation results and displays them on a display. In this way, power distribution in the fuel assemblies can rapidly be computed to thereby secure the improvement of the performance and safety of the reactor. (Seki, T.)

Demonstration tokamak fusion power plant for early realization of net electric power generation

International Nuclear Information System (INIS)

Hiwatari, R.; Okano, K.; Asaoka, Y.; Shinya, K.; Ogawa, Y.

2005-01-01

A demonstration tokamak fusion power plant Demo-CREST is proposed as the device for early realization of net electric power generation by fusion energy. The plasma configuration for Demo-CREST is optimized to satisfy the electric breakeven condition (the condition for net electric power, P e net = 0 MW) with the plasma performance of the ITER reference operation mode. This optimization method is considered to be suitable for the design of a demonstration power plant for early realization of net electric power generation, because the demonstration power plant has to ensure the net electric generation. Plasma performance should also be more reliably achieved than in past design studies. For the plasma performance planned in the present ITER programme, net electric power from 0 to 500 MW is possible with Demo-CREST under the following engineering conditions: maximum magnetic field 16 T, thermal efficiency 30%, NBI system efficiency 50% and NBI current drive power restricted to 200 MW. By replacing the blanket system with one of higher thermal efficiency, a net electric power of about 1000 MW is also possible so that the performance of the commercial plant with Demo-CREST can also be studied from the economic point of view. The development path from the experimental reactor 'ITER' to the commercial plant 'CREST' through the demonstration power plant 'Demo-CREST' is proposed as an example of the fast track concept. (author)

Design study on evaluation for power conversion system concepts in high temperature gas cooled reactor with gas turbine

International Nuclear Information System (INIS)

Minatsuki, Isao; Mizokami, Yorikata

2007-01-01

The design studies on High Temperature Gas Cooled Reactor with Gas Turbine (HTGR-GT) have been performed, which were mainly promoted by Japan Atomic Energy Agency (JAEA) and supported by fabricators in Japan. HTGR-GT plant feature is almost determined by selection of power conversion system concepts. Therefore, plant design philosophy is observed characteristically in selection of them. This paper describes the evaluation and analysis of the essential concepts of the HTGR-GT power conversion system through the investigations based on our experiences and engineering knowledge as a fabricator. As a result, the following concepts were evaluated that have advantages against other competitive one, such as the horizontal turbo machine rotor, the turbo machine in an individual vessel, the turbo machine with single shaft, the generator inside the power conversion vessel, and the power conversion system cycle with an intercooler. The results of the study can contribute as reference data when the concepts will be selected. Furthermore, we addressed reasonableness about the concept selection of the Gas Turbine High Temperature Reactor GTHTR300 power conversion system, which has been promoted by JAEA. As a conclusion, we recognized the GTHTR300 would be one of the most promising concepts for commercialization in near future. (author)

Power reactors in Member States. 1978 edition

International Nuclear Information System (INIS)

1978-01-01

The computer-based reactor listing gives information on reactor core characteristics and plant systems for all power reactors in operation under construction and planned. The following two tables are included to give a general picture of the overall situation: Reactor types and net electrical power; Reactor units and net electrical power by country and cumulated by year

Dose estimates in Japan following the Chernobyl reactor accident

International Nuclear Information System (INIS)

Togawa, Orihiko; Homma, Toshimitsu; Iijima, Toshinori; Midorikawa, Yuji.

1988-02-01

Estimates have been made of the maximum individual doses and the collective doses in Japan following the Chernobyl reactor accident. Based on the measured data of ground deposition and radionuclide concentrations in air, raw milk, milk on sale and leafy vegetables, the doses from some significant radionuclides were calculated for 5 typical exposure pathways; cloudshine, groundshine, inhalation, ingestion of milk and leafy vegetables. The maximum effective dose equivalents for hypothetical individuals were calculated to be 1.8 mrem for adults, 3.7 mrem for children and 6.0 mrem for infants. The collective effective dose equivalent in Japan was estimated to be 5.8 x 10 4 man · rem; 0.50 mrem of the average dose per capita. (author)

Surveillance of nuclear power reactors

International Nuclear Information System (INIS)

Marini, J.

1983-01-01

Surveillance of nuclear power reactors is now a necessity imposed by such regulatory documents as USNRC Regulatory Guide 1.133. In addition to regulatory requirements, however, nuclear reactor surveillance offers plant operators significant economic advantages insofar as a single day's outage is very costly. The economic worth of a reactor surveillance system can be stated in terms of the improved plant availability provided through its capability to detect incidents before they occur and cause serious damage. Furthermore, the TMI accident has demonstrated the need for monitoring certain components to provide operators with clear information on their functional status. In response to the above considerations, Framatome has developed a line of products which includes: pressure vessel leakage detection systems, loose part detection systems, component vibration monitoring systems, and, crack detection and monitoring systems. Some of the surveillance systems developed by Framatome are described in this paper

Summary of space nuclear reactor power systems, 1983--1992

Energy Technology Data Exchange (ETDEWEB)

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

Summary of space nuclear reactor power systems, 1983--1992

International Nuclear Information System (INIS)

Buden, D.

1993-01-01

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power

Safeguarding on-power fuelled reactors - instrumentation and techniques

International Nuclear Information System (INIS)

Waligura, A.; Konnov, Y.; Smith, R.M.; Head, D.A.

1977-01-01

Instrumentation and techniques applicable to safeguarding reactors that are fuelled on-power, particularly the CANDU type, have been developed. A demonstration is being carried out at the Douglas Point Nuclear Generating Station in Canada. Irradiated nuclear materials in certain areas - the reactor and spent fuel storage bays - are monitored using photographic and television cameras, and seals. Item accounting is applied by counting spent-fuel bundles during transfer from the reactor to the storage bay and by placing these spent-fuel bundles in a sealed enclosure. Provision is made for inspection and verification of the bundles before sealing. The reactor's power history is recorded by a track-etch power monitor. Redundancy is provided so that the failure of any single piece of equipment does not invalidate the entire safeguards system. Several safeguards instruments and devices have beeen developed and evaluated. These include a super-8 mm surveillance camera system, a television surveillance system, a spent-fuel bundle counter, a device to detect dummy fuel bundles, a cover for enclosing a stack of spent-fuel bundles, and a seal suitable for underwater installation and ultrasonic interrogation. The information provided by these different instruments should increase the effectiveness of Agency safeguards and, when used in combination with other measures, will facilitate inspection at reactor sites

The disappointments for nuclear energy in Japan; Les deconvenues pour l'energie nucleaire au Japon

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

Several dysfunctions are reported in this paper: A reactor (Onagawa) closed after a nitrogen leakage; a small leakage of radioactive water in the nuclear power plant of Mihama assessment raised to five deaths, the operator stops its nuclear power plants for inspection, the Japan face to its ageing nuclear power plants, the truth about the cost of M.O.X., the seven reactors of Japan closed for inspection after cracks and leaks hidden to authorities, Tokai MURA accident. (N.C.)

Operating US power reactors

International Nuclear Information System (INIS)

Silver, E.G.

1988-01-01

This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

Present status and future perspective of research and test reactors in JAERI

International Nuclear Information System (INIS)

Baba, Osamu; Kaieda, Keisuke

1999-01-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

Present status and future perspective of research and test reactors in JAERI

Energy Technology Data Exchange (ETDEWEB)

Baba, Osamu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Kaieda, Keisuke

1999-08-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

The IAEA power reactor information system - PRIS

International Nuclear Information System (INIS)

Laue, H.J.; Qureshi, A.; Skjoeldebrand, R.; White, D.

1983-01-01

The IAEA Power Reactor Information System, PRIS, is based on a collection of basic design data and operating experience data which the IAEA started in 1970. PRIS is used for annual publications on 'Power Reactors in Member States', 'Operating Experience with Nuclear Power Stations in Member States', which gives annual operating information for individual plants, and a 'Performance Analysis Report' summarizing each year's and earlier experience. Since 1973 information has been collected in a systematic manner on significant plant outages (= more than 10 full power hours). There is now information on more than 10,000 outages in the system which permits some conclusions to be drawn both in regard to individual plants and to categories of plants on the significance of different outage reasons and different types of equipment failures. PRIS has not been intended to be a component reliability information system as an international data collection must stop short of the level of detail which would be needed for that purpose. The objectives of PRIS have been to provide a factual background for assumptions on parameters which are essential for economic evaluations and for systems operation planning (load factor and availability). The outage information does, however, lend itself to conclusions about generic problems in different categories of plants and it can be used by an individual operator to find other plants where information about particular problems can be obtained. It would also now be possible to use PRIS for setting availability goals based on experience and not only on theoretical design considerations. The paper demonstrates the conclusions which can be drawn from 662 reactor years of operation of light and heavy water pressurized reactors and 390 reactor years of boiling water reactors and, in particular, the role that the main heat removal system and its components have played in the equipment failure category

Technical Bases to Consider for Performance and Demonstration Testing of Space Fission Reactors

International Nuclear Information System (INIS)

Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

2004-01-01

Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as 'Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Will the test article accurately represent the flight system? Are the costs too restrictive?' have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems. (authors)

Building good relationships with neighbors of Japan's oldest plant, Tsuruga

International Nuclear Information System (INIS)

Hata, Emi

1992-01-01

Since its establishment in 1957 as a pioneer company of nuclear power development in Japan, the Japan Atomic Power Company (JAPC) has gained a great deal of experience with construction and operation of four nuclear power plants - one gas-cooled reactor, two boiling water reactors (BWRs), and one pressurized water reactor (PWR) - at two sites, Tsuruga and Tokai. To gain the understanding and cooperation of the local community, the Tsuruga station must keep running. Each employee is encouraged to make every possible effort not only to ensure the safe and reliable operation of the two units, but also to ensure conscientious coexistence and coprosperity within the local community. The Tsuruga office in the city and the Public Relations (PR) Pavilion (visitor's center) at the site work together as an open window of communication with the local community. Under these basic philosophies, various good neighbor activities are developed and carried out

Electric-power economy of Japan

International Nuclear Information System (INIS)

Dobrochotov, V.I.; Wolfberg, D.B.

