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Sample records for safe research reactor

  1. Standards for safe operation of research reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The safety of research reactors is based on many factors such as suitable choice of location, design and construction according to the international standards, it also depends on well trained and qualified operational staff. These standards determine the responsibilities of all who are concerned with the research reactors safe operation, and who are responsible of all related activities in all the administrative and technical stages in a way that insures the safe operation of the reactor

  2. Safe operation of critical assemblies and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-09-15

    Some countries have accumulated considerable experience in the operation of these reactors and have in the process developed safe practices. On the other hand, other countries which have recently acquired, or will soon acquire, such reactors do not have sufficient background of experience with them to have developed full knowledge regarding their safe operation. In this situation, the International Atomic Energy Agency has considered that it would be useful to make available to all its Member States a set of recommendations on the safe operation of these reactors, based on the accumulated experience and best practices. The Director General accordingly nominated a Pane Ion Safe Operation of Critical Assemblies and Research Reactors to assist the Agency's Secretariat in drafting such recommendations

  3. Technical and safe development features of modern research reactor

    International Nuclear Information System (INIS)

    Wang Jiaying; Dong Duo

    1998-01-01

    The development trend of research reactor in the world, and development situation in China are introduced. Up to now, some research reactors have serviced for long time and equipment have aged, not to be satisfied for requirement of science and technology development. New research reactors must been developed. The technical features and safe features of new type research reactor in China, for example: multi-pile utilization, compact core of high flux, high automation level of control, reactor two independent shutdown systems, great coefficient of negative temperature, passive safety systems, reliable residual heat removal system are studied

  4. Safe decommissioning of the Romanian VVR-S research reactor

    International Nuclear Information System (INIS)

    Garlea, C.; Garlea, I.; Kelerman, C.; Rodna, A.

    2002-01-01

    The VVR-S Romania research reactor was operated between 1957-1997, at 2 MW nominal power, for research and radioisotopical production. The detailed decommissioning plan was developed between 1995-1998, in the frame of the International Atomic Energy Agency Technical assistance project ROM/9/017. The proposed strategy agreed by the counterpart as well as international experts was stage 1. In 1997, an independent analysis performed by European Commission experts, in the frame of PHARE project PH04.1/1994 was dedicated to the 'Study of Soviet Design Research Reactors', had consolidated the development of the project emphasizing technical options of safe management for radioactive wastes and VVR-S spent fuel. The paper presents the main technical aspects as well as those of social impact, which lead to the establishment of strategy for safe management of decommissioning. Technical analysis of the VVR-S reactor and associated radwaste facilities (Radioactive Waste Treatment Plant - Magurele and National Repository Baita-Bihor) proved the possibility of the classical method utilization for dismantling of the facility and treatment-conditioning-disposal of the arrised wastes in safe conditions. The decommissioning plan at stage 2 has been developed based on radiological safety assessment, evaluation of radwaste inventory (removed as well as preserved on site), cost analysis and environmental impact. Technical data were provided by the R and D programme including neutron calculations and experiments, radiological characterizing (for facility and its influence area), seismic analysis and environmental balance during the operation and after shut down of the reactor. A special chapter is dedicated to regulatory issues concerning the development of decommissioning under nuclear safety. Based on the Fundamental Norms of Radiological Safety, the Regulatory Body defined the clearance levels and safety criteria for the process. The development of National Norms for the

  5. Prospects for inherently safe reactors

    International Nuclear Information System (INIS)

    Barkenbus, J.N.

    1988-01-01

    Public fears over nuclear safety have led some within the nuclear community to investigate the possibility of producing inherently safe nuclear reactors; that is, reactors that are transparently incapable of producing a core melt. While several promising designs of such reactors have been produced, support for large-scale research and development efforts has not been forthcoming. The prospects for commercialization of inherently safe reactors, therefore, are problematic; possible events such as further nuclear reactor accidents and superpower summits, could alter the present situation significantly. (author)

  6. Safe Operation of Critical Assemblies and Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-05-15

    This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

  7. Safe Operation of Critical Assemblies and Research Reactors

    International Nuclear Information System (INIS)

    1961-01-01

    This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

  8. A logic scheme for regulating safe operation of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Effat, A; Rahman, F A [Operational Saety Dept, National Center of Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    This investigation presents a logic scheme for regulating the safe operation of research reactor in accordance with the new revision of SS-35 and revised by the 10 CFR. It emphasizes the regulatory inspection and enforcement (RI end E) during the reactor operation phase. IT is developed to provide information, guidance and recommendations to be taken when constructing the RI and E program that could be applied to the operational phase of the egyptian Research Reactors. In the operational phase, the regulatory inspection (RI) means an examination, observation, measurement, or test undertaken or on behalf of the nuclear regulatory body (NRB) during operation to verify that the nuclear materials, components, systems and structures as well operational activities, processes, procedures and personnel competence and performance are in accordance with the requirements established or the provisions approved by NRB or specified in the operational license or contained in regulations. Regulatory inspection includes both routine and non-routine ones. Any of them may be announced or unannounced. The problems identified by the RI must be resolved by the proper RE actions. The RE actions include investigative and corrective RE actions. These RI and E procedures for regulating safe operation of research reactors are presented as flow charts and then developed as a computer logic scheme. The software program is very efficient, very friendly, very simple and is interactive in nature such that the program asks the user certain questions about essential steps that guide the (RI and E) for research reactors, and user responds. The program proceeds based on this response until all the necessary steps for (RI and E) are accomplished. 5 figs.

  9. Assessment methodology applicable to safe decommissioning of Romanian VVR-S research reactor

    International Nuclear Information System (INIS)

    Baniu, O.; Vladescu, G.; Vidican, D.; Penescu, M.

    2002-01-01

    The paper contains the results of research activity performed by CITON specialists regarding the assessment methodology intended to be applied to safe decommissioning of the research reactors, developed taking into account specific conditions of the Romanian VVR-S Research Reactor. The Romanian VVR-S Research Reactor is an old reactor (1957) and its Decommissioning Plan is under study. The main topics of paper are as follows: Safety approach of nuclear facilities decommissioning. Applicable safety principles; Main steps of the proposed assessment methodology; Generic content of Decommissioning Plan. Main decommissioning activities. Discussion about the proposed Decommissioning Plan for Romanian Research Reactor; Safety risks which may occur during decommissioning activities. Normal decommissioning operations. Fault conditions. Internal and external hazards; Typical development of a scenario. Features, Events and Processes List. Exposure pathways. Calculation methodology. (author)

  10. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    International Nuclear Information System (INIS)

    Faghihi, F.; Mirvakili, S.M.

    2009-01-01

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ ex ), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10 3 Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  11. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of)

    2009-06-15

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ({rho}{sub ex}), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10{sup 3}Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  12. Intrinsically Safe and Economical Reactor (ISER)

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki; Asahi, Yoshiro

    1991-01-01

    The Intrinsically Safe and Economical Reactor (ISER) is designed based on the principle of a process inherent ultimate safe reactor, PIUS, a so-called inherently safe reactor (ISR). ISER has been developed joingly by the members of the Kanagawa Institute of Technology, the University of Tokyo, the Japan Atomic Energy Research Institute (JAERI) and several industrial firms in Japan. This paper describes the requirements for the next generation of power reactor, the safety design philosphy of ISR and ISER, the controllability of ISER and the results of analyses of some of the design-based accidents (DBA) of ISER, namely station blackout, accidents in which the pressurizer relief valve becomes jammed and stuck in open position and tube breaks in the steam generator. It is concluded that the ISER can ensure a wide range of contraollabitily and fuel integrity for all the analysed DBAs. (orig.)

  13. Proposition of innovative and safe design of grid plate for Tehran research reactor

    International Nuclear Information System (INIS)

    Jalali, H.R.; Fadaei, A.H.

    2017-01-01

    Highlights: • An innovative and safe design for grid plate in research reactors proposed. • New grid plate acts as an independent shutdown system. • Neutronic and transient calculation was done using MTR-PC package. • Calculations show that the performance and safety of new design are acceptable. - Abstract: The purpose of this paper is to propose an innovative and safe design of grid plate for Tehran research reactor (TRR) without any reduction in its performance in comparison with the current operation. The new grid plate consisted of two joined cubic with empty walls which are place of fuels and heavy water, respectively. The proposed design is such that the reactor core is divided into two distinct parts using the heavy water. The heavy water is inserted in the walls of the new grid plate. The new design of grid plate by keeping the characteristics of the previous version creates the possibility of shutting the reactor down in critical condition. In this paper, at initial step, a simulation of acceptable benchmark for Tehran research reactor is performed which could be considered reliable and comparable with SAR (Safety Analysis Report) data. In the next step, two different designs are proposed for grid plate and then are applied to reactor core using simulation tools. For the proposed design: core excess reactivity, shutdown margin, control rod worth, neutron flux and kinetic parameters are calculated. Furthermore, the transient analysis was performed for the new design to check the status of reactor safety. Obtained results show that all neutronic parameters for the first operating core and the new design are comparable, and there is no reduction in the efficiency of reference core. Moreover, in the current design, a diverse and independent shutdown system for TRR was included. Nuclear reactor analysis codes including MTR-PC package were employed to carry out these calculations.

  14. Safe operation of research reactors and critical assemblies. Code of practice and annexes. 1984 ed

    International Nuclear Information System (INIS)

    1984-01-01

    The safe operation of research reactors and critical assemblies (hereafter termed 'reactors') requires proper design, construction, management and supervision. This Code of Practice deals mainly with management and supervision. The provisions of the Code apply to the whole life of the reactor, including modification, updating and upgrading. The Code may be subject to revision in the light of experience and the state of technology. The Code is aimed at defining minimum requirements for the safe operation of reactors. Emphasis is placed on which safety requirements should be met rather than on specifying how these requirements may be met. The Code also provides guidance and information to persons and authorities responsible for the operation of reactors. The Code recommends that documents dealing with the operation of reactors and including safety analyses be prepared and submitted for review and approval to a regulatory body. Operation would be authorized on the understanding that it would comply with limits and conditions designed to ensure safety. The Code covers a wide range of reactor types, which gives rise to a variety of safety issues. Safety issues applicable to specific reactor types only (e.g. fast reactors) are not necessarily covered in this Code. Some of the recommendations in the Code are not directly applicable to critical assemblies. A recommendation may therefore be interpreted according to the type of reactor concerned. In such cases the words 'adequate' and 'appropriate' are used to mean 'adequate' or 'appropriate' for the type of reactor under consideration.

  15. Inherently safe light water reactors

    International Nuclear Information System (INIS)

    Ise, Takeharu

    1987-01-01

    Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

  16. Method of safely operating nuclear reactor

    International Nuclear Information System (INIS)

    Ochiai, Kanehiro.

    1976-01-01

    Purpose: To provide a method of safely operating an nuclear reactor, comprising supporting a load applied to a reactor container partly with secondary container facilities thereby reducing the load borne by the reactor container when water is injected into the core to submerge the core in an emergency. Method: In a reactor emergency, water is injected into the reactor core thereby to submerge the core. Further, water is injected into a gap between the reactor container and the secondary container facilities. By the injection of water into the gap between the reactor container and the secondary container facilities a large apparent mass is applied to the reactor container, as a result of which the reactor container undergoes the same vibration as that of the secondary container facilities. Therefore, the load borne by the reactor container itself is reduced and stress at the bottom part of the reactor container is released. This permits the reactor to be operated more safely. (Moriyama, K.)