1975-01-01

This is a survey on a) development and present capacity of electricity-supply companies in Japan, b) the structural shift in the capacity of power plants which took place from 1966 until 1974, arranged according to thermal, nuclear and hydraulic power stations, c) the structural shift in the use of fossile fuels, also from 1966 until 1974, d) the major thermal and nuclear power stations and pump storage plants under construction and in operation, e) interconnected operation. The survey ends with the development study of the Japanese Government being outlined. (GG/LN) [de

Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

2007-03-01

The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

Reference reactor module for NASA's lunar surface fission power system

International Nuclear Information System (INIS)

Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Werner, James; Qualls, Louis; Radel, Ross

2009-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO 2 -fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

Financing the electric power utilities, especially the nuclear power in Japan

International Nuclear Information System (INIS)

Tajima, T.

1975-04-01

Electric power demands in Japan have shown a remarkable growth at an annual rate of 12% since 1965. Nine electric power companies have invested large amounts of money so far, amounting to over 1 trillion yen every year since 1972. A survey of the electric power supply system and an estimation of the electric power demands in 1980 and in 1985 are given. It is expected that the main portion of electric power in the future will gradually be generated by nuclear plants. Financial features of the electrical power utilities, the credit risk of the electric power utilities, and the raising of funds by electric power utilities are discussed. It is concluded that it will be necessary (1) to expand the capital market, (2) to enable the electric power companies to issue a sufficient amount of bonds, (3) to make the Government financing institutions, such as the Japan Development Bank, provide the electric power companies with larger funds on a long-term and low-interest rate basis, and (4) even to take such drastic steps as subsidizing interest on private loans to the electric power companies. (B.P.)

Safeguarding on-power fuelled reactors - instrumentation and techniques

International Nuclear Information System (INIS)

Waligura, A.; Konnov, Y.; Smith, R.M.; Head, D.A.

1977-05-01

Instrumentation and techniques applicable to safeguarding reactors that are fuelled on-power, particularly the CANDU type, have been developed. A demonstration is being carried out at the Douglas Point Nuclear Generating Station in Canada. Irradiated nuclear materials in certain areas - the reactor and spent fuel storage bays - are monitored using photographic and television cameras, and seals. Item accounting is applied by counting spent-fuel bundles during transfer from the reactor to the storage bay and by placing these spent-fuel bundles in a sealed enclosure. Provision is made for inspection and verification of the bundles before sealing. The reactor's power history is recorded by a Track-Etch power monitor. Redundancy is provided so that the failure of any single piece of equipment does not invalidate the entire safeguards system. Several safeguards instruments and devices have been developed and evaluated. These include a super-8-mm surveillance camera system, a television surveillance system, a spent-fuel bundle counter, a device to detect dummy fuel bundles, a cover for enclosing a stack of spent-fuel bundles, and a seal suitable for underwater installation and ultrasonic interrogation. (author)

The electronuclear program of Japan

International Nuclear Information System (INIS)

Mori, Kazuhisa

1978-01-01

Japan, depending on imports for 80% of its energy supply, introduced a nuclear power program, which now comprises 14 reactors in operation (8,000 MW., that is 8% of its electricity production), and 15 reactors (14,000MW) under construction or being investigated. The objective for 1985 is from 26,000 to 33,000 MW. Japanese industry committed the error of placing too much confidence in its American licensers and not undertaking enough of its own research. This resulted in having a too small number of nuclear stations available, approximately 50%. Japan secured its uranium supply (Niger), its enrichment facilities (U.S.A. and Eurodif), while studying the centrifugation process, and facilities for reprocessing irradiated fuels (France and U.K.) while, at the same time, seeking national independence regarding its entire fuel cycle. The siting of nuclear power stations comes up against local opposition, which is being overcome by a taxation on electricity used to subsidize local collectivities favorable to the construction of a power station, whether hydro-electric, thermal or nuclear [fr

Seismic measures and defence in depth of nuclear power plant. Lessons learned from the great east Japan earthquake

International Nuclear Information System (INIS)

Ochiai, Kanehiro

2011-01-01

The Great East Japan Earthquake occurred in March 11, 2011 brought about severe accident at nuclear power plant, which gave significant lessons to nuclear experts concerned with safety measures. Concepts of defence in depth was basic philosophy to assure safety of nuclear power plant even against uncertainties exceeding design basis. This concept consisted of prevention, monitoring, and action to mitigate consequences of failures such as a series of physical barriers between the reactor core and the environment, which were called multiple safety systems, each with backup and designed to accommodate human error. As for natural disaster, depth of recognition of characteristic of natural phenomena and its effect and engineering judgment was of prime importance. Different waveforms of ground motion at Fukushima and Onagawa at the Great East Japan Earthquake showed that design ground motion should have large uncertainties. To cope with uncertainties of ground motion, robust seismic measures based on experience were such as design of static seismic intensity and rigid structure of natural period less than 0.1 sec. As for tsunami, defence in depth measures were prepared for the cooling of reactor core, spent fuel and related electric generation equipment with taking into account 1) time lag between tsunami generation and arrival, 2) tsunami affected area could be limited by coastal levee or anti-inundation measure, 3) system redundancy could be assured by different locations of equipments and 4) repair works could be done by shipment of replacement equipment from outside due to limitation of affected regional area. Success examples of Onagawa, Tokai unit 2, Fukushima Daiichi unit 6 and Fukushima Daini Nuclear Power Plants could suggest definite tsunami defence in depth measures. Containment vent system as final heat sink and emergency condenser as reactor core cooling at outage should be properly utilized for Fukushima Daiichi unit 1 Nuclear Power Plant. (T. Tanaka)

Human resource development in the beginning phase of nuclear technology development in Japan

International Nuclear Information System (INIS)

Yamashita, Kiyonobu

2015-01-01

Japan Research Reactor No.1 (JRR-1) was constructed as the first nuclear reactor in Japan and reached the first criticality in 1957. The construction of both the first BWR and the first PWR were started in the same year 1967 and they started power operation in the same year 1970. Engineers of electrical utilities and vendors gave efforts to have knowledge for reactor engineering mainly on the job training with high self-motivation to contribute for nuclear technology development. A part of them participated in the reactor engineering training course of the JAERI. (author)

Progress report on fast breeder reactor development in Japan, July - September 1977

International Nuclear Information System (INIS)

1978-06-01

As for the experimental fast breeder reactor ''Joyo'', the low power performance tests have been continued, and the measurements of reactor noise, the reactivity of fuel assemblies, power distribution and Na-void effect have been made. Efforts have been exerted to develop the required maintenance equipments, to manufacture the transfer rotor maintenance facilities, and to construct the spent fuel storing and cooling facilities. The analysis and calculation of the core characteristics have been in progress. The design work on the prototype fast breeder reactor ''Monju'' has been continued, and the development of the computer codes for the design has progressed. Informations have been gathered regarding the technological developments of LMFBRs overseas. The surveys on the site for ''Monju'' have been carried out. The design and research works on the demonstration reactor were started, and the general design factors such as the steam condition and the plant layout have been studied. As for the research and development of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structural materials, safety, and steam generators, the progresses are reported in detail. High performance neutron detectors for nuclear instrumentation have been under development, and the tagging gas method for fuel failure detection and location system has been tested. (Kako, I.)

Nuclear situation in Japan

International Nuclear Information System (INIS)

2006-01-01

This analysis takes stock on the nuclear situation in Japan. It discusses the ambitious equipment program in collaboration with the France, the destabilization of the japanese nuclear industry following the accidents and the energy policy evolutions. It presents the projects of the japanese nuclear industry: the Monju reactor restart, the Pluthermal project, the reprocessing power plant of Rokkasho Mura, the new reactors, the russian weapons dismantling, the ITER site selection and the buy out of Westinghouse by Toshiba. (A.L.B.)