  17. Inherently safe reactors

    International Nuclear Information System (INIS)

    Maartensson, Anders

    1992-01-01

    A rethinking of nuclear reactor safety has created proposals for new designs based on inherent and passive safety principles. Diverging interpretations of these concepts can be found. This article reviews the key features of proposed advanced power reactors. An evaluation is made of the degree of inherent safety for four different designs: the AP-600, the PIUS, the MHTGR and the PRISM. The inherent hazards of today's most common reactor principles are used as reference for the evaluation. It is concluded that claims for the new designs being inherently, naturally or passively safe are not substantiated by experience. (author)

  18. Management of research reactor ageing

    International Nuclear Information System (INIS)

    1995-03-01

    As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs

  19. Management of research reactor ageing

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs.

  20. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  1. Fail-safe reactivity compensation method for a nuclear reactor

    Science.gov (United States)

    Nygaard, Erik T.; Angelo, Peter L.; Aase, Scott B.

    2018-01-23

    The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on the constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.

  2. Safe operation and maintenance of research reactor

    International Nuclear Information System (INIS)

    Munsorn, S.

    1999-01-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U 3 O 8 - Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  3. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

    1999-10-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  4. Ageing of research reactors

    International Nuclear Information System (INIS)

    Ciocanescu, M.

    2001-01-01

    Historically, many of the research institutions were centred on a research reactor facility as main technological asset and major source of neutrons for research. Important achievements were made in time in these research institutions for development of nuclear materials technology and nuclear safety for nuclear energy. At present, ageing of nuclear research facilities among these research reactors and ageing of staff are considerable factors of reduction of competence in research centres. The safe way of mitigation of this trend deals with ageing management by so called, for power reactors, Plant Life Management and new investments in staff as investments in research, or in future resources of competence. A programmatic approach of ageing of research reactors in correlation with their actual and future utilisation, will be used as a basis for safety evaluation and future spending. (author)

  5. Decommissioning of the Neuherberg Research Reactor (FRN)

    International Nuclear Information System (INIS)

    Demmeler, M.; Rau, G.; Strube, D.

    1982-01-01

    The Neuherberg Research Reactor is of type TRIGA MARK III with 1 MW steady state power and pulsable up to 2000 MW. During more than ten years of operation 12000 MWh and 6000 reactor pulses had been performed. In spite of its good technical condition and of permanent safe operation without any failures, the decommissioning of the Neuherberg research reactor was decided by the GSF board of directors to save costs for maintaining and personnel. As the mode of decommissioning the safe enclosure was chosen which means that the fuel elements will be transferred back to the USA. All other radioactive reactor components will be enclosed in the reactor block. Procedures for licensing of the decommissioning, dismantling procedures and time tables are presented

  6. Licensing issues for inherently safe fast reactors

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Lee, S.; Okrent, D.

    1986-01-01

    There has been considerable interest recently in a new generation of liquid metal reactor (LMR) concepts in the US. Some significant changes in regulatory philosophy will be required if the anticipated cost advantages of inherently safe designs are to be achieved. The defense in depth philosophy will need to be significantly re-evaluated in the context of inherently safe reactors. It is the purpose of this paper to begin such a re-evaluation of this regulatory philosophy

  7. Safety-related parameters for the MAPLE research reactor and a comparison with the IAEA generic 10-MW research reactor

    International Nuclear Information System (INIS)

    Carlson, P.A.; Lee, A.G.; Smith, H.J.; Ellis, R.J.

    1989-07-01

    A summary is presented of some of the principle safety-related physics parameters for the MAPLE Research Reactor, and a comparison with the IAEA Generic 10-MW Reactor is given. This provides a means to assess the operating conditions and fuelling requirements for safe operation of the MAPLE Research Reactor under accepted standards

  8. Materials for passively safe reactors

    International Nuclear Information System (INIS)

    Simnad, T.

    1993-01-01

    Future nuclear power capacity will be based on reactor designs that include passive safety features if recent progress in advanced nuclear power developments is realized. There is a high potential for nuclear systems that are smaller and easier to operate than the current generation of reactors, especially when passive or intrinsic characteristics are applied to provide inherent stability of the chain reaction and to minimize the burden on equipment and operating personnel. Taylor, has listed the following common generic technical features as the most important goals for the principal reactor development systems: passive stability, simplification, ruggedness, case of operation, and modularity. Economic competitiveness also depends on standardization and assurance of licensing. The performance of passively safe reactors will be greatly influenced by the successful development of advanced fuels and materials that will provide lower fuel-cycle costs. A dozen new designs of advanced power reactors have been described recently, covering a wide spectrum of reactor types, including pressurized water reactors, boiling water reactors, heavy-water reactors, modular high-temperature gas-cooled reactors (MHTGRs), and fast breeder reactors. These new designs address the need for passive safety features as well as the requirement of economic competitiveness

  9. Siting of research reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

  10. Pioneering SUPER - Small Unit Passively-safe Enclosed Reactor - 15559

    International Nuclear Information System (INIS)

    Bhownik, P.K.; Gairola, A.; Shamim, J.A.; Suh, K.Y.; Suh, K.S.

    2015-01-01

    This paper presents the basic features of the Small Unit Passively-safe Enclosed Reactor abbreviated as SUPER, a new reactor system that has been designed and proposed at the Seoul National University's Department of Energy Systems Engineering. SUPER is a small modular reactor system or SMR that is cooled by sub-cooled as well as supercritical water. As a new member of SMRs, SUPER is a small-scale nuclear plant that is designed to be factory-manufactured and shipped as modules to be assembled at a site. The concept offers promising answers to many questions about nuclear power including proliferation resistance, waste management, safety, and startup costs. SUPER is a customized paradigm of a supercritical water reactor or SCWR, a type sharing commonalities with the current fleet of light water reactors, or LWRs. SUPER has evolved from the System-integrated Modular Advance Reactor, or SMART, being developed at the Korea Atomic Energy Research Institute, or KAERI. SUPER enhanced the safety features for robustness, design/equipment simplification for natural convection, multi-purpose application for co-generation flexibilities, suitable for isolated or small electrical grids, just-in-time capacity addition, short construction time, and last, but not least, lower capital cost per unit. The primary objectives of SUPER is to develop the conceptual design for a safe and economic small, natural circulation SCWR, to address the economic and safety attributes of the concept, and to demonstrate its technical feasibilities. (authors)

  11. Improvement of research reactor sustainability

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Paunoiu, C.; Toma, C.; Preda, M.; Ionila, M.

    2010-01-01

    The Research Reactors as is well known have numerous applications in a wide range of science technology, nuclear power development, medicine, to enumerate only the most important. The requirements of clients and stack-holders are fluctuating for the reasons out of control of Research Reactor Operating Organization, which may ensure with priority the safety of facility and nuclear installation. Sustainability of Research Reactor encompasses several aspects which finally are concentrated on safety of Research Reactor and economical aspects concerning operational expenses and income from external resources. Ensuring sustainability is a continuous, permanent activity and also it requests a strategic approach. The TRIGA - 14 MW Research Reactor detains a 30 years experience of safe utilization with good performance indicators. In the last 4 years the reactor benefited of a large investment project for modernization, thus ensuring the previous performances and opening new perspectives for power increase and for new applications. The previous core conversion from LEU to HEU fuel accomplished in 2006 ensures the utilization of reactor based on new qualified European supplier of TRIGA LEU fuel. Due to reduction of number of performed research reactors, the 14 MW TRIGA modernized reactor will play a significant role for the following two decades. (author)

  12. Code on the safety of nuclear research reactors: Operation

    International Nuclear Information System (INIS)

    1992-01-01

    The purpose of this publication is to provide the essential requirements and recommendations for the safe operation of research reactors, with emphasis on the supervisory and managerial aspects. However, the publication also provides some guidance and information on topics concerning all the organizations involved in operation. These objectives are expressed in terms of requirements and recommendations for the safe operation of research reactors. Emphasis is placed on the safety requirements that shall be met rather than on the ways in which they can be met. The requirements and recommendations may form the foundation necessary for a Member State to develop regulations and safety criteria for its research reactor programme.

  13. SIR - small is safe [in reactor design

    International Nuclear Information System (INIS)

    Hayns, M.

    1989-01-01

    A joint USA-UK venture has been initiated to design a small nuclear reactor which offers low capital cost, greater flexibility and a potentially lower environmental impact. Called Safe Integral Reactor (SIR), the lead unit could be built in the United Kingdom Atomic Energy Authority's (UKAEA's) Winfrith site if the design is accepted by the UK Nuclear Installations Inspectorate (NII). This article describes the 320 MWe reactor unit that is the basis of the design being developed. (author)

  14. Small intrinsically safe reactor implications

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1985-01-01

    Reviewing the history of nuclear power, it is found that peaceful uses of nuclear power are children of the war-like atom. Importance of special growth in a shielded environment is emphasized to exploit fully the advantages of nuclear power. Nuclear power reactors must be safe for their assimilation into society from the points of view of both technology and social psychology. ISR/ISER is identified as a missing link in the development of nuclear power reactors from this perspective and advocated for international development and utilization, being unleashed from the concerns of politicization, safety, and proliferation

  15. Improving nuclear safety at international research reactors: The Integrated Research Reactor Safety Enhancement Program (IRRSEP)

    International Nuclear Information System (INIS)

    Huizenga, David; Newton, Douglas; Connery, Joyce

    2002-01-01

    Nuclear energy continues to play a major role in the world's energy economy. Research and test reactors are an important component of a nation's nuclear power infrastructure as they provide training, experiments and operating experience vital to developing and sustaining the industry. Indeed, nations with aspirations for nuclear power development usually begin their programs with a research reactor program. Research reactors also are vital to international science and technology development. It is important to keep them safe from both accident and sabotage, not only because of our obligation to prevent human and environmental consequence but also to prevent corresponding damage to science and industry. For example, an incident at a research reactor could cause a political and public backlash that would do irreparable harm to national nuclear programs. Following the accidents at Three Mile Island and Chernobyl, considerable efforts and resources were committed to improving the safety posture of the world's nuclear power plants. Unsafe operation of research reactors will have an amplifying effect throughout a country or region's entire nuclear programs due to political, economic and nuclear infrastructure consequences. (author)

  16. ISAT promises fail-safe computer-based reactor protection

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    AEA Technology's ISAT system is a multiplexed microprocessor-based reactor protection system which has very extensive self-monitoring capabilities and is inherently fail safe. It provides a way of addressing software reliability problems that have tended to hamper widespread introduction of computer-based reactor protection. (author)

  17. Training and Certification of Research Reactor Personnel

    International Nuclear Information System (INIS)

    Zarina Masood

    2011-01-01

    The safe operation of a research reactor requires that reactor personnel be fully trained and certified by the relevant authorities. Reactor operators at PUSPATI TRIGA Reactor underwent extensive training and are certified, ever since the reactor first started its operation in 1982. With the emphasis on enhancing reactor safety in recent years, reactor operator training and certification have also evolved. This paper discusses the changes that have to be implemented and the challenges encountered in developing a new training programme to be in line with the national standards. (author)

  18. Related activities on management of ageing of Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lam, Pham Van [Reactor Dept., Nuclear Research Institute, Dalat (Viet Nam)

    1998-10-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the previous 250 kW TRIGA-MARK II reactor. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The renovated reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. Since then DNRR has been operated safely. Retained structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding are 35 years old. During the recent years, in-service inspection has been carried out, the reactor control and instrumentation system were renovated due to ageing and obsolescence of its components, reactor general inspection and refurbishment were performed. Efforts are being made to cope with ageing of old reactor components to maintain safe operation of the DNRR. (author)

  19. Inherently safe characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    1989-01-01

    This report is based on a detailed study which was carried out by Colenco (a company of the Motor-Columbus Group) on behalf of the Commission of the European Communities (CEC). It presents a summary of this study and concentrates more on the generic issues involved in the subject of inherent safety in nuclear power plants. It is assumed that the reader is reasonably familiar with the design outline of the systems included in the report. The report examines the role of inherent design features in achieving the safety of nuclear power plants as an alternative to the practice, which is largely followed in current reactors, of achieving safety by the addition of engineered safety features. The report examines current reactor systems to identify the extent to which their characteristics are either already inherently safe or, on the other hand, have inherent characteristics that require protective action to be taken. It then considers the advantages of introducing design changes to improve their inherent safety characteristics. Next, it looks at some new reactor types for which claims of inherent safety are made to see to what extent these claims are justified. The general question is then considered whether adoption of the inherently safe reactors would give advantages (by reducing risk in real terms or by improving the public acceptability of nuclear power) which are sufficient to offset the expected high costs and the technical risks associated with any new technology

  20. Development status of PIUS/ISER - a inherently safe reactor for the international use

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1987-01-01

    It is just in early 1980s that LWR-based nuclear power has become a substantial power source. Though the safety level of nuclear power is always claimed to be sufficiently high by the industry, it rests on the idea of defense in depth, the calculation by probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA). The TMI-2 and Chernobyl-4 accidents occurred in the industrially most advanced countries. In this paper, an alternative way to safe nuclear power is sought in so-called inherently safe reactors (ISR) including the LWR type PIUS/ISER. With proper consideration into the design of nuclear reactor plants, those can be made basically safe through the use of passive safe mechanism for their design. In short, an ISR is a nuclear power reactor which has passive and intrinsic core cooling capability and automatic shutdown capability. As the nuclear power reactors which are currently claimed to be inherently safe, there are the process inherent and ultimately safe reactor (PIUS) of ASEA-ATOM Sweden and the inherently safe and economical reactor (ISER) of the University of Tokyo, Japan, of LWR type. The current status of the development, the reliability, and some technical problems of ISER/PIUS and the attitude of various countries toward ISER/PIUS are described. (Kako, I.)

  1. Safe new reactor for radionuclide production

    International Nuclear Information System (INIS)

    Gray, P.L.

    1995-01-01

    In late 1995, DOE is schedule to announce a new tritium production unit. Near the end of the last NPR (New Production Reactors) program, work was directed towards eliminating risks in current designs and reducing effects of accidents. In the Heavy Water Reactor Program at Savannah River, the coolant was changed from heavy to light water. An alternative, passively safe concept uses a heavy-water-filled, zircaloy reactor calandria near the bottom of a swimming pool; the calandria is supported on a light-water-coolant inlet plenum and has upflow through assemblies in the calandria tubes. The reactor concept eliminates or reduces significantly most design basis and severe accidents that plague other deigns. The proven, current SRS tritium cycle remains intact; production within the US of medical isotopes such as Mo-99 would also be possible

  2. 105-H Reactor Interim Safe Storage Project Final Report

    International Nuclear Information System (INIS)

    Ison, E.G.

    2008-01-01

    The following information documents the decontamination and decommissioning of the 105-H Reactor facility, and placement of the reactor core into interim safe storage. The D and D of the facility included characterization, engineering, removal of hazardous and radiologically contaminated materials, equipment removal, decontamination, demolition of the structure, and restoration of the site. The ISS work also included construction of the safe storage enclosure, which required the installation of a new roofing system, power and lighting, a remote monitoring system, and ventilation components.

  3. Analysis of the interim safe storage of reactors at the Hanford site

    International Nuclear Information System (INIS)

    Wang Hailiang

    2014-01-01

    The nine production reactors, i.e. B, C, D, DR, F, H, KE, KW and N, at the Hanford site are all water-cooled and graphite-moderated reactors with natural uranium fuel. In 1993, the U.S. Department of Energy (DOE) decided to put eight production reactors (except for B) into Interim Safe Storage (ISS) for 75 years followed by deferred one-piece removal. Reactor B will remain as a national historical landmark. By the end of 2013, six reactors C, F, D, DR, H and N had been successfully put into the ISS. Reactors KE and KW will be put into the ISS in the coming years. Taking reactor C as an example, this paper mainly talks about how to put the production reactors in the Interim Safe Storage, e.g. how to make site preparation, how to construct the safe storage enclosure (SSE) and how to perform surveillance and maintenance during the ISS period, etc. (authors)

  4. Safety in the utilization and modification of research reactors

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide presents guidelines, approved by international consensus, for the safe utilization and modification of research reactors. While the Guide is most applicable to existing reactors, it is also recommended for use by organizations planning to put a new reactor into operation. 1 fig

  5. Iser: an international inherently safe reactor concept

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1988-01-01

    Iser is a modular standardised 200-300 MWe power reactor based on the PIUS principle. It differs from PIUS in being simpler, and making full use of existing steel-vessel-based LWR technology. Iser is an inherently safe reactor concept under development in Japan. It is a generic concept, not a patented commodity, and it is expected that an international association to develop the concept will be formed. (U.K.)

  6. Project management plan for Reactor 105-C Interim Safe Storage project

    International Nuclear Information System (INIS)

    Plagge, H.A.

    1996-09-01

    Reactor 105-C (located on the Hanford Site in Richland, Washington) will be placed into an interim safe storage condition such that (1) interim inspection can be limited to a 5-year frequency; (2) containment ensures that releases to the environmental are not credible under design basis conditions; and (3) final safe storage configuration shall not preclude or significantly increase the cost for any decommissioning alternatives for the reactor assembly.This project management plan establishes plans, organizational responsibilities, control systems, and procedures for managing the execution of Reactor 105-C interim safe storage activities to meet programmatic requirements within authorized funding and approved schedules

  7. Safe actinide disposition in molten salt reactors

    International Nuclear Information System (INIS)

    Gat, U.

    1997-01-01

    Safe molten salt reactors (MSR) can readily accommodate the burning of all fissile actinides. Only minor compromises associated with plutonium are required. The MSRs can dispose safely of actinides and long lived isotopes to result in safer and simpler waste. Disposing of actinides in MSRs does increase the source term of a safety optimized MSR. It is concluded that the burning and transmutation of actinides in MSRs can be done in a safe manner. Development is needed for the processing to handle and separate the actinides. Calculations are needed to establish the neutron economy and the fuel management. 9 refs

  8. Technical specifications: Health Physics Research Reactor

    International Nuclear Information System (INIS)

    1986-03-01

    These technical specifications define the key limitations that must be observed for safe operation of the Health Physics Research Reactor (HPRR) and an envelope of operation within which there is assurance that these limits will not be exceeded

  9. Decommissioning of the research reactors at the Russian Research Centre Kurchatov Institute

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoy, N.N.; Ryantsev, E.P.; Kolyadin, V.I.; Kucharkin, N.E.; Melkov, E.S.; Gorlinsky, Yu.E.; Kyznetsova, T.I.; Bulkin, B.K.

    2002-01-01

    The Kurchatov Institute is the largest research center of Russia in the field of nuclear science and engineering. It comprises more than 10 research institutes and scientific-technological complexes carrying out research work in the field of safe development of atomic engineering, controlled thermonuclear fusion, and plasma physics, nuclear physics and elementary particle physics, research reactors, radiation materials technology, solid state physics and superconductivity, molecular and chemical physics, and also perspective know-how's, information science and ecology. This report is basically devoted to the decommissioning of the research reactor installations, in particular to the reactor MR because of the volume and complexity of actions involved. (author)

  10. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H. [NUKEM GmbH, Alzenau (Germany)

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  11. Supply of enriched uranium for research reactors

    International Nuclear Information System (INIS)

    Mueller, H.

    1997-01-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA's ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future

  12. 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

  13. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  14. Proceedings of the workshop on intrinsically safe and economical reactors

    International Nuclear Information System (INIS)

    Livingston, R.S.

    1985-12-01

    This report presents the proceedings of a workshop concerning the design of inherently safe reactors. This paper emphasizes Japanese contributions to this subject, especially small reactors. Nine analytics were prepared for this report

  15. Conceptual design of multipurpose compact research reactor

    International Nuclear Information System (INIS)

    Nagata, Hiroshi; Kusunoki, Tsuyoshi; Hori, Naohiko; Kaminaga, Masanori

    2012-01-01

    Conceptual design of the high-performance and low-cost multipurpose compact research reactor which will be expected to construct in the nuclear power plant introduction countries, started from 2010 in JAEA and nuclear-related companies in Japan. The aims of this conceptual design are to achieve highly safe reactor, economical design, high availability factor and advanced irradiation utilization. One of the basic reactor concept was determined as swimming pool type, thermal power of 10MW and water cooled and moderated reactor with plate type fuel element same as the JMTR. It is expected that the research reactors are used for human resource development, progress of the science and technology, expansion of industry use, lifetime extension of LWRs and so on. (author)

  16. Probabilistic safety assessment for research reactors

    International Nuclear Information System (INIS)

    1986-12-01

    Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

  17. Design strategy for control of inherently safe reactors

    International Nuclear Information System (INIS)

    Chisholm, G.H.

    1984-01-01

    Reactor power plant safety is assured through a combination of engineered barriers to radiation release (e.g., reactor containment) in combination with active reactor safety systems to shut the reactor down and remove decay heat. While not specifically identified as safety systems, the control systems responsible for continuous operation of plant subsystems are the first line of defense for mitigating radiation releases and for plant protection. Inherently safe reactors take advantage of passive system features for decay-heat removal and reactor shutdown functions normally ascribed to active reactor safety systems. The advent of these reactors may permit restructuring of the present control system design strategy. This restructuring is based on the fact that authority for protection against unlikely accidents is, as much as practical, placed upon the passive features of the system instead of the traditional placement upon the PPS. Consequently, reactor control may be simplified, allowing the reliability of control systems to be improved and more easily defended

  18. Nuclear design of ISER [intrinsically safe and economical reactor

    International Nuclear Information System (INIS)

    Yamano, Naoki; Yokoyama, Takashi

    1985-01-01

    A preliminary core design work on ISER (Intrinsically Safe and Economical Reactor) based on the concept of the PIUS reactor of ASEA-ATOM is performed in order to grasp the characteristics of the reactor core and the fuel management scheme. Certain relations between the fuel specifications and the cycle length are estimated. Items of improvement on the ISER core characteristics and problems to be considered on the nuclear design are presented. Experiments to be considered are also discussed in conjunction with the development of experimental reactor (ISER-E)

  19. Safe Management Of Fast Reactors: Towards Sustainability

    International Nuclear Information System (INIS)