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

Aging countermeasures for nuclear power plants in Japan and PLEC's activities

International Nuclear Information System (INIS)

Tajima, K.; Ishikawa, M.; Koyama, M.

2002-01-01

Full text: This summarizes aging countermeasure program of the nuclear power plants performed by the Japanese Government and industries, and describes current R and D program and utilization of the research results for the aged plants. The regulatory bodies (NISA of METI: the Nuclear Industry Safety Agency of the Ministry of Economy, Trade and Industry, MITI was reorganized to the Ministry of Economy, Trade and Industry (METI) in 2001) reviewed the results of technical assessments of aging phenomena of structures and components of nuclear power plants, conducted by the electric utilities. Major technical research projects related to aging of nuclear power plants in Japan have been conducted by Japan Power Engineering and Inspection Corporation (JAPEIC) under the asepses of the regulatory bodies. The results of the research projects were applied to the review and technical assessments. Environmental Fatigue Tests of Nuclear Power Plants Materials for Reliability Verification Project (EFT Project), has been conducting the fatigue tests for carbon steels, low alloy steels and austenitic steels under the simulated light water reactor (LWR) water environmental conditions. Based on the results, EFT project developed the models for evaluating fatigue initiation life reduction in LWR environment. The models were applied to the fatigue evaluation in the technical assessments mentioned above. On the other hand, the Nuclear Power Plant Life Engineering Center (PLEC) entrusted by NISA is conducting to coordinate planning of technology research on aging, to integrate technologies and knowledge of aging for management and assessment, to promote practical applications of results of technical research, and to promote sharing of information on aging technology. As to codes and standards the PLEC plans to develop them, based on the research results of the national projects performed by JAPEIC, with completion of the projects in order. Current planning technical standard is based to

An overview of in-service inspection for nuclear power plants in Japan

International Nuclear Information System (INIS)

Takano, Eishiro

1996-01-01

In nuclear power stations, for maintaining the safety of machinery and equipment, it is stipulated to carry out periodic inspection. The contents of the periodic inspection and the inspections of important matters are explained. In-service inspection (ISI) is the nondestructive inspection for confirming the soundness of reactor pressure vessels, main pipings, valves and others. The stipulations on the scope and time of the periodic inspection and ISI, the course of introducing ISI into Japan, and the basic way of thinking on ISI are described. Pre-service inspection (PSI) is carried out for the purpose of collecting and recording the initial data for ISI. The range of the objects of nondestructive inspection and leak test and the range of the objects of leak test only are shown. As to nondestructive inspection, the parts to be inspected, the frequency of inspection and the methods of inspection, and the method of leak test are described. As the present state of inspection technology, the automatic ultrasonic flow detectors for reactor vessels, piping welded parts and reactor vessel stud bolts, and the eddy current flaw detector for steam generator tubes are explained. (K.I.)

Nuclear power in Japan and the USA

International Nuclear Information System (INIS)

Titterton, E.

1979-06-01

The development of the nuclear power industry in Japan and the USA is discussed. The author lists the number of nuclear power plants operating, under construction and planned and considers the contribution made by nuclear power stations to the total electricity generated. The advantages of nuclear power to both countries are outlined and forecasts are made of the role to be played by nuclear power in future years

Catalogue of requirements for a plant-specific safety inspection of German nuclear power plants taking into account the Fukushima-I (Japan) events; Anforderungskatalog fuer anlagenbezogene Ueberpruefungen deutscher Kernkraftwerke unter Beruecksichtigung der Ereignisse in Fukushima-I (Japan)

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-03-30

The catalogue of requirements for a plant-specific safety inspection of German nuclear power plants taking into account the Fukushima-I (Japan) events worked out by the German RSK (reactor safety commission) includes the following inspection topics: natural events like earth quakes, floods, weather-based consequences and possible superposition; civilization-based events like airplane crash, gas release, reactor accident consequences for neighboring units, terroristic impacts, external attacks on computer-based control systems. Further event-independent assumptions have to be considered: station blackout, long-term emergency power supply requirement, failure of auxiliary cooling water supply, efficacy of preventive measures, aggravating boundary conditions for the performance of emergency measures.

International standardization of safety requirements for fast reactors

International Nuclear Information System (INIS)

2011-06-01

Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

Outline of geothermal power generation in Japan

Energy Technology Data Exchange (ETDEWEB)

Ezaki, Y

1960-01-01

The utilization of geothermal energy in electrical power generation throughout the world is described. Details of generating capacity and cost are given for Larderello, Italy; Wairakei, New Zealand: and the Geysers, USA. In Japan three types of conversion systems are used. These include the direct use of steam, direct use of hot water and binary fluid type systems. The history of Japanese investigation and exploitation of geothermal energy is reviewed and the status of the Matsukawa, Hakone, Otake and Takenoyu geothermal power plants is discussed. It is recommended that laws be enacted in Japan to encourage the development of this form of energy conversion.

Reactor cold neutron source facility, the first in Japan

International Nuclear Information System (INIS)

Utsuro, Masahiko; Maeda, Yutaka; Kawai, Takeshi; Tashiro, Tameyoshi; Sakakibara, Shoji; Katada, Minoru.

1986-01-01

In the Research Reactor Institute, Kyoto University, the first cold neutron source facility for the reactor in Japan was installed, and various tests are carried out outside the reactor. Nippon Sanso K.K. had manufactured it. After the prescribed tests outside the reactor, this facility will be installed soon in the reactor, and its outline is described on this occasion. Cold neutrons are those having very small energy by being cooled to about-250 deg C. Since the wavelength of the material waves of cold neutrons is long, and their energy is small, they are very advantageous as an experimental means for clarifying the structure of living body molecules and polymers, the atom configuration in alloys, and atomic and molecular movements by neutron scattering and neutron diffraction. The basic principle of the cold neutron source facility is to irradiate thermal neutrons on a cold moderator kept around 20 K, and to moderate and cool the neutrons by nuclear scattering to convert to cold neutrons. The preparatory research on cold neutrons and hydrogen liquefaction, the basic design to put the cold neutron source facility in the graphite moderator facility, the safety countermeasures, the manufacture and quality control, the operation outside the reactor and the performance are reported. The cold neutron source facility comprises a cold moderator tank and other main parts, a deuterium gas tank, a helium refrigerator and instrumentation. (Kako, I.)

HTTR demonstration test plan for industrial utilization of nuclear heat

International Nuclear Information System (INIS)

Sato, Hiroyuki; Ohashi, Hirofumi; Yan, Xing L.; Kubo, Shinji; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki

2014-09-01

Japan Atomic Energy Agency has been conducting research and development with a central focus on the utilization of High Temperature engineering Test Reactor (HTTR), the first High Temperature Gas-cooled Reactor (HTGR) in Japan, towards the realization of industrial use of nuclear heat. Several studies have made on the integration of the HTTR with thermochemical iodine-sulfur process and steam methane reforming hydrogen production plant (H 2 plant) as well as helium gas turbine power conversion system. In addition, safety standards for coupling a H 2 plant to a nuclear facility has been investigated. Based on the past design information, the present study identified test items to be validated in the HTTR demonstration test to accomplish a formulation of safety requirement and design consideration for coupling a H 2 plant to a nuclear facility as well as confirmation of overall performance of helium gas turbine system. In addition, plant concepts for the heat utilization system to be connected with the HTTR are investigated. (author)

Reactor control system. PWR

International Nuclear Information System (INIS)

2009-01-01

At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

Reactor power control device

International Nuclear Information System (INIS)

Kobayashi, Akira.

1980-01-01

Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

A review of fast reactor program in Japan (April 2001 - March 2002)

International Nuclear Information System (INIS)

Nagata, T.; Ieda, Y.

2002-01-01

This report describes the research and development activities on fast reactors in Japan thru April 2001 to March 2002. In December 2001, the Cabinet decided the Plan for Reorganization of Government-funded Corporations including the merger of JNC and the Japan Atomic Energy Research Institute (JAERI). A law to set up a new entity is supposed to be submitted to the National Diet by the Japanese Fiscal Year (JFY) 2004. In the Experimental Fast Reactor Joyo, thirty-five duty cycle operations and thirteen special tests with the MK-II core were completed by June 2000 without any fuel pin failures or serious plant trouble. The reactor is currently being upgraded to the MK-III core. Though a fire broke out in the maintenance building of Joyo in October 2001, the Mk-III construction work was restarted in February 2002. In the Prototype Fast Breeder Reactor Monju, countermeasures against sodium leakage have already been drawn up based on Monju comprehensive safety review. The safety licensing examination for the plant modification of Monju is undergoing. As for the Feasibility Study on Commercialized Fast Reactor Cycle Systems, JFY2001 was the first year of its second phase. A three-year period from JFY2001 to 2003 is the initial term of this phase. During this term, research activities are being focused on the design of the candidate concepts and fundamental tests of key technologies. An interim summary of these activities will be checked and reviewed, and based on the results; the research for JFY 2004 to 2005 will be conducted in order to narrow down the number of alternatives for the fast reactor cycle. (author)

Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

Science.gov (United States)

Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

2013-01-01

Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-07-01

A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

Characteristics of self-powered neutron detectors used in power reactors

International Nuclear Information System (INIS)

Todt, W.H.