    Dreimanis, Andrejs

    2015-01-01

    An interdisciplinary systemic approach to socio-technical optimization of nuclear energy management is proposed, by recognizing a) the rising requirements to nuclear safety being realized using fast reactors (FR), b) the actuality to maintain and educate qualified workforce for fast reactors, c) the reactor safety and public awareness as the keystones for improving attitude to implement novel reactors. Knowledge management and informational support firstly is needed in: 1) technical issues: a) nuclear energy safety and reliability, b) to develop safe and economic technologies; 2) societal issues: a) general nuclear awareness, b) personnel education and training, c) reliable staff renascence, public education, stakeholder involvement, e).risk management. The key methodology - the principles being capable to manage knowledge and information issues: 1) a self-organization concept, 2) the principle of the requisite variety. As a primary source of growth of internal variety is considered information and knowledge. Following questions are analyzed indicating the ways of further development: a) threats in peaceful use of nuclear energy, b) basic features of nuclear risks, including terrorism, c) human resource development: basic tasks and instruments, d) safety improvements in technologies, e) advanced research and nuclear awareness improvement There is shown: public education, social learning and the use of mass media are efficient mechanisms forming a knowledge-creating community thereby reasoning to facilitate solution of key socio-technical nuclear issues: a) public acceptance of novel nuclear objects, b) promotion of adequate risk perception, and c) elevation of nuclear safety level and adequate risk management resulting in energetic and ecological sustainability. (author)

  20. Concept of an inherently-safe high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

    2012-01-01

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  1. Manual for the operation of research reactors

    International Nuclear Information System (INIS)

    1965-01-01

    The great majority of the research reactors in newly established centres are light-water cooled and are often also light-water moderated. Consequently, the IAEA has decided to publish in its Technical Reports Series a manual dealing with the technical and practical problems associated with the safe and efficient operation of this type of reactor. Even though this manual is limited to light-water reactors in its direct application and presents the practices and experience at one specific reactor centre, it may also be useful for other reactor types because of the general relevance of the problems discussed and the long experience upon which it is based. It has, naturally, no regulatory character but it is hoped that it will be found helpful by staff occupied in all phases of the practical operation of research reactors, and also by those responsible for planning their experimental use. 23 refs, tabs

  2. Research reactor collaboration in the Asia-Pacific region

    International Nuclear Information System (INIS)

    Jun, Byung Jin

    2006-01-01

    The number of research reactors over the world has been decreasing since its peak in the middle of the 1970s, and it is predicted to decrease more rapidly than before in the future. International collaboration on research reactors is an effective way for their continued safe service to human welfare in various technical areas. The number of new research reactors under construction or planned for in the Asia-Pacific region is the greatest in the world. Among the regional collaboration activities on research reactors, safety has been the most important subject followed by neutron activation analysis, radioisotope production and neutron beam applications. It is understood that more regional collaboration on basic technologies important for the safety, management and utilization of the research reactors is demanding. The new project proposal of the Forum for Nuclear Cooperation in Asia on 'Research Reactor Technology for Effective Utilization' is understood to meet the demands. Meanwhile, there is a consensus on the need for research reactor resource sharing in the region. As a result of the review on the international collaboration activities in the region, the author suggests a linkage between the above new project and IAEA/RCA project considering a possible sharing of research reactor resources in the region. (author)

  3. Proceedings of the 1997 workshop on the utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-10-01

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

  4. The experiences of research reactor accident to safety improvement

    International Nuclear Information System (INIS)

    Wiranto, S.

    1999-01-01

    The safety of reactor operation is the main factor in order that the nuclear technology development program can be held according the expected target. Several experience with research reactor incidents must be learned and understood by the nuclear program personnel, especially for operators and supervisors of RSG-GA. Siwabessy. From the incident experience of research reactor in the world, which mentioned in the book 'Experience with research reactor incidents' by IAEA, 1995, was concluded that the main cause of research reactor accidents is understandless about the safety culture by the nuclear installation personnel. With learn, understand and compare between this experiences and the condition of RSG GA Siwabessy is expended the operators and supervisors more attention about the safety culture, so that RSG GA Siwabessy can be operated successfull, safely according the expected target

  5. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  6. Fail-safe logic elements for use with reactor safety systems

    International Nuclear Information System (INIS)

    Bobis, J.P.; McDowell, W.P.

    1976-01-01

    A complete fail-safe trip circuit is described which utilizes fail-safe logic elements. The logic elements used are analog multipliers and active bandpass filter networks. These elements perform Boolean operations on a set of AC signals from the output of a reactor safety-channel trip comparator

  7. Twenty years of health physics research reactor operation

    International Nuclear Information System (INIS)

    Sims, C.S.; Gilley, L.W.

    1983-01-01

    The Health Physics Research Reactor at the Oak Ridge National Laboratory has been in regular use for more than two decades. Safe operation of this fast reactor over this extended period indicates that (1) fundamental design, (2) operational procedures, (3) operator training and performance, (4) maintenance activites, and (5) management have all been eminently satisfactory. The reactor and its uses are described, the operational history and significant events are reviewed, and operational improvements and maintenance are discussed

  8. A completely automatic operation type super-safe fast reactor, RAPID. Its application to dispersion source on lunar and earth surfaces

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Kawasaki, Akira; Iwamura, Takamichi

    2002-01-01

    At a viewpoint of flexible measures to future electric power demands, expectation onto a small-scale reactor for dispersion source is increasing gradually. This is thought to increase its importance not only for a source at proximity of its market in advanced nations but also for the one in developing nations. A study on development of the completely automatic operation type super-safe fast reactor, RAPID (refueling by all pins integrated design) has been carried out as a part of the nuclear energy basic research promoting system under three years project since 1999 by a trust of the Japan Atomic Energy Research Institute to a group of the Central Research Institute of Electric Power Industry (CRIEPI) and so on. As the reactor is a lithium cooled fast reactor with 200 Kw of electric output supposing to use at lunar surface, it can be applied to a super-small scale nuclear reactor on the earth, and has feasibility to become a new option of future nuclear power generation. On the other hand, CRIEPI has investigated on various types of fast reactors (RAPID series) for fast reactor for dispersion source on the earth. Here was introduced on such super-safe fast reactors at a center of RAPID-L. (G.K.)

  9. Inherently safe high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yamada, Masao; Hayakawa, Hitoshi

    1987-01-01

    It is recognized in general that High Temperature Gas-cooled Reactors have remarkable characteristics in inherent safety and it is well known that credits of the time margin have been admitted for accident evaluation in the licensing of the currently operating prototype HTGRs (300 MWe class). Recently, more inherently safe HTGRs are being developed in various countries and drawing attention on their possibility for urban siting. The inherent safety characteristics of these HTRs differ each other depending on their design philosophy and on the features of the components/structures which constitute the plant. At first, the specific features/characteristics of the elemental components/structures of the HTRs are explained one by one and then the overall safety features/characteristics of these HTR plants are explained in connection with their design philosophy and combination of the elemental features. Taking the KWU/Interatom Modular Reactor System as an example, the particular design philosophy and safety characteristics of the inherently safe HTR are explained with a result of preliminary evaluation on the possibility of siting close to densely populated area. (author)

  10. Safe design and operation of fluidized-bed reactors: Choice between reactor models

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1990-01-01

    For three different catalytic fluidized bed reactor models, two models presented by Werther and a model presented by van Deemter, the region of safe and unique operation for a chosen reaction system was investigated. Three reaction systems were used: the oxidation of benzene to maleic anhydride, the

  11. Research reactor developments in Australia

    International Nuclear Information System (INIS)

    Godfrey, Robert

    1998-01-01

    The Australian Nuclear Science and Technology Organization (ANSTO) operates the 10 MW research reactor, HIFAR, at the Lucas Heights site approximately 30 kilometres south of Sydney. Although recent reviews and inspections have confirmed that HIFAR operates safely by an adequate margin and has minimal impact, it was concluded that the reactor design and age places limitations on its operation and utilization, and that HIFAR is approaching the end of its economic life. In September 1997, a decision was made by the Australian Government to found ANSTO for the construction of a replacement research on the existing Lucas Heights site, subject to the requisite environmental assessment process. A draft EIS has been prepared and is currently undergoing public review. A design specification is in preparation, and a research reactor vendor pre-qualification process has been initiated. Spent fuel shipments have been made to Dounreay and to the Savannah River Site, and discussions are continuing regarding the disposition of the existing spent fuel and that arising form HIFAR's remaining operation. (author)

  12. The inherently-safe power reactor DYONISOS (Dynamic Nuclear Inherently-Safe Reactor Operating with Spheres)

    International Nuclear Information System (INIS)

    Taube, M.; Lanfranchi, M.; Weissenfluh, Th. von; Ligou, J.; Yadigaroglu, G.; Taube, P.

    1986-01-01

    A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydro-dynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h. (author)

  13. Research reactor decommissioning experience - concrete removal and disposal -

    International Nuclear Information System (INIS)

    Manning, Mark R.; Gardner, Frederick W.

    1990-01-01

    Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limits for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations

  14. Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

  15. Australia's new high performance research reactor

    International Nuclear Information System (INIS)

    Miller, R.; Abbate, P.M.

    2003-01-01

    A contract for the design and construction of the Replacement Research Reactor was signed in July 2000 between ANSTO and INVAP from Argentina. Since then the detailed design has been completed, a construction authorization has been obtained, and construction has commenced. The reactor design embodies modern safety thinking together with innovative solutions to ensure a highly safe and reliable plant. Also significant effort has been placed on providing the facility with diverse and ample facilities to maximize its use for irradiating material for radioisotope production as well as providing high neutron fluxes for neutron beam research. The project management organization and planing is commensurate with the complexity of the project and the number of players involved. (author)

  16. Technical specifications: Health Physics Research Reactor

    International Nuclear Information System (INIS)

    1979-02-01

    The technical specifications define the key limitations that must be observed for safe operation of the Health Physics Research Reactor (HPRR) and an envelope of operation within which there is assurance that these limits will not be exceeded. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

  17. Progress on PRISM, an inherently safe, economic, and testable advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Tippets, F.E.; Salerno, L.N.; Boardman, C.E.; Kwant, W.; Murata, R.E.; Snyder, C.R.

    1987-01-01

    This paper reports progress on the design of PRISM (Power Reactor Inherently Safe Module) under the DOE-sponsored innovative reactor program now in its third year at General Electric. The purpose of this program is to develop a design for an inherently safe, reliable, and marketable liquid metal fast reactor power plant. The PRISM design approach includes the following key elements: Compact sodium-cooled pool-type reactor modules that are sized to enable factory fabrication, economical shipment to inland as well as water-side sites, and economical full-scale prototype testing for design certification; Nuclear safety-related envelope limited to the reactor modules and their service systems; Inherent, passive shutdown heat removal for loss-of-cooling events; Inherent, passive reactivity shutdown for failure-to-scram events

  18. Status of the RA research reactor decommissioning project

    International Nuclear Information System (INIS)

    Ljubenov, V.; Nikolic, D.; Pesic, M.; Milosevic, M.; Kostic, Lj.; Steljic, M.; Sotic, O.; Antic, D. . E-mail address of corresponding author: vladan@vin.bg.ac.yu; Ljubenov, V.)

    2005-01-01

    The 6.5 MW heavy water RA research reactor at the VINCA Institute of Nuclear Sciences operated from 1959 to 1984. After 18 years of extended shutdown in 2002 it was decided that the reactor shutdown should be final. Preliminary decommissioning activities have been initiated by the end of 2002 under the Technical Co-operation Programme of the International Atomic Energy Agency. The objective of the project is to implement safe, timely and cost-effective decommissioning of the RA reactor up to unrestricted use of the site. Decommissioning project is closely related to two other projects: Safe Removal of the RA Reactor Spent Nuclear Fuel and Radioactive Waste Management in VINCA Institute. The main phases of the project include preparation of the detailed decommissioning plan, radiological characterization of the reactor site, dismantling and removal of the reactor components and structures, decontamination, final radiological site survey and the documentation of all the activities in order to obtain the approval for unrestricted use of the facility site. In this paper a review of the activities related to the preparation and realization of the RA reactor decommissioning project is given. Status of the project's organizational and technical aspects as for July 2004 are presented and plans for the forthcoming phases of the project realization are outlined. (author)

  19. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  20. Supervisory and managerial aspects on the safe operation of the Egyptian second research reactor

    International Nuclear Information System (INIS)

    Abdelrazek, I.D.; Shokr, A.M.

    2000-01-01

    ETRR-2 is a multipurpose reactor for radioisotopes production, neutron beam experiments, basic and applied research in physics and engineering, fuels and material tests, and for training. The reactor is an open pool type, 22 MW power, with average thermal neutron flux of 1.4x10 14 n/cm 2 sec, cooled and moderated by light water and with beryllium reflectors. Various experimental devices and irradiation facilities are integrated with the reactor. The reactor has been licensed for operation in November 1998. Several principles and regulations have been applied to all the reactor project stages to achieve safety. Moreover, other several principals, regulations, and aspects are enforced by the AEA, National Center for Nuclear Safety and Radiation Control, NCNSRC, and reactor management to achieve safety during reactor operation and utilization. Responsibility on Safety and Supervision Aspects AEA chairman has the ultimate responsibility on the reactor safety during operation and utilization. The primary responsible on the safety is the ETRR-2 supervisor, who is supervising the ETRR-2 site that includes the ETRR-2 reactor, Fuel Manufacturing Plant, and other two projects (under execution): Radioisotopes Production Plant and Dry Fuel Storage Facility. ETRR-2 supervisor is responsible to ensure that: the reactor is operated in accordance with the safety requirements by qualified and trained personnel, updating and enforcement of the reactor mandatory documentation, and the services are adequate for the reactor operation. He is responsible, also, to guarantee that: the reactor manager has enough authority and resources to carry out his function effectively and the reactor is kept operated in agreement with established procedures. A Technical Revision Committee, TRC, is formed to advice the ETRR-2 supervisor on the safety of the reactor and experiments. The committee members are from AEA experts with no direct relation to the reactor and experiments being performed. Members

  1. French experience in research reactor fuel transportation

    International Nuclear Information System (INIS)

    Raisonnier, Daniele

    1996-01-01

    Since 1963 Transnucleaire has safely performed a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied suitable packaging for all types of nuclear fuel cycle radioactive material from front-end and back-end products and for power or for research reactors. Transportation of spent fuel from power reactors are made on a regular and industrial basis, but this is not yet the case for the transport of spent fuel coming from research reactors. Each shipment is a permanent challenge and requires a reactive organization dealing with all the transportation issues. This presentation will explain the choices made by Transnucleaire and its associates to provide and optimize the corresponding services while remaining in full compliance with the applicable regulations and customer requirements. (author)

  2. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

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

  3. Activities of research-reactor-technology project in FNCA from FY2005 to FY2007. Sharing neutronics calculation technique for core management and utilization of research reactors

    International Nuclear Information System (INIS)

    2010-07-01

    RRT project (Research-Reactor-Technology Project) was carried out with the theme of 'sharing neutronics calculation technique for core management and utilization of research reactors' in the framework of FNCA (Forum for Nuclear Cooperation in Asia) from FY2005 to FY2007. The objective of the project was to improve and equalize the level of neutronics calculation technique for the reactor core management among participating countries to assure the safe and stable operation of research reactors and the promotion of the effective utilization. Neutronics calculation codes, namely SRAC code system and MVP code, were adopted as common codes. Participating countries succeeded in applying the common codes to analyzing the core of each domestic research reactor. Some participating countries succeeded in applying the common codes to analyzing for utilization of own research reactors. Activities of RRT project have improved and equalized the level of neutronics calculation technique among participating countries. (author)

  4. Application of Code Of Conduct on the Safety of Research Reactor (RTP)

    International Nuclear Information System (INIS)

    Ligam, A.S.; Ahmad Nabil Abd Rahim; Zarina Masood

    2014-01-01

    The implementation and the practices of the effective safety system at research reactors are important to ensure that the worker, public and environment do not receive any abnormal causes. Many international safety related support agencies for research reactor such as International Atomic Energy Agency (IAEA) providing guidelines that can be applied to enhance and strengthen the enforcement of safety namely Code of Conduct on the Safety of Research Reactor (IAEA/CODEOC/RR/2006). The excellent safety management, reliability, and maintainability of RTP reactor structures, coupled with personnel numerous lessons and experiences learned, Reactor TRIGA PUSPATI research reactor providing Nuclear Malaysia personnel and visitor the very safe working and visiting environment. This paper will discuss the status, practices and improvement strategies over the past few years. (author)

  5. Technology which led to the westinghouse inherently safe liquid metal reactor

    International Nuclear Information System (INIS)

    Schmidt, J.E.; Coffield, R.D.; Doncals, R.A.; Kalinowski, J.E.; Markley, R.A.

    1985-01-01

    The Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor programs resulted in an understanding of liquid metal reactor behavior that is being used to design inherent safety capability into liquid metal reactors. Technological advances give the same beneficial operating characteristics of conventional liquid metal reactors, however, the addition of inherently safe design features precludes the initiation of hypothetical core disruptive accidents. These innovative features permit inherent safety capability to be demonstrated with more than adequate margins. Also, the variety of inherent safety features provides the designers with options in selecting inherent design features for a specific reactor application

  6. Detection of fission products release in the research reactor 'RA' spent fuel storage pool

    International Nuclear Information System (INIS)

    Matausek, M.V.; Vukadin, Z.; Pavlovic, S.; Maksin, T.; Idakovic, Z.; Marinkovic, N.

    1997-05-01

    Spent fuel resulting from 25 years of operating the 6.5/10 MW thermal heavy water moderated and cooled research reactor RA at the VINCA Institute is presently all stored in the temporary spent fuel storage pool in the basement of the reactor building. In 1984, the reactor was shut down for refurbishment, which for a number of reasons has not yet been completed. Recent investigations show that independent of the future status of the research reactor, safe disposal of the so far irradiated fuel must be the subject of primary concern. The present status of the research reactor RA spent fuel storage pool at the VINCA Institute presents a serious safety problem. Action is therefore initiated in two directions. First, safety of the existing spent fuel storage should be improved. Second, transferring spent fuel into another, presumably dry storage space should be considered. By storing the previously irradiated fuel of the research reactor RA in a newly built storage space, sufficient free space will be provided in the existing spent fuel storage pool for the newly irradiated fuel when the reactor starts operation again. In the case that it would be decided to decommission the research reactor RA, the newly built storage space would provide safe disposal for the fuel irradiated so far

  7. 105-C Reactor interim safe storage project technology integration plan

    International Nuclear Information System (INIS)

    Pulsford, S.K.

    1997-01-01

    The 105-C Reactor Interim Safe Storage Project Technology Integration Plan involves the decontamination, dismantlement, and interim safe storage of a surplus production reactor. A major goal is to identify and demonstrate new and innovative D and D technologies that will reduce costs, shorten schedules, enhance safety, and have the potential for general use across the RL complex. Innovative technologies are to be demonstrated in the following areas: Characterization; Decontamination; Waste Disposition; Dismantlement, Segmentation, and Demolition; Facility Stabilization; and Health and Safety. The evaluation and ranking of innovative technologies has been completed. Demonstrations will be selected from the ranked technologies according to priority. The contractor team members will review and evaluate the demonstration performances and make final recommendations to DOE

  8. Project Experiences in Research Reactor Ageing Management, Modernization and Refurbishment. Report of a Technical Meeting on Research Reactor Ageing Management, Modernization and Refurbishment

    International Nuclear Information System (INIS)

    2014-08-01

    Research reactors have played an important role in several scientific fields for around 60 years: in the development of nuclear science and technology; in the valuable generation of radioisotopes for various applications; and in the development of human resources and skills. Moreover, research reactors have been effectively utilized to support sustainable development in more than 60 countries worldwide. More than half of all operating research reactors are now over 40 years old, with many exceeding their originally conceived design life. The majority of operating research reactors face challenges due to the negative impacts of component and system ageing, which manifest in a number of forms. This situation was highlighted by a serious medical isotope supply crisis which peaked in mid-2010, when several major producing reactors underwent prolonged shutdowns due to extensive necessary overhauls of various systems. Several facilities have established a proactive systematic approach to managing ageing or mitigating its impact on safety and availability of isotopes. Others have tried to prevent or remedy the drawbacks of ageing on a case by case basis. Overall, a large body of knowledge related to ageing issues exists in many Member States. Collecting and sharing this information within the research reactor community can provide a solid foundation to develop a more systematic approach — that is, an ageing management programme to prevent negative consequences of ageing on the safety, and the operability and lifetime of operating, or even future, reactors. It may also help organizations to manage research reactors that have been in an extended shutdown state by ensuring that any required systems are operated and maintained in a safe manner prior to final decommissioning and disposal of fuel to safe storage facilities. Sharing experiences from projects undertaken to refurbish or replace equipment and systems, satisfy safety and regulatory requirements, improve

  9. Project Experiences in Research Reactor Ageing Management, Modernization and Refurbishment. Report of a Technical Meeting on Research Reactor Ageing Management, Modernization and Refurbishment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-08-15

    Research reactors have played an important role in several scientific fields for around 60 years: in the development of nuclear science and technology; in the valuable generation of radioisotopes for various applications; and in the development of human resources and skills. Moreover, research reactors have been effectively utilized to support sustainable development in more than 60 countries worldwide. More than half of all operating research reactors are now over 40 years old, with many exceeding their originally conceived design life. The majority of operating research reactors face challenges due to the negative impacts of component and system ageing, which manifest in a number of forms. This situation was highlighted by a serious medical isotope supply crisis which peaked in mid-2010, when several major producing reactors underwent prolonged shutdowns due to extensive necessary overhauls of various systems. Several facilities have established a proactive systematic approach to managing ageing or mitigating its impact on safety and availability of isotopes. Others have tried to prevent or remedy the drawbacks of ageing on a case by case basis. Overall, a large body of knowledge related to ageing issues exists in many Member States. Collecting and sharing this information within the research reactor community can provide a solid foundation to develop a more systematic approach — that is, an ageing management programme to prevent negative consequences of ageing on the safety, and the operability and lifetime of operating, or even future, reactors. It may also help organizations to manage research reactors that have been in an extended shutdown state by ensuring that any required systems are operated and maintained in a safe manner prior to final decommissioning and disposal of fuel to safe storage facilities. Sharing experiences from projects undertaken to refurbish or replace equipment and systems, satisfy safety and regulatory requirements, improve

  10. Surveillance and Maintenance Plan for the 105-C Reactor Safe Storage Enclosure

    International Nuclear Information System (INIS)

    Logan, T. E.

    1998-01-01

    This document provides a plan for implementing surveillance and maintenance activities to ensure that the 105-C Reactor Safe Storage Enclosure is maintained in a safe, environmentally secure, and cost-effective manner until subsequent closure during the final disposition phase of decommissioning

  11. Submersion-Subcritical Safe Space (S4) reactor

    International Nuclear Information System (INIS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

  12. New fuels for proliferation-safe research reactors. SCK-CEN a kingpin in international research

    International Nuclear Information System (INIS)

    Van den Berghe, S.