1997-01-01

Self-Powered Neutron Detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors world-wide. The basic properties of these radiation sensors are described including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs which are being effectively used in in-core instrumentation systems for pressurised water, heavy water and graphite moderated light water reactors. Examples are also shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurised water and heavy water reactors worldwide. This paper is a summary of a new IEC standard to be issued in 1996 describing the characteristics and test methods of self-powered detectors used in nuclear power reactors. (author)

Power conditioning system for a nuclear reactor

International Nuclear Information System (INIS)

Higashigawa, Yuichi; Joge, Toshio.

1981-01-01

Purpose: To provide a power conditioning system for a BWR type reactor which has a function to be automatically operated within a range that the relationship between the heat power of the reactor and the electric power of an electric generator does not lose the safety of fuel by eliminating the unnecessary fluctuation of the power of the reactor. Constitution: A load request error signal fed from a conventional turbine control system to recirculation flow regulator is eliminated, and a reactor power conditioning system is newly provided, to which an electric generator power signal, a reactor average power area monitor signal and a load request signal are inputted. Thus, the load request signal is compared directly with the electric power of the electric generator, the recirculation flow rate is controlled by the compared result, and whether the correlation between the heat power of the reqctor and the electric power of the generator satisfies the correlation determined to prove the safety of fuel or not is checked. If this correlation is satisfied, the recirculation flow rate is merely automatically controlled. (Yoshino, Y.)

Status report of shielding investigation in Japan

International Nuclear Information System (INIS)

Shindo, M.

1964-01-01

The Japan Atomic Energy Research Institute (JAERI) was established in 1954, and immediately proceeded with the construction of a research reactor. The first symposium in Japan on nuclear energy was held in 1957. Most of the papers presented in the field of reactor shielding were limited to shielding materials and their fabrication. In the first stage of our investigations, our efforts were devoted to practical design studies of reactor shielding. As a result of these studies, it was found that the formulae at hand for calculations were inadequate, but at that time no electronic computer was available in Japan nor were theoretical calculations very actively undertaken. Problems on nuclear ship shielding had been investigated at the Ship Research Institute, since 1956 and many fruitful results had been obtained. About that time the Japan Atomic Industry Forum started activities and took the initiative in organizing shielding research. Research workers in the shipbuilding industry in particular have been seriously studying shielding problems. Few years after the first symposium, problems concerning more fundamental studies were treated by many research workers. Shielding experiments using radioisotopes were carried out and many fruitful results were obtained. They are described in the this paper. Medium size electronic computers became available in Japan, permitting a theoretical study group to make an active contribution. They produced some codes, and their results are also described in the following sections. This constituted the second stage of our investigations. A swimming-pool reactor, JRR-4 (Japan Research Reactor-4), has been under construction at JAERI since 1962 and will become critical in autumn 1964. After characteristic tests it will be a very powerful tool for the shielding investigations. This id the beginning of the third stage of investigations

Impacts on power reactor health physics programs

International Nuclear Information System (INIS)

Meyer, B.A.

1991-01-01

The impacts on power reactor health physics programs form implementing the revised 10 CFR Part 20 will be extensive and costly. Every policy, program, procedure and training lesson plan involving health physics will require changes and the subsequent retraining of personnel. At each power reactor facility, hundreds of procedures and thousands of people will be affected by these changes. Every area of a power reactor health physics program will be affected. These areas include; ALARA, Respiratory Protection, Exposure Control, Job Coverage, Dosimetry, Radwaste, Effluent Accountability, Emergency Planning and Radiation Worker Training. This paper presents how power reactor facilities will go about making these changes and gives possible examples of some of these changes and their impact on each area of power reactor health physics program

Thermionic reactors for space nuclear power

Science.gov (United States)

Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

1985-01-01

Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

Results on safety research for five years (from fiscal year 1996 to 2000). A field of power reactors

International Nuclear Information System (INIS)

2001-10-01

This safety research carried out by the Japan Nuclear Cycle Development Institute (JNC) for five years ranged from 1996 to 2000 fiscal year, was performed according to the safety research basic plan (from 1996 to 2000 fiscal year) established on March, 1996 (revised again on May, 2000). This report was arranged on a field on power reactors (all subjects on fields of advanced conversion reactor and a subject on power reactor in a field of seismic resistant and probability theoretical safety evaluation) by combining its research results for five years ranged from 1996 to 2000 fiscal year with general outlines on the safety research basic plan. Here were shown outlines on the safety research basic plan, aims and subjects on safety research at a field of power reactors, a list of survey sheets on safety research result, and survey sheets on safety research results. The survey sheets containing research field, title, organization, researcher name, researching period, names of cooperative organization, using facilities, research outline, research results, established contents, application, and research trends, are ranged to 5 items on advanced conversion reactor, 29 items on high breeder reactor, 1 item on seismic resistance, and 5 items on probability theoretical safety evaluation. (G.K.)

Ten years of KRB Gundremmingen demonstration power station

International Nuclear Information System (INIS)

Facius, H. von; Ettemeyer, R.

1976-01-01

In August 1976 the first large nuclear power station in the Federal Republic, the KRB Gundremmingen plant with a net power of 237 MWe, has been in operation ten years. The construction of KRB as a demonstration plant was a major step forward on the way to the economic utilization of nuclear power for German utilities. Design and operation of the plant have decisively influenced the further development of the technology of light water reactors in the Federal Republic. Unlike the Kahl Experimental Nuclear Power Station (VAK), which was a test facility designed to generate experience and to train personnel, the decision to build KRB from the outset was conditional upon the fulfillment of economic criteria. Here are some of the aspects in which KRB has greatly influenced the development of nuclear power station technology: first application of internal steam-water separation instead of a steam drum with a water content of the steam of less than 1%; construction of a reactor buildung with all the necessary safety factors; solution of the corrosion and erosion problems linked with the use of a saturated steam turbine; special measures taken to prevent the turbine from speeding up due to post-evaporation effects after shutdown. Detailed comments are devoted to the subjects of availability, causes of failure and repair work. (orig.) [de

Human resources in nuclear power program

International Nuclear Information System (INIS)

Machi, Sueo

2008-01-01

Nuclear power utilization within 2020 horizon is expanding in Asia, particularly in Japan, China, India, Republic of Korea, Vietnam and Indonesia. The nuclear energy policy iof Japan sees the increase of nuclear power contribution for energy security and to control CO 2 emission with the contribution ratio through the 21 st century kept at the current level of 30-40% or even higher. Japan expects its first reprocessing plant to be operational in 2007 and its first commercial fast breeder reactor operational in 2050. Starting with her experience with the operation of its first research reactor in 1957, a power demonstration reactor from USA in 1963; the first commercial 166 MW power plant from UK in 1966 and then its first commercial 375 MW light water reactor from USA in 1970, Japan developed her own nuclear reactor technology. Today, Japan has 55 operating nuclear power plants (NPPs) totaling 49 GW which supply 30% of its electricity needs. There are two NPPs under construction and 11 additional NPPs to be completed by 2017. Japan's experience showed that engineers in the nuclear, mechanical, electrical, material and chemical fields are needed to man their nuclear power plant. For the period 1958 to about 1970, there was a rapid increase in the number of students enrolled for their bachelor of science majoring in nuclear science and technology but this number of enrollees leveled off beyond 1970 up to 2002. For those pursuing their masters of science degree in this field, there was a steady but moderate rise in the number of students from 1958 to 2002. The population of students in the Ph.D program in nuclear science and technology had the lowest number of enrollees and lowest level of increase from 1958 to 2002. The courses offered at the university for nuclear power are nuclear reactor physics and engineering, nuclear reactor safety engineering and radiation safety. Prior to graduation, the students undergo training at a nuclear research institute, nuclear power

Reactor power control device

International Nuclear Information System (INIS)

Imaruoka, Hiromitsu.

1994-01-01

A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

A compact reactor/ORC power source

International Nuclear Information System (INIS)

Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

1986-01-01

A compact power source that combines an organic Rankine cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for componenet development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analyses combined with a finite element thermal analysis have aided in the power source design. The analysis have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high

Nuclear reactor power control device

International Nuclear Information System (INIS)

Koshi, Yuji; Sakata, Akira; Karatsu, Hiroyuki.

1987-01-01

Purpose: To control abrupt changes in neutron fluxes by feeding back a correction signal obtained from a deviation between neutron fluxes and heat fluxes for changing the reactor core flow rate to a recycling flow rate control system upon abrupt power change of a nuclear reactor. Constitution: In addition to important systems, that is, a reactor pressure control system and a recycling control system in the power control device of a BWR type power plant, a control circuit for feeding back a deviation between neutron fluxes and heat fluxes to a recycling flow rate control system is disposed. In the suppression circuit, a deviation signal is prepared in an adder from neutron flux and heat flux signals obtained through a primary delay filter. The deviation signal is passed through a dead band and an advance/delay filter into a correction signal, which is adapted to be fed back to the recycling flow rate control system. As a result, the reactor power control can be conducted smoothly and it is possible to effectively suppress the abrupt change or over shoot of the neutron fluxes and abrupt power change. (Kamimura, M.)

Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

International Nuclear Information System (INIS)

1978-01-01

Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

Cascade: a high-efficiency ICF power reactor

International Nuclear Information System (INIS)

Pitts, J.H.

1985-01-01

Cascade attains a net power-plant efficiency of 49% and its cost is competitive with high-temperature gas-cooled reactor, pressurized-water reactor, and coal-fired power plants. The Cascade reactor and blanket are made of ceramic materials and activation is 6 times less than that of the MARS Tandem Mirror Reactor operating at comparable power. Hands-on maintenance of the heat exchangers is possible one day after shutdown. Essentially all tritium is recovered in the vacuum system, with the remainder recovered from the helium power conversion loop. Tritium leakage external to the vacuum system and power conversion loop is only 0.03 Ci/d

The 33 years of research reactors in JAERI

International Nuclear Information System (INIS)

1990-11-01

The development and utilization of atomic energy in Japan began with the installation of JRR-1 reactor which attained the criticality in August, 1957, thus the third fire was lighted for the first time in Japan. JRR-2 was constructed as a full scale versatile research reactor, which attained the criticality in October, 1960, and since 1962, it has accomplished the role of the reactor for joint utilization. JRR-3 is the first reactor made in Japan by concentrating Japanese technologies in it, to develop and improve Japanese atomic energy technology. It attained the criticality in September, 1962, and has been used as a versatile research reactor. In 1960, Research Reactor Management Department was founded. JRR-4 was constructed as the research reactor for shielding for developing a nuclear-powered ship, which attained the criticality in January, 1965. The first hot laboratory in Japan for carrying out the post-irradiation test on the fuel and materials irradiated in these research reactors was installed in 1961. The JRR-1 was stopped in September, 1968, and is used as the commemorative exhibition hall. The JRR-3 was reconstructed, and attained the criticality in March, 1990, using 20 % enriched uranium fuel. The course of the research reactors for 33 years is reported. (K.I.)

Nuclear power plant with several reactors

Energy Technology Data Exchange (ETDEWEB)

Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

1972-05-10

A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

Power distribution forecasting device for reactors

International Nuclear Information System (INIS)

Tsukii, Makoto

1981-01-01

Purpose: To save expensive calculations on the forecasting of reactor power distribution. Constitution: Core status (CSD) such as entire coolant flow rate, pressures in the reactor, temperatures at the outlet and inlet and positions for control rods are inputted into a power distribution calculation device to calculate the power distribution based on physical models intermittently. Further, present power distribution is calculated based on in-core neutron flux measured values and CSD in a process control computer. Further, the ratio of the calculation results of the latter to those of the former is calculated, stored and inputted into a correction device to correct the forecast power distribution obtained by the power distribution calculation device. This enables to forecast the power distribution with excellent responsivity in the reactor site. (Furukawa, Y.)

Annual report of the Japan Atomic Energy Research Institute, for fiscal 1989

International Nuclear Information System (INIS)

1990-01-01

Japan Atomic Energy Research Institute has promoted the research on nuclear safety, the research and development of high temperature engineering and nuclear fusion which are the leading projects bringing about the breakthrough in atomic energy technology, the research on radiation utilization and the research and development of nuclear-powered ships, following the 'Plan of development and long term utilization of atomic energy' decided in 1987, as the central, general research institute in atomic energy field in Japan. Also the advanced basic research for opening atomic energy frontier and various international cooperation as well as the cooperation in Japan have been promoted. The engineering safety of nuclear facilities and environmental safety, the construction of the Nuclear Fuel Cycle Safety Engineering Research Facility, the design of the High Temperature Engineering Test Reactor and the various tests related to it, the reconstruction of JT-60 for increasing the current, the design of a nuclear fusion reactor, the high utilization of radiation using ion beam, the construction of Sekinehama Port for the nuclear-powered ship 'Mutsu', the power increasing test of the reactor of the Mutsu, the reconstruction of JRR-3 and others are reported. (K.I.)

Power distribution monitor in a nuclear reactor

International Nuclear Information System (INIS)

Uematsu, Hitoshi

1983-01-01

Purpose: To enable accurate monitoring for the reactor power distribution within a short time in a case where abnormality occurs in in-core neutron monitors or in a case where the reactor core state changes after the calibration for the neutron monitors. Constitution: The power distribution monitor comprises a power distribution calculator adapted to be inputted counted values from a reactor core present state data instruments and calculate the neutron flux distribution in the reactor core and the power distribution based on previously incorporated physical models, an RCF calculator adapted to be inputted with the counted values from the in-core neutron monitors and the neutron flux distribution and the power distribution calculated in the power distribution calculator and compensate the counted errors included in the counted values form the in-core neutron monitors and the calculation errors included in the power distribution calculated in the power distribution calculator to thereby calculate the power distribution within the reactor core, and an input/output device for the input of the data required for said power distribution calculator and the display for the calculation result calculated in the RCF calculator. (Ikeda, J.)

Power reactors in Member States. 1979 edition

International Nuclear Information System (INIS)

1979-01-01

This is the fifth issue of a periodic computer-based listing of nuclear power reactors, presenting the situation as of 1 May 1979. The basic design data for all reactors in operation, under construction, planned and shut down have been included. The following two tables are included to give a general picture of the overall situation: Table I: Reactor types and net electrical power. Table II: Reactor units and net electrical powered by country cummulated by year

Operating data monitoring and fatigue evaluation systems and findings for boiling water reactors in Japan

International Nuclear Information System (INIS)

Maekawa, O.; Kanazawa, Y.; Takahashi, Y.; Tani, M.

1995-01-01

The reactor pressure vessel (RPV) is one of the most critical components of a boiling water reactor (BWR) when utilities think about plant life extension (PLEX). Design stress analysis sometimes reports very high fatigue usage factors for such portions of RPVs as stud bolt, feedwater nozzle and support skirt.In order to evaluate design margin and to eliminate excessive conservatism in this design analysis to pave the way for PLEX, Japanese BWR utilities jointly with BWR manufacturers in Japan established a programme (1) to acquire plant operational data on line for specific parameters used in stress analysis, (2) to evaluate margin in the design using measured plant data best estimate boundary conditions for stress analysis, and (3) to establish a simplified fatigue analysis method for BWR RPV.A plant data acquisition system, named OPEDAS, has been developed and installed in Tokyo Electric Power Company's 1100MWe BWR at its Kashiwazaki-Kariwa nuclear power plant. Best estimate stress analysis using measured in-plant data has been carried out and the results show considerable margin in fatigue usage factor over the design. A simplified fatigue analysis method using in-plant data has been developed with the Green's function, although some limitations have been identified for its use. ((orig.))

Introduction to the neutron kinetics of nuclear power reactors

CERN Document Server

Tyror, J G; Grant, P J

2013-01-01

An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

Power reactor information system (PRIS)

International Nuclear Information System (INIS)

1989-06-01

Since the very beginning of commercial operation of nuclear power plants, the nuclear power industry worldwide has accumulated more than 5000 reactor years of experience. The IAEA has been collecting Operating Experience data for Nuclear Power Plants since 1970 which were computerized in 1980. The Agency has undertaken to make Power Reactor Information System (PRIS) available on-line to its Member States. The aim of this publication is to provide the users of PRIS from their terminals with description of data base and communication systems and to show the methods of accessing the data

Technical Evaluation on Ageing Management in Nuclear Power in Japan -Life Extension over 40 Years at Japanese Nuclear Power Plants-

International Nuclear Information System (INIS)

Tanaka, H.; Iwasaki, M.; Miyama, S.