    2011-01-01

    A number of research reactors in the world, including BR2, are still working on highly enriched proliferation sensitive nuclear fuel. Extremely high-density fuel is required to make possible the switch to low-enriched fuel without loss of effectiveness. In the past few years, SCK-CEN has developed into the global test station for new, low-enriched fuels.

  13. The ultimate safe (US) Reactor: A concept for the third millenium

    International Nuclear Information System (INIS)

    Gat, U.

    1986-01-01

    The Ultimate Safe (U.S.) Reactor is based on a novel safety concept. Fission products in the reactor are allowed to accumulate only to a level at which they would constitute a harmless source term. Removal of fission products also removes the decay heat - the driving force for the source term. The reactor has no excess criticality and is controlled by the reactivity temperature coefficient. Safety is inherent and passive. Waste is removed from the site promptly

  14. Power Trip Set-points of Reactor Protection System for New Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Yang, Soohyung

    2013-01-01

    This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

  15. Conceptual design of inherently safe integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. I.; Chang, M. H.; Lee, D. J. and others

    1999-03-01

    The design concept of a 300 MWt inherently safe integral reactor(ISIR) for the propulsion of extra large and superhigh speed container ship was developed in this report. The scope and contents of this report are as follows : 1. The state of the art of the technology for ship-mounted reactor 2. Design requirements for ISIR 3. Fuel and core design 4. Conceptual design of fluid system 5. Conceptual design of reactor vessel assembly and primary components 6. Performance analyses and safety analyses. Installation of two ISIRs with total thermal power of 600MWt and efficiency of 21% is capable of generating shaft power of 126,000kW which is sufficient to power a container ship of 8,000TEU with 30knot cruise speed. Larger and speedier ship can be considered by installing 4 ISIRs. Even though the ISIR was developed for ship propulsion, it can be used also for a multi-purpose nuclear power plant for electricity generation, local heating, or seawater desalination by mounting on a movable floating barge. (author)

  16. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  17. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  18. TRIGA 14 MW Research Reactor Status and Utilization

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  19. MAPLE research reactor safety uncertainty assessment methodology

    International Nuclear Information System (INIS)

    Sills, H.E.; Duffey, R.B.; Andres, T.H.

    1999-01-01

    The MAPLE (multipurpose Applied Physics Lattice Experiment) reactor is a low pressure, low temperature, open-tank-in pool type research reactor that operates at a power level of 5 to 35 MW. MAPLE is designed for ease of operation, maintenance, and to meet today's most demanding requirements for safety and licensing. The emphasis is on the use of passive safety systems and environmentally qualified components. Key safety features include two independent and diverse shutdown systems, two parallel and independent cooling loops, fail safe operation, and a building design that incorporates the concepts of primary containment supported by secondary confinement

  20. Possibility evaluation of eliminating the saturated control fuel element from Tehran research reactor core

    International Nuclear Information System (INIS)

    Mirvakili, S.M.; Keyvani, M.; Arshi, S. Safaei; Khalafi, H.

    2012-01-01

    Highlights: ► We show safe operation of Tehran research reactor without one of its control rods. ► We propose an optimum new core configuration by fuel management calculations. ► We calculate neutronic and thermal hydraulic parameters of the new core. ► Parameters are consistent with the safety criteria. - Abstract: In this study the possibility of safe operation of Tehran research reactor (TRR) providing the elimination of one control rod is evaluated. One of the control fuel elements (CFEs) of TRR has been reached the maximum permissible burn-up and due to the impossibility of fresh fuel assembly provision under current situation, providing an optimum core configuration which satisfies safe operation conditions by applying fuel management calculations is essential. In order to ensure the safe and stable operation of recently proposed configuration for TRR core, neutronic and thermal hydraulic parameters of the new core are calculated and compared with the safety criteria. The results show good compatibility with reactor safety criteria, and provide desired shutdown margin and safety reactivity factor.

  1. A conceptual design of intrinsically safe and economical reactor (ISER)

    International Nuclear Information System (INIS)

    Oda, Junro

    1985-01-01

    The purpose of this paper is to describe the reference conceptual designs of the ISER which were prepared for the ISER development forum in Japan. At the forum, participants from influential utilities, academia, as well as companies in the nuclear industry, discussed the development of the inherently safe reactor over the last two years. The conceptual designs described in this paper are preliminary trial designs at an early stage and essentially versions of the PIUS reactor developed by ASEA-ATOM. A notable feature of the ISER which is different from the original PIUS is its use of a steel reactor pressure vessel for reducing plant construction costs and improving plant performance

  2. Myrrha, new polyvalent research reactor

    International Nuclear Information System (INIS)

    Khatcheressian, Nayiri; Haj Tahar, Malek

    2015-01-01

    Myrrha (Multi-purpose hybrid research reactor for high-tech applications) is the first prototype of sub-critical nuclear reactor driven by a particle accelerator (an ADS, accelerator-driven system) at semi-industrial scale (50-100 MW), a safe and easy-to-control technology. In an interview, the manager of this project recalls his curriculum, presents and comments the characteristics of Myrrha, outlines why these ADS are so interesting to produce radio-isotopes, comments the variety of countries and companies involved in this project, outlines the peculiarities of Myrrha in terms of safety and the main technological challenges (a mixing of lead and bismuth for the coolant, control of corrosion by oxygen, an improved reliability based on redundant design and fault tolerance, MOX as fuel). He also evokes competing technologies

  3. Best Safety Practices for the Operation of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Boeck, H.; Villa, M. [Atominstitute of the Austrian Universities, 1020 Vienna (Austria)

    2002-07-01

    A survey on administrative, organisational and technical aspects for the safe and efficient operation of a 250 kW TRIGA Mark II research reactor is given. The replacement of the I and C system is discussed, maintenance procedures are presented and the fuel management is described. (author)

  4. Best Safety Practices for the Operation of Research Reactors

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2002-01-01

    A survey on administrative, organisational and technical aspects for the safe and efficient operation of a 250 kW TRIGA Mark II research reactor is given. The replacement of the I and C system is discussed, maintenance procedures are presented and the fuel management is described. (author)

  5. Summary of advanced LMR [Liquid Metal Reactor] evaluations: PRISM [Power Reactor Inherently Safe Module] and SAFR [Sodium Advanced Fast Reactor

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G.

    1989-10-01

    In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) [Berglund, 1987] and the Sodium Advanced Fast Reactor (SAFR) [Baumeister, 1987], were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the ''inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II [NED, 1986]. The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs

  6. Dynamic operator actions analysis for inherently safe fast reactors and light water reactors

    International Nuclear Information System (INIS)

    Ho, V.; Apostolakis, G.

    1988-01-01

    A comparative dynamic human actions analysis of inherently safe fast reactors (ISFRs) and light water reactors (LWRs) in terms of systems response and estimated human error rates is presented. Brief overviews of the ISFR and LWR systems are given to illustrate the design differences. Key operator actions required by the ISFR reactor shutdown and decay heat removal systems are identified and are compared with those of the LWR. It is observed that, because of the passive nature of the ISFR safety-related systems, a large time window is available for operator actions during transient events. Furthermore, these actions are fewer in number, are less complex, and have lower error rates and less severe consequences than those of the LWRs. We expect the ISFR operator errors' contribution to risk is smaller (at least in the context of the existing human reliability models) than that of the LWRs. (author)

  7. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  8. RA research reactor - properties and experimental capabilities

    International Nuclear Information System (INIS)

    Milosevic, M.; Martinc, R.

    1978-01-01

    The brief survey of the Reactor RA exploitation experience, as well as the reactor equipment state, after 18 years of operation is presented. The results of efforts spent on reactor characteristics improvement in order to ensure safe and reliable reactor operation for next 15-20 years, are described [sr

  9. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  10. Status and future of the WWR-M research reactor in Kiev

    Energy Technology Data Exchange (ETDEWEB)

    Bazavov, D.A.; Gavrilyuk, V.I.; Kirischuk, V.I.; Kochetkov, V.V.; Lysenko, M.V.; Makarovskiy, V.N.; Scherbachenko, A.M.; Shevel, V.N.; Slisenko, V.I. [Institute for Nuclear Research, Kiev (Ukraine)

    2001-07-01

    Kiev WWR-M Research Reactor, operated at maximum power of 10 MW, was put into operation in 1960 and during its 40-years history has been used to perform numerous studies in different areas of science and technology. Due to a number of technical problems the Research Reactor, the only one in Ukraine, was shut down in 1993 and then put into operation in 1999 again. Now there is an intention to reconstruct Kiev Research Reactor. The upgraded Research Reactor would allow solving such problems as the safe operation of Ukrainian NPPs, radioisotope production and, naturally, fundamental and applied research. The main problem for the successful operation of Kiev Research Reactor is the management and storage of spent fuel at the site, since after core unloading the spent fuel storage appears to be practically completed. So it is absolutely necessary to ship the most part of the spent fuel for reprocessing and as soon as possible. Besides, there is a need to build up the new spent fuel storage, because the tank of available storage requires careful inspection for corrosion. (author)

  11. FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

  12. On the classification of structures, systems and components of nuclear research and test reactors

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    2009-01-01

    The classification of structures, systems and components of nuclear reactors is a relevant issue related to their design because it is directly associated with their safety functions. There is an important statement regarding quality standards and records that says Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The definition of the codes, standards and technical requirements applied to the nuclear reactor design, fabrication, inspection and tests may be seen as the main result from this statement. There are well established guides to classify structures, systems and components for nuclear power reactors such as the Pressurized Water Reactors but one can not say the same for nuclear research and test reactors. The nuclear reactors safety functions are those required to the safe reactor operation, the safe reactor shutdown and continued safe conditions, the response to anticipated transients, the response to potential accidents and the control of radioactive material. So, it is proposed in this paper an approach to develop the classification of structures, systems and components of these reactors based on their intended safety functions in order to define the applicable set of codes, standards and technical requirements. (author)

  13. Nuclear safety requirements for operation licensing of Egyptian research reactors

    International Nuclear Information System (INIS)

    Ahmed, E.E.M.; Rahman, F.A.

    2000-01-01

    From the view of responsibility for health and nuclear safety, this work creates a framework for the application of nuclear regulatory rules to ensure safe operation for the sake of obtaining or maintaining operation licensing for nuclear research reactors. It has been performed according to the recommendations of the IAEA for research reactor safety regulations which clearly states that the scope of the application should include all research reactors being designed, constructed, commissioned, operated, modified or decommissioned. From that concept, the present work establishes a model structure and a computer logic program for a regulatory licensing system (RLS code). It applies both the regulatory inspection and enforcement regulatory rules on the different licensing process stages. The present established RLS code is then applied to the Egyptian Research Reactors, namely; the first ET-RR-1, which was constructed and still operating since 1961, and the second MPR research reactor (ET-RR-2) which is now in the preliminary operation stage. The results showed that for the ET-RR-1 reactor, all operational activities, including maintenance, in-service inspection, renewal, modification and experiments should meet the appropriate regulatory compliance action program. Also, the results showed that for the new MPR research reactor (ET-RR-2), all commissioning and operational stages should also meet the regulatory inspection and enforcement action program of the operational licensing safety requirements. (author)

  14. New developments in transportation for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mondanel, J.L. [Transnucleaire, F-75008 Paris (France)

    1998-07-01

    For more than 30 years, Transnucleaire has been performing safely a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied numerous packagings for all types of nuclear fuel cycle radioactive materials: for front-end and back-end products and for power and research reactors. Since the last meeting held in Bruges, Transnucleaire has been continuously involved in transportation activities for fresh and irradiated materials for research reactors. We are pleased to take the opportunity in this meeting to share with reactor operators, official bodies and other partners, the on-going developments in transportation and associated services. Special attention will be paid to the starting of transports of MTR spent fuel elements to the La Hague reprocessing plant where COGEMA offers reprocessing services on a long-term basis to reactors operators. Detailed information is provided on regulatory issues, which may affect transport activities: evolution of the regulations, real experiences of recent transportation and development of new packaging designs. Options and solutions will be proposed by Transnucleaire to improve the situation for continuation of national and international transports at an acceptable price whilst maintaining an ultimate level of safety (author)

  15. New safe reactor, but not first of a kind engineering

    International Nuclear Information System (INIS)

    Nurdin, M.