2012-01-01

The Japanese commercial reactors have accumulated more than 40 years operating experience since Tsuruga-1 started commercial operation in March 1970. However, the severe accident occurred at Fukushima nuclear power plants triggered by East Japan Great Earthquake on March 11 last year. The facts that all the reactors having experienced core meltdown had operated for more than 30 years and Fukushima Daiichi unit 1 had just received the approval for its 40th year Ageing Management Technical Evaluation results from the Japanese government caused increasing distrust among the public in nuclear power plants operating for a long period of time. However, investigations of the accident conducted so far have not revealed any evidence that ageing degradation accelerated the accident. In addition, the analysis of seismic accelerations of the earthquake did not show that any component function was lost due to the accident. Considering these facts, I would like to discuss the issues to be continuously pursued and to be additionally implemented as part of the plat life management activities. In addition, I will introduce the efforts made by the Japanese utilities following the accident. (author)

Tendencies in operating power reactors

International Nuclear Information System (INIS)

Brinckmann, H.F.

1987-01-01

A survey is given about new tendencies in operating power reactors. In order to meet the high demands for control and monitoring of power reactors modern procedures are applicated such as the incore-neutron flux detection by means of electron emission detectors and multi-component activation probes, the noise diagnostics as well as high-efficient automation systems

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.

Market potential of heat utilization of modular HTR in Japan

International Nuclear Information System (INIS)

Ide, Akira; Tasaka, Kanji.

1993-01-01

HTR is considered to be the most suitable reactor type to use in the field other than power generation. So it is useful to know market potential of this type of reactor in Japan to justify its development. This potential was estimated to be about 400 200MWt modular HTR reactors. This number will be double if the market of hydrogen is developed. (J.P.N.)

Computerized reactor power regulation with logarithmic controller

International Nuclear Information System (INIS)

Gossanyi, A.; Vegh, E.

1982-11-01

A computerized reactor control system has been operating at a 5 MW WWR-SM research reactor in the Central Research Institute for Physics, Budapest, for some years. This paper describes the power controller used in the SPC operating mode of the system, which operates in a 5-decade wide power range with +-0.5% accuracy. The structure of the controller easily limits the minimal reactor period and produces a reactor transient with constant period if the power demand changes. (author)

Strategy of nuclear power in Korea, non-nuclear-weapon state and peaceful use of nuclear power

International Nuclear Information System (INIS)

Nagasaki, Takao

2005-01-01

The nuclear power plant started at Kori in Korea in April, 1978. Korea has carried out development of nuclear power as a national policy. The present capacity of nuclear power plants takes the sixes place in the world. It supplies 42% total power generation. The present state of nuclear power plant, nuclear fuel cycle facility, strategy of domestic production of nuclear power generation, development of next generation reactor and SMART, strategy of export in corporation with industry, government and research organization, export of nuclear power generation in Japan, nuclear power improvement project with Japan, Korea and Asia, development of nuclear power system with nuclear diffusion resistance, Hybrid Power Extraction Reactor System, radioactive waste management and construction of joint management and treatment system of spent fuel in Asia are stated. (S.Y.)

The working lifetime of nuclear power plants and new types of power reactors

International Nuclear Information System (INIS)

Bataille, Ch.; Birraux, C.

2003-01-01

The report on the working lifetime of nuclear power plants and new reactor types, by Mr Christian Bataille, deputy for the Nord, and Mr Claude Birraux, deputy for Haute-Savoie as well as President of the Office, supplements the studies carried out by the Parliamentary Office on the Safety of Nuclear Installations and Radioactive Wastes: it examines the remaining working life of the EDF nuclear power plants and the current status of projects that might, if circumstances were right, replace the reactors at present in service. The report investigates the different physical and other factors that influence the ageing of nuclear power plants and tackles the question of whether the design life of 40 years could be exceeded in practice. The whole issue of French nuclear power plant is put in perspective and compared with the situation of nuclear plants in Finland, Sweden, Germany and the United States, from the technical and regulatory standpoints. Believing that any attempt to optimise the working lifetime of the power plants currently in service must be accompanied by simultaneous moves aimed at their replacement, Messrs. Christian Bataille and Claude Birraux go on to review in detail the various light water reactor projects being proposed around the world for completion by 2015, as developments of existing models, in particular the EPR reactor of Framatome ANP, characterised by its competitiveness. They suggest that a first such reactor should be built as quickly as possible. Describing the other nuclear systems being investigated by research organisations not only in France but also in the United States and Sweden, Mrs. Christian Bataille and Claude Birraux review the objectives of these and the circumstances in which they might be developed, which would be unlikely to be before 2035 in view of the technological problems to be overcome and the industrial demonstration plants that would be needed

Demonstration tokamak-power-plant study (DEMO)

International Nuclear Information System (INIS)

1982-09-01

A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li 2 O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li 2 O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B 4 C, lead, and FE 1422 structural material

Power reactors operational diagnosis

International Nuclear Information System (INIS)

Dach, K.; Pecinka, L.

1976-01-01

The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

Small and medium power reactors 1987

International Nuclear Information System (INIS)

1987-12-01

This TECDOC follows the publication of TECDOC-347 Small and Medium Power Reactors Project Initiation Study - Phase I published in 1985 and TECDOC-376 Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power programme. It consists of two parts: 1) Guidelines for the Introduction of Small and Medium Power Reactors in Developing Countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of Small and Medium Power Reactors in developing countries; 2) Up-dated Information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex I of the above mentioned TECDOC-347. Figs

Reactor power system deployment and startup

International Nuclear Information System (INIS)

Wetch, J.R.; Nelin, C.J.; Britt, E.J.; Klein, G.; Rasor Associates, Inc., Sunnyvale, CA; California Institute of Technology, Pasadena)

1985-01-01

This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems. 5 references

Small and medium power reactors 1987

Science.gov (United States)

1987-12-01

This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

Japan's electric power industry: responding to the challenges of the 3Es

International Nuclear Information System (INIS)

Park, J.

1999-01-01

With the rapid push toward deregulation in the power markets of North America, the European Union, and emerging economies, the business environment of the global power market has been dramatically altered in recent years. Profit margins, strategic overseas investments, and shareholder equity have replaced stability and predictability as the new business paradigm in the international electric power industry. Although Japan's electric power industry has not undergone the same degree of market liberalization experienced by its counterparts in North America and the European Union, this does not mean that Japan has escaped the challenges of the 3Es (economic, energy, and environmental changes), which have transformed the international electric power industry in recent years. With companies and consumers in Japan paying the highest price for electricity in the industrialized world, the high price of electricity represents an important business competitiveness issue for firms in energy - and export-intensive industries. This is particularly true now that Japan is desperately trying to end the economic turmoil caused by over regulation and a weak banking system. What remains to be seen is if Japan will be able to deregulate its electric power industry and at the same time, comply with the Kyoto climate change pledge of reducing the country's greenhouse emissions by 6% (compared to 1990 levels) in the next 10 years and continue to meet its energy security objective of expanding the use of nuclear power in the national energy supply. The complex interplay of domestic and international pressures on the national power market need to be explored in order to fully understand the policy challenges facing Japan's electric power industry. (author)

Research and development for the future nuclear power

Energy Technology Data Exchange (ETDEWEB)

Morimoto, Hideo [Japan International Cooperation Agency, Tokyo (Japan)

2002-11-01

This paper consists of nuclear power technologies in Japan, its states of other countries, the today's objects, investment, change of the research and development paradigm, new type of reactor, public research and target research and resource. The new types of reactor investigated in Japan are FBR, 4S, aqueous homogenous reactor, gas reactor and molten-salt reactor. On the basis of correspondence to environment of market and materialization of business model, nuclear power has to cooperate with electric power side. The international joint research should be investigated, because the investment is limited. There are three references such as Report of nuclear power section in the total source energy investigation (2001): http://www.meti.go.jp/report/data/g10627aj.html, OECD/NEA (2002): http://www.neafr/html/ndd/reports/2002/nea3969.html and public research: http://www.iae.or.jp/koubo/koubo.html. (S.Y.)

Research and development for the future nuclear power

International Nuclear Information System (INIS)

Morimoto, Hideo

2002-01-01

This paper consists of nuclear power technologies in Japan, its states of other countries, the today's objects, investment, change of the research and development paradigm, new type of reactor, public research and target research and resource. The new types of reactor investigated in Japan are FBR, 4S, aqueous homogenous reactor, gas reactor and molten-salt reactor. On the basis of correspondence to environment of market and materialization of business model, nuclear power has to cooperate with electric power side. The international joint research should be investigated, because the investment is limited. There are three references such as Report of nuclear power section in the total source energy investigation (2001): http://www.meti.go.jp/report/data/g10627aj.html, OECD/NEA (2002): http://www.neafr/html/ndd/reports/2002/nea3969.html and public research: http://www.iae.or.jp/koubo/koubo.html. (S.Y.)

Operational method for demonstrating fuel loading integrity in a reactor having accessible 235U fuel

International Nuclear Information System (INIS)

Ward, D.R.