    1997-01-01

    New safe reactor but not a first of a kind engineering is a new reactor concept to fulfill the need on Small Reactor for power generation, both for electricity and for co-generation. Nuclear reactor system of this concept in certain degree has similar design compared to the established and successful reactor systems now in operation; so the material used for the same function and purpose is not the same. The strategy or choice adopted in achieving this concept will be automatically shown by the inspiration or philosophy of ''not to re-invent the wheel.'' Based on the above mentioned strategy, a certain degree of experimental verification and justification are of course needed/necessary to know better the deviations and the differences from the existing nuclear reactor concepts and further to anticipate of course precisely engineering behaviour of the proposed concept. Physical and engineering discussion on the proposed concept are main objectives of this paper in which most of the scope and objectives of this IAEA TCM on Small Reactors with minimized Staffing and/or Remote Monitoring are elaborated. They are discussed in such a way to give the technical and economical background of the proposed concept. (author)

  16. Seismic upgrading of the Brookhaven High Flux Beam Research Reactor

    International Nuclear Information System (INIS)

    Subudhi, M.

    1985-01-01

    In recent years the High Flux Beam Research (HFBR) reactor facility at Brookhaven National Laboratory (BNL) was upgraded from 40 to 50 MW power level. The reactor plant was built in the early sixties to the seismic design requirements of the period, using the static load approach. While the plant power level was upgraded, the seismic design was also improved according to current design criteria. This included the development of new floor response spectra for the facility and an overall seismic analysis of those systems important to the safe shutdown of the reactor. Items included in the reanalysis are the containment building with its internal structure, the piping systems, tanks, equipment, and heat exchangers. This paper describes the procedure utilized in developing the floor response spectra for the existing facility. Also included in the paper are the findings and recommendations, based on the seismic analysis, regarding the seismic adequacy of structural and mechanical systems vital to achieving the safe shutdown of the reactor. 11 references, 4 figures, 1 table

  17. Design and research of fuel element for pulsed reactor

    International Nuclear Information System (INIS)

    Tian Sheng

    1994-05-01

    The fuel element is the key component for pulsed reactor and its design is one of kernel techniques for pulsed reactor. Following the GA Company of US the NPIC (Nuclear Power Institute of China) has mastered this technique. Up to now, the first pulsed reactor in China (PRC-1) has been safely operated for about 3 years. The design and research of fuel element undertaken by NPIC is summarized. The verification and evaluation of this design has been carried out by using the results of measured parameters during operation and test of PRC-1 as well as comparing the design parameters published by others

  18. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    1975-12-01

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  19. Diagnosis of electric equipment at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Truong Sinh

    1999-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type of its kind in the world: Soviet-designed core and control system harmoniously integrated into the left-over infrastructure of the former American-made TRIGA MARK II reactor, which includes the reactor tank and shielding, graphite reflector, beam tubes and thermal column. The reactor is mainly used for radioisotope and radiopharmaceutical production, elemental analysis using neutron activation techniques, neutron beam exploitation, silicon doping, and reactor physics experimentation. For safe operation of the reactor maintenance work has been carried out for the reactor control and instrumentation, reactor cooling, ventilation, radiomonitoring, mechanical, normal electric supply systems as well as emergency electric diesel generators and the water treatment station. Technical management of the reactor includes periodical maintenance as required by technical specifications, training, re-training and control of knowledge for reactor staff. During recent years, periodic preventive maintenance (PPM) has been carried out for the electric machines of the technological systems. (author)

  20. Safe operation of critical assemblies and research reactors. Code of practice and Technical appendix. 1971 ed

    International Nuclear Information System (INIS)

    Cox, J.

    1971-01-01

    This book is in two parts. The first is a Code of Practice for the Safe Operation of Critical Assemblies and Research Reactors, prepared as a result of a meeting of experts which took place in Vienna on 20-24 May 1968. The Code has been prepared by the International Atomic Energy Agency in co-operation with the World Health Organization, and its publication is sponsored by both organizations. In addition, the Code was approved by the Board of Governors of the International Atomic Energy Agency on 16 December 1968 as part of the Agency's safety standards, which are applied to operations undertaken by Member States with the assistance of the Agency. The Board, in approving the publication of the present book, also recommended Member States to take the Code into account in the formulation of national regulations and recommendations. The second part of the book is a Technical Appendix to give information and illustrative samples that would be helpful in implementing the Code of Practice. This second part, although published under the same cover, is not part of the Code. An extensive Bibliography, amplifying the Technical Appendix, is included at the end.

  1. Research nuclear reactor operation management

    International Nuclear Information System (INIS)

    Preda, M.; Carabulea, A.

    2008-01-01

    Some aspects of reactor operation management are highlighted. The main mission of the operational staff at a testing reactor is to operate it safely and efficiently, to ensure proper conditions for different research programs implying the use of the reactor. For reaching this aim, there were settled down operating plans for every objective, and procedure and working instructions for staff training were established, both for the start-up and for the safe operation of the reactor. Damages during operation or special situations which can arise, at stop, start-up, maintenance procedures were thoroughly considered. While the technical skill is considered to be the most important quality of the staff, the organising capacity is a must in the operation of any nuclear facility. Staff training aims at gaining both theoretical and practical experience based on standards about staff quality at each work level. 'Plow' sheet has to be carefully done, setting clear the decision responsibility for each person so that everyone's own technical level to be coupled to the problems which implies his responsibility. Possible events which may arise in operation, e.g., criticality, irradiation, contamination, and which do not arise in other fields, have to be carefully studied. One stresses that the management based on technical and scientific arguments have to cover through technical, economical and nuclear safety requirements a series of interlinked subprograms. Every such subprograms is subject to some peculiar demands by the help of which the entire activity field is coordinated. Hence for any subprogram there are established the objectives to be achieved, the applicable regulations, well-defined responsibilities, training of the personnel involved, the material and documentation basis required and activity planning. The following up of positive or negative responses generated by experiments and the information synthesis close the management scope. Important management aspects

  2. Physical Characteristics of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Ngo Quang Huy

    1994-10-01

    The operation of the TRIGA MARK II reactor of nominal power 250 KW has been stopped as all the fuel elements have been dismounted and taken away in 1968. The reconstruction of the reactor was accomplished with Russian technological assistance after 1975. The nominal power of the reconstructed reactor is of 500 KW. The recent Dalat reactor is unique of its kind in the world: Russian-designed core combined with left-over infrastructure of the American-made TRIGA II. The reactor was loaded in November 1983. It has reached physical criticality on 1/11/1983 (without central neutron trap) and on 18/12/1983 (with central neutron trap). The power start up occurred in February 1984 and from 20/3/1984 the reactor began to be operated at the nominal power 500 KW. The selected reports included in the proceedings reflect the start up procedures and numerous results obtained in the Dalat Nuclear Research Institute and the Centre of Nuclear Techniques on the determination of different physical characteristics of the reactor. These characteristics are of the first importance for the safe operation of the Dalat reactor

  3. Technological aspects of intrinsically safe and economical reactor (ISER)

    International Nuclear Information System (INIS)

    Yamada, Nobuyuki; Oda, Junro; Yamanaka, Kazuo; Sugawara, Ichiro.

    1987-01-01

    ISER is a modified version of process inherent ultimate safe reactor (PIUS) developed by ASEA-ATOM, Sweden, and follows the same inherent safety principle, that is, passive reactor shutdown through the introduction of borated pool water into a core via an interface, and passive decay heat removal by natural circulation. The most significant deviation from the PIUS is that the ISER employs a steel reactor pressure vessel enclosed in the reactor pit, instead of a prestressed concrete reactor pressure vessel of the PIUS. The merits of using steel pressure vessels are siting versatility including barge-mounted plants, low cost, the standardization and serial production of total NSSSs through the weight reduction and compaction of primary system, as well as the possibility of utilizing current LWR technology, which minimizes R and D effort. In this paper, the design features of the latest version of ISERs are shown, and the specific problems of the key components are discussed. The primary system consists of a primary coolant loop and a borated water pool, which are connected with upper and lower interfaces. The nuclear design and thermohydraulic design, the operation and maintenance, and the design features of a steam generator, a pressurizer, interfaces and so on are described. (Kako, I.)

  4. Project management plan for the 105-C Reactor interim safe storage project. Revision 1

    International Nuclear Information System (INIS)

    Miller, R.L.

    1997-01-01

    In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

  5. Sodium heat pipe module test for the SAFE-30 reactor prototype

    International Nuclear Information System (INIS)

    Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

    2001-01-01

    Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

  6. Safety aspect of long-life small safe power reactors

    International Nuclear Information System (INIS)

    Zaki, S.; Sekimoto, H.

    1995-01-01

    Safety aspects of several design options of long-life small safe fast power reactors using nitride fuel and lead-bismuth as coolant are discussed. In the present study hypothetical accidents are simulated for these reactors, i.e., unprotected simultaneous ULOF (total loss of primary pumping system) and UTOP (rod run out transient over power) accidents, caused by the simultaneous withdrawal of all control rods. The proposed designs have some important safety characteristics as low reactivity swing (only 0.2-0.25$), and negative coolant void coefficient over whole burnup period. Effectively negative value of all components of reactivity during an accident is observed. The safety performances of the balance, pancake, and tall slender type of core, each of them satisfy reactivity and negative coolant void coefficient constraint, against the above accident are compared. The simulation results show that all of the design options can survive the above accidents without the help of reactor scram and without the need of operator actions. (author)

  7. History of Discharge of Radionuclide from Nuclear Malaysia Research Reactor

    International Nuclear Information System (INIS)

    Mohd Nahar Othman; Ismail Sulaiman

    2013-01-01

    After more than 40 years the operation of Malaysian Nuclear Agency research reactor, this is the first time, the discharge emission radionuclide to environment is recorded and analyzed in detail. Starting from 1984, radionuclide Ar-41 had been analyzed manually by Research Officers but their finding is not recorded in any Journal. That is responsible of Safety and Health Division to make sure the safety of the workers and public living around Malaysian Nuclear Agency receive safe dose. This paper will report dose that had been discharge to the environment starting from 1984 to 2012 and it detail calculations.After more than 40 years the operation of Malaysian Nuclear Agency research reactor, this is the first time, the discharge emission radionuclide to environment is recorded and analyzed in detail. Starting from 1984, radionuclide Ar-41 had been analyzed manually by Research Officers but their finding is not recorded in any Journal. That is responsible of Safety and Health Division to make sure the safety of the workers and public living around Malaysian Nuclear Agency receive safe dose. This paper will report dose that had been discharge to the environment starting from 1984 to 2012 and it detail calculations. (author)

  8. Trends in fusion reactor safety research

    International Nuclear Information System (INIS)

    Herring, J.S.; Holland, D.F.; Piet, S.J.