1979-07-01

The Health Physics Research Reactor is a small pulse reactor at the Oak Ridge National Laboratory. It is desirable for the operator to be able to demonstrate on a routine basis that all the fuel pieces are present in the reactor core. Accordingly, a technique has been devised wherein the control rod readings are recorded with the reactor at delayed critical and corrections are made to compensate for the effects of variations in reactor height above the floor, reactor power, core temperature, and the presence of any massive neutron reflectors. The operator then compares these readings with the values expected based on previous operating experience. If this routine operational check suggests that the core fuel loading might be deficient, a more rigorous follow-up may be made

Power controlling method for BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1983-01-01

Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

Research reactors for the social safety and prosperous neutron use

International Nuclear Information System (INIS)

Ito, Yasuo

2000-01-01

The present status of nuclear reactors in Japan and the world was briefly described in this report. Aiming to construct a background of stable future society dependent on nuclear energy, the necessity to establish an organization for research reactors in Japan was pointed out. There are a total of 468 reactors in the world, but only 248 of them are running at present and most of them are superannuated. In Japan, 15 research reactors are running and 8 of them are under collaborative utilization, but not a few of them have various problems. In the education of atomic energy, a reactor is dispensable for understanding its working principle through practice learning. Furthermore, a research reactor has important roles for development of power reactor in addition to various basic studies such as activation analysis, fission track, biological irradiation, neutron scattering, etc. Application of a reactor has been also progressing in industrial and medical fields. However, operation of the reactors has become more and more difficult in Japan because of a large running cost and a lack of residential consensus for nuclear reactor. Here, the author proposed an establishment of organization of research reactor in order to promote utilization of a reactor in the field of education, rearing of professionals and science and engineering. (M.N.)

Fracture toughness requirements of reactor vessel material in evaluation of the safety analysis report of nuclear power plants

International Nuclear Information System (INIS)

Widia Lastana Istanto

2011-01-01

Fracture toughness requirements of reactor vessel material that must be met by applicants for nuclear power plants construction permit has been investigated in this paper. The fracture toughness should be described in the Safety Analysis Reports (SARs) document that will be evaluated by the Nuclear Energy Regulatory Agency (BAPETEN). Because BAPETEN does not have a regulations or standards/codes regarding the material used for the reactor vessel, especially in the fracture toughness requirements, then the acceptance criteria that applied to evaluate the fracture toughness of reactor vessel material refers to the regulations/provisions from the countries that have been experienced in the operation of nuclear power plants, such as from the United States, Japan and Korea. Regulations and standards used are 10 CFR Part 50, ASME and ASTM. Fracture toughness of reactor vessel materials are evaluated to ensure compliance of the requirements and provisions of the Regulatory Body and the applicable standards, such as ASME or ASTM, in order to assure a reliability and integrity of the reactor vessels as well as providing an adequate safety margin during the operation, testing, maintenance, and postulated accident conditions over the reactor vessel lifetime. (author)

Systems aspects of a space nuclear reactor power system

Science.gov (United States)

Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.

1988-01-01

Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1979-09-01

Research activities in the Division of Reactor Engineering in fiscal 1978 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committees on Reactor Physics and in Decommissioning of Nuclear Facilities. (author)

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.)

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

Economic and energetic assessment for the situation in Japan after Fukushima

Energy Technology Data Exchange (ETDEWEB)

Ramos A, R.; Puente E, F., E-mail: rodrigo.ramos.aranda@gmail.com [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

In March 11 2011, a 9.0 mega thrust earthquake originated off the coast of Japan; as of now it has been the strongest ever recorded in the island. It produced a tsunami with waves over 15 meters high. The dimensions of the tsunami exceeded even the most pessimist scenarios for which the country was prepared. This happened in the Fukushima Daiichi nuclear power plant. Where there were containment walls built to withstand waves of up to 10 meters of height; the backup generation system was flooded, this left the plant disconnected from the grid and led to 7 degree nuclear accident in the Ines scale. Following the accident, the Japanese government decided to turn off all of its nuclear reactors and reevaluate the sector safety regulations. This decision had valid reasons and was backed up by most of the general public. However, the not so apparent consequences of turning the nuclear switch from one day to another need to be understood. In order to replace the energy produced by the nuclear power plants liquid natural gas and coal was imported for electric power generation. As of to day this extra imports cost more Japan than 90 billion dollars per year. this huge energy bill has led the country to perceive 20 month of trade deficits. This in addition to an exponential increase in greenhouse gases emissions and electricity prices so high that have forced hundreds of businesses to close, has demonstrated the importance of the nuclear sector in Japan. This situation left the Japanese government with no other option and in February 2014 it announced that they will restart their nuclear reactors. It is still uncertain what role will nuclear energy have in Japan. The objective of the following document is to show an analysis of the economical consequences of the Fukushima accident. There are many reports that address the environmental impacts, lessons learned from the accident and the changes in the safety regulations. In the contrary, the present highlights the importance

Economic and energetic assessment for the situation in Japan after Fukushima

International Nuclear Information System (INIS)

Ramos A, R.; Puente E, F.

2014-10-01

In March 11 2011, a 9.0 mega thrust earthquake originated off the coast of Japan; as of now it has been the strongest ever recorded in the island. It produced a tsunami with waves over 15 meters high. The dimensions of the tsunami exceeded even the most pessimist scenarios for which the country was prepared. This happened in the Fukushima Daiichi nuclear power plant. Where there were containment walls built to withstand waves of up to 10 meters of height; the backup generation system was flooded, this left the plant disconnected from the grid and led to 7 degree nuclear accident in the Ines scale. Following the accident, the Japanese government decided to turn off all of its nuclear reactors and reevaluate the sector safety regulations. This decision had valid reasons and was backed up by most of the general public. However, the not so apparent consequences of turning the nuclear switch from one day to another need to be understood. In order to replace the energy produced by the nuclear power plants liquid natural gas and coal was imported for electric power generation. As of to day this extra imports cost more Japan than 90 billion dollars per year. this huge energy bill has led the country to perceive 20 month of trade deficits. This in addition to an exponential increase in greenhouse gases emissions and electricity prices so high that have forced hundreds of businesses to close, has demonstrated the importance of the nuclear sector in Japan. This situation left the Japanese government with no other option and in February 2014 it announced that they will restart their nuclear reactors. It is still uncertain what role will nuclear energy have in Japan. The objective of the following document is to show an analysis of the economical consequences of the Fukushima accident. There are many reports that address the environmental impacts, lessons learned from the accident and the changes in the safety regulations. In the contrary, the present highlights the importance

Reactor power cutback system test experience at YGN 4

International Nuclear Information System (INIS)

Chi, Sung Goo; Kim, Se Chang; Seo, Jong Tae; Eom, Young Meen; Wook, Jeong Dae; Choi, Young Boo

1995-01-01

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

SP-100 space reactor power system readiness

International Nuclear Information System (INIS)

Josloff, A.T.; Matteo, D.N.; Bailey, H.S.

1992-01-01

This paper discusses the SP-100 Space Reactor Power System which is being developed by GE, under contract to the U.S. Department of Energy, to provide electrical power in the range of 10's to 100's of kW. The system represents an enabling technology for a wide variety of earth orbital and interplanetary science missions, nuclear electric propulsion (NEP) stages, and lunar/Mars surface power for the Space Exploration Initiative (SEI). The technology and design is now at a state of readiness to support the definition of early flight demonstration missions. Of particular importance is that SP-100 meets the demanding U.S. safety performance, reliability and life requirements. The system is scalable and flexible and can be configured to provide 10's to 100's of kWe without repeating development work and can meet DoD goals for an early, low-power demonstration flight in the 1996-1997 time frame

Reactor power automatically controlling method and device for BWR type reactor

International Nuclear Information System (INIS)

Murata, Akira; Miyamoto, Yoshiyuki; Tanigawa, Naoshi.

1997-01-01

For an automatic control for a reactor power, when a deviation exceeds a predetermined value, the aimed value is kept at a predetermined value, and when the deviation is decreased to less than the predetermined value, the aimed value is increased from the predetermined value again. Alternatively, when a reactor power variation coefficient is decreased to less than a predetermine value, an aimed value is maintained at a predetermined value, and when the variation coefficient exceeds the predetermined value, the aimed value is increased. When the reactor power variation coefficient exceeds a first determined value, an aimed value is increased to a predetermined variation coefficient, and when the variation coefficient is decreased to less than the first determined value and also when the deviation between the aimed value and an actual reactor power exceeds a second determined value, the aimed value is maintained at a constant value. When the deviation is increased or when the reactor power variation coefficient is decreased, since the aimed value is maintained at predetermined value without increasing the aimed value, the deviation is not increased excessively thereby enabling to avoid excessive overshoot. (N.H.)

Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1981-09-01

Research activities in the Division of Reactor Engineering in fiscal 1980 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

Power market restructuring in Asia : Russia, China, India, and Japan

International Nuclear Information System (INIS)

Hammons, T.J.; Zhong, J.; Mukhopadhyay, S.; Kurihara, I.

2008-01-01

Many countries are now in the process of deregulating their power industries in order to promote growth and competitiveness. This paper discussed power market restructuring activities in Russia, China, India and Japan. Economic convergence points in Russian and Asian power markets were reviewed. The state of Russia and China's power industry after the implementation of recent power restructuring initiatives was discussed. Technical characteristics of the industries were evaluated, and market development plans were outlined. Regional electricity markets in Asia were discussed, as well as issues related to domestic and foreign investment. Institutional reforms were reviewed, and individual outlines of revisions for the power industries of the 4 countries were presented. The study demonstrated that structuring processes vary from country to country. Differences in restructuring patterns were attributed to economic differences; country-specific features established within the electric power industry; and attitudes towards deregulation. It was concluded that the reforms adopted by the countries will lead to the expansion of national electric power systems. 23 refs., 4 tabs., 6 figs

Power Reactor Information System (PRIS)

International Nuclear Information System (INIS)

Spiegelberg, R.

1992-01-01

The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

Compact power reactor

International Nuclear Information System (INIS)

Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

Self-powered detectors for power reactors: an overview

International Nuclear Information System (INIS)

Ma, J.

2006-01-01

In this paper, Self-Powered Detectors (SPDs) for applications in nuclear power reactors have been reviewed. Based on their responses to radiation, these detectors can be divided into delayed response Self-Powered Neutron Detector (SPND), prompt response SPND and Self-Powered Gamma Detector (SPGD). The operational principles of these detectors are presented and their distinctive characteristics are examined accordingly. The analytical models and Monte Carlo method to calculate the responses of these detectors to neutron flux and external gamma rays are reviewed. The paper has also considered some related signal processing techniques, such as detector calibrations and detector signal compensations. Furthermore, a couple of failure modes have also been analyzed. Finally, applications of SPD in nuclear power reactors are summarized. (author)

Self-powered detectors for power reactors: an overview

Energy Technology Data Exchange (ETDEWEB)

Ma, J. [Univ. of Western Ontario, Dept. of Mechanical and Materials Engineering, London, Ontario (Canada)]. E-mail: jma64@uwo.ca

2006-07-01

In this paper, Self-Powered Detectors (SPDs) for applications in nuclear power reactors have been reviewed. Based on their responses to radiation, these detectors can be divided into delayed response Self-Powered Neutron Detector (SPND), prompt response SPND and Self-Powered Gamma Detector (SPGD). The operational principles of these detectors are presented and their distinctive characteristics are examined accordingly. The analytical models and Monte Carlo method to calculate the responses of these detectors to neutron flux and external gamma rays are reviewed. The paper has also considered some related signal processing techniques, such as detector calibrations and detector signal compensations. Furthermore, a couple of failure modes have also been analyzed. Finally, applications of SPD in nuclear power reactors are summarized. (author)

Fractals in Power Reactor Noise

International Nuclear Information System (INIS)

Aguilar Martinez, O.

1994-01-01

In this work the non- lineal dynamic problem of power reactor is analyzed using classic concepts of fractal analysis as: attractors, Hausdorff-Besikovics dimension, phase space, etc. A new non-linear problem is also analyzed: the discrimination of chaotic signals from random neutron noise signals and processing for diagnosis purposes. The advantages of a fractal analysis approach in the power reactor noise are commented in details

Development of telerobotic manipulators for reactor dismantling work

International Nuclear Information System (INIS)

Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

1991-01-01

This paper describes the amphibious electrical manipulators JARM-10, JART-25, JART-100 and JARM-25 which were developed in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. They are multi-functional telerobotic light-duty (10 and 25 daN) and heavy-duty (100 daN) Manipulators which can be used in hostile environments in reactor dismantling work such as high radiation, underwater work and electrical noise. Each manipulator can be operated in either a bilateral master-slave, a teach-and-playback or a programmed control mode. By combining these modes appropriately, it is possible to perform complex tasks of remote handling. The usefulness of the telerobotic systems for dismantling nuclear reactors has been demonstrated by successful application of the JARM-25 for remote underwater dismantlement of highly radioactive reactor internals of complex form of an experimental nuclear power reactor. (author)

Summary view on demonstration reactor safety

International Nuclear Information System (INIS)

Satoh, Kazuziro; Kotake, Shoji; Tsukui, Yutaka; Inagaki, Tatsutoshi; Miura, Masanori

1991-01-01

This work presents a summary view on safety design approaches for the demonstration fast breeder reactor (DFBR). The safety objective of DFBR is to be at lea as safe as a LWR. Major safety issues discussed in this paper are; reduction of sodium void reactivity worth, adoption of self-actuated mechanism in the backup shutdown system, use of the direct reactor auxiliary cooling system (DRACS), provision of the containment system. (author)

Nuclear power developments in the Asia-Pacific region

International Nuclear Information System (INIS)

Irwin, T.

2001-01-01

There are 438 nuclear power reactors operating in the world. Of these, 95 are in the Asia-Pacific region. Of the 36 reactors currently under construction in the world, 19 are in the Asia-Pacific region. Of the 44 planned reactors in the world, 36 are in this region. At the start of the 'New Nuclear Century' the Asia-Pacific region has become the main area for growth and innovation in nuclear power. This paper describes the nuclear power developments in each country and examines the status of the construction programme and the planned projects. Countries included are China, India, Japan, Democratic People's Republic of Korea (DPRK), Republic of Korea, Pakistan and Taiwan. New projects include the HTR in China, Advance Breeder Water Reactors in Japan, KEDO in the DPRK and the Advance Pressurised Water Reactor in the Republic of Korea

Nuclear Power Reactors in the World. 2013 Ed

International Nuclear Information System (INIS)

2013-01-01

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

Nuclear power reactors of new generation

International Nuclear Information System (INIS)

Ponomarev-Stepnoi, N.N.; Slesarev, I.S.

1988-01-01

The paper presents discussions on the following topics: fuel supply for nuclear power; expansion of the sphere of nuclear power applications, such as district heating; comparative estimates of power reactor efficiencies; safety philosophy of advanced nuclear plants, including passive protection and inherent safety concepts; nuclear power unit of enhanced safety for the new generation of nuclear power plants. The emphasis is that designers of new generation reactors face a complicated but technically solvable task of developing highly safe, efficient, and economical nuclear power sources having a wide sphere of application

Reference reactor module for NASA's lunar surface fission power system

Energy Technology Data Exchange (ETDEWEB)

Poston, David I [Los Alamos National Laboratory; Kapernick, Richard J [Los Alamos National Laboratory; Dixon, David D [Los Alamos National Laboratory; Werner, James [INL; Qualls, Louis [ORNL; Radel, Ross [SNL

2009-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

Molten salt reactors and possible scenarios for future nuclear power deployment

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Loiseaux, J. M.; Billebaud, A.; Brissot, R.; David, S.; Garzenne, C.; Laulan, O.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Meplan, O.; Michel-Sendis, F.; Nuttin, A.; Perdu, F.

2004-01-01

An important fraction of the nature energy demand may be satisfied by nuclear power. In this context, the possibilities of worldwide nuclear deployment are studied. We are convinced that the Molten Salt Reactors may play a central role in this deployment. The Molten Salt Reactor needs to be coupled to a reprocessing unit in order to extract the Fission Products which poison the core. The efficiency of this reprocessing has a crucial influence on reactor behavior especially for the breeding ratio. The Molten Salt Breeder Reactor project was based on an intensive reprocessing for high breeding purposes. A new concept of Thorium Molten Salt Reactor is presented here. Including this new concept in the worldwide nuclear deployment, to satisfy these power needs, we consider three typical scenarios, based on three reactor types: Pressurized Water Reactor, Fast Neutron Reactor and Thorium Molten Salt Reactor. The aim of this paper is to demonstrate, in a first hand that a Thorium Molten Salt Reactor can be realistic, with correct temperature coefficients and at least iso-breeder with slow reprocessing and new geometry; on the other hand that such Molten Salt Reactors enable a successful nuclear deployment, while minimizing fuel and waste management problems. (authors)

Nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-06-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Lifetimes of 7 to 10 yr at full power, at converter operating temperatures of 1275 to 1675 0 K, are being studied. The systems are being designed such that no single-failure modes exist that will cause a complete loss of power. In fact, to meet the long lifetimes, highly redundant design features are being emphasized. Questions have been raised about safety since the COSMOS 954 incident. ''Fail-safe'' means to prevent exposure of the population to radioactive material, meeting the environmental guidelines established by the U.S. Government have been and continue to be a necessary requirement for any space reactor program. The major safety feature to prevent prelaunch and launch radioactive material hazards is not operating the reactor before achieving the prescribed orbit. Design features in the reactor ensure that accidental criticality cannot occur. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit, where orbital lifetimes are practically indefinite, the safety considerations are negligible. Orbits below 400 to 500 nautical miles are the ones where a safety issue is involved in case of satellite malfunction. The potential missions, the question of why reactors are being considered as a prime power candidate, reactor features, and safety considerations will be discussed