    1991-01-01

    Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g. for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions. 18 refs

  9. Proceedings of the 1998 workshop on the utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-10-01

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

  10. Past and present supply of enriched uranium for research reactors in the European Union

    International Nuclear Information System (INIS)

    Mueller, H.

    2002-01-01

    In the last decade research reactor operators have focused mainly on the issues of disposal of spent research reactor fuel and the development of high density fuels. The safe supply of fresh uranium did not receive as much attention. This is surprising since the United States - who was the main supplier for LEU and HEU since the late 1950's - stopped supplying non-US research reactors with enriched uranium a decade ago. The reason for this stop of supply is described in this paper. This paper explains how research reactors in the E U continued to operate during the last decade, in spite of the fact that their primary supply source had not provided LEU and HEU over the same period. (author)

  11. German concept and status of the disposal of spent fuel elements from German research reactors

    International Nuclear Information System (INIS)

    Komorowski, K.; Storch, S.; Thamm, G.

    1995-01-01

    Eight research reactors with a power ≥ 100 kW are currently being operated in the Federal Republic of Germany. These comprise three TRIGA-type reactors (power 100 kW to 250 kW), four swimming-pool reactors (power 1 MW to 10 MW) and one DIDO type reactor (power 23 MW). The German research reactors are used for neutron scattering for basic research in the field of solid state research, neutron metrology, for the fabrication of isotopes and for neutron activation analysis for medicine and biology, for investigating the influence of radiation on materials and for nuclear fuel behavior. It will be vital to continue current investigations in the future. Further operation of the German research reactors is therefore indispensable. Safe, regular disposal of the irradiated fuel elements arising now and in future operation is of primary importance. Furthermore, there are several plants with considerable quantities of spent fuel, the safe disposal of which is a matter of urgency. These include above all the VKTA facilities in Rossendorf and also the TRIGA reactors, where disposal will only be necessary upon decommissioning. The present paper report is concerned with the disposal of fuel from the German research reactors. It briefly deals with the situation in the USA since the end of 1988, describes interim solutions for current disposal requirements and then mainly concentrates on the German disposal concept currently being prepared. This concept initially envisages the long-term (25--50 years) dry interim storage of fuel elements in special containers in a central German interim store with subsequent direct final disposal without reprocessing of the irradiated fuel

  12. Reactor physics studies at the Zittau Training and research reactor ZLFR

    Energy Technology Data Exchange (ETDEWEB)

    Konschak, K.; Horche, W.; Honisch, H.; Berger, J. (Ingenieurhochschule Zittau (German Democratic Republic). Sektion Kraftwerksanlagenbau und Energieumwandlung); Doerschel, B. (Technische Univ., Dresden (German Democratic Republic). Sektion Physik)

    1982-04-01

    It is reported on experimental studies during the start-up period of the Zittau training and research reactor ZLFR. The critical mass obtained is in good agreement with the calculated value. It corresponds to a core charge of 90 fuel assemblies ECH-1. The shutdown reactivity of the safety rod and of the three control rods is 3.2% in total. The reactivity effects due to shuffling, internals, and configuration modifications as well as to intentional or unintentional changes in the operating conditions have been analyzed from the viewpoint of safe operation.

  13. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

    The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

  14. Innovative research reactor core designed. Estimation and analysis of gamma heating distribution

    International Nuclear Information System (INIS)

    Setiyanto

    2014-01-01

    The Gamma heating value is an important factor needed for safety analysis of each experiments that will be realized on research reactor core. Gamma heat is internal heat source occurs in each irradiation facilities or any material irradiated in reactor core. This value should be determined correctly because of the safety related problems. The gamma heating value is in general depend on. reactor core characteristics, different one and other, and then each new reactor design should be completed by gamma heating data. The Innovative Research Reactor is one of the new reactor design that should be completed with any safety data, including the gamma heating value. For this reasons, calculation and analysis of gamma heating in the hole of reactor core and irradiation facilities in reflector had been done by using of modified and validated Gamset computer code. The result shown that gamma heating value of 11.75 W/g is the highest value at the center of reactor core, higher than gamma heating value of RSG-GAS. However, placement of all irradiation facilities in reflector show that safety characteristics for irradiation facilities of innovative research reactor more better than RSG-GAS reactor. Regarding the results obtained, and based on placement of irradiation facilities in reflector, can be concluded that innovative research reactor more safe for any irradiation used. (author)

  15. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

    The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  16. Welcoming Address & Opening Remarks [International Conference on Research Reactors: Safe Management and Effective Utilization, Vienna (Austria), 16-20 November 2015

    International Nuclear Information System (INIS)

    Chudakov, M.

    2017-01-01

    For more than 60 years, research reactors have been centres of innovation and productivity for nuclear science and technology programmes in 67 countries around the world. Research reactors provide a multidisciplinary environment to catalyse scientific, industrial, medical and agricultural development. They are facilities for nuclear education and training of young scientists and technicians, and they can contribute to the development of nuclear power programmes. According to the IAEA Research Reactor Database, there are 246 research reactors currently in operation in 55 countries, and close to 30 new research reactor projects are at different stages of implementation. Many of the operating reactors are several decades old and face ageing management issues. These reactors must be operated and maintained with due regard to safety and security. Some reactors face challenges with sustainable supply of fresh fuel. Others are looking to improve utilization, which is linked to justifying adequate resources for operation, maintenance and refurbishment. As some of the fuel return programmes are expected to wind down in the near term, the community will need to find solutions for spent fuel and waste management. And taking into account the large number of reactors, about 140, no longer in operation, as well as ageing reactors coming to the end of their lifecycles, decommissioning is an important area of sharing experience and best practice. You will have an opportunity to discuss these and other issues over the course of the conference.

  17. Low-temperature thermionics in space nuclear power systems with the safe-type fast reactor

    International Nuclear Information System (INIS)

    Zrodnikov, A.V.; Yarygin, V.I.; Lazarenko, G.E.; Zabudko, A.N.; Ovcharenko, M.K.; Pyshko, A.P.; Mironov, V.S.; Kuznetsov, R.V.

    2007-01-01

    The potentialities of the use of the low-temperature thermionic converters (TIC) with the emitter temperature ≤ 1500 K in the space nuclear power system (SNPS) with the SAFE-type (Safe Affordable Fission Engine) fast reactor proposed and developed by common efforts of American experts have been considered. The main directions of the 'SAFE-300-TEG' SNPS (300 kW(thermal)) design update by replacing the thermoelectric converters with the low-temperature high-performance thermionic converters (with the barrier index V B ≤ 1.9 eV and efficiency ≥ 10%) meant for a long-term operation (5 years at least) as the components of the SAFE-300-TIC SNPS for a Lunar base have been discussed. The concept of the SNPS with the SAFE-type fast reactor and low-temperature TICs with specific electric power of about 1.45 W/cm 2 as the components of the SAFE-300-TIC system meeting the Nasa's initial requirements to a Lunar base with the electric power demand of about 30 kW(electrical) for robotic mission has been considered. The results, involving optimization and mass-and-size estimation, show that the SAFE-300-TIC system meets the initial requirements by Nasa to the lunar base power supply. The main directions of the system update aimed at the output electric power increase up to 100 kW(electrical) have also been presented. (authors)

  18. Power Excursion Accident Analysis of Research Water Reactor

    International Nuclear Information System (INIS)

    Khaled, S.M.; Doaa, G.M.

    2009-01-01

    A three-dimensional neutronic code POWEX-K has been developed, and it has been coupled with the sub-channel thermal-hydraulic core analysis code SV based on the Single Mass Velocity Model. This forms the integrated neutronic/thermal hydraulics code system POWEX-K/SV for the accident analysis. The Training and Research Reactors at Budapest University of Technology and Economics (BME-Reactor) has been taken as a reference reactor. The cross-section generation procedure based on WIMS. The code uses an implicit difference approach for both the diffusion equations and thermal-hydraulics modules, with reactivity feedback effects due to coolant and fuel temperatures. The code system was applied to analyzing power excursion accidents initiated by ramp reactivity insertion of 1.2 $. The results show that the reactor is inherently safe in case of such accidents i.e. no core melt is expected even if the safety rods do not fall into the core

  19. Implications of inherent safe nuclear power system

    International Nuclear Information System (INIS)

    Song, Yo-Taik

    1987-01-01

    The safety of present day nuclear power reactors and research reactors depends on a combination of design features of passive and active systems, and the alert judgement of their operators. A few inherently safe designs of nuclear reactors for power plants are currently under development. In these designs, the passive systems are emphasized, and the active systems are minimized. Also efforts are made to eliminate the potential for human failures that initiate the series of accidents. If a major system fails in these designs, the core is flooded automatically with coolants that flow by gravity, not by mechanical pumps or electromagnetic actuators. Depending on the choice of the coolants--water, liquid metal and helium gas--there are three principal types of inherently safe reactors. In this paper, these inherently safe reactor designs are reviewed and their implications are discussed. Further, future perspectives of their acceptance by nuclear industries are discussed. (author)

  20. Comparative Study on Research Reactor Design Requirements between IAEA and Korea

    International Nuclear Information System (INIS)

    Chang, Won Joon; Yune, Young Gill; Song, Myung Ho; Cho, Seung Ho

    2013-01-01

    This study has identified the gaps in the safety requirements for design of research reactors of Korea comparing with those of the IAEA. The review results showed that the gaps have arisen mainly from the following aspects: - The differences in the characteristics of design and operation between power reactor and research reactor - Enhancement of the level of safety of nuclear reactor facility - Consideration of advanced safety technologies. The review results will be utilized to reflect the IAEA safety requirements for design of research reactors into those of Korea, which will contribute to enhancing the level of safety and improving the technical standards of research reactors of Korea. The IAEA safety standards encompass international consensus to strengthen the nuclear safety and to reflect the latest advancement of nuclear safety technologies. Also, they provide reliable means to ensure the effective fulfillment of obligations under the various international safety conventions. Many countries have adopted the IAEA safety standards as their national standards in nuclear regulations. Since Korea has exported research reactor technologies abroad these days and will continue to export them in the future, it is desirable to harmonize domestic safety requirements for research reactor with those of the IAEA. The KINS (Korea Institute of Nuclear Safety) has performed a review of the IAEA safety requirements for design of research reactors comparing with those of Korea. The purpose of this comparative study is to harmonize the safety requirements for the design of research reactors of Korea with those of the IAEA as a member state of the IAEA, and to encompass global efforts to enhance the nuclear safety and to reflect the latest advancement of nuclear safety technologies into the safety requirements for the design of research reactors of Korea. Design requirements for structures, systems, and components of research reactors important to safety, which are required to

  1. Research reactor support

    International Nuclear Information System (INIS)

    2005-01-01

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  2. Management of spent fuel from research and prototype power reactors and residues from post-irradiation examination of fuel

    International Nuclear Information System (INIS)

    1989-09-01

    The safe and economic management of spent fuel is important for all countries which have nuclear research or power reactors. It involves all aspects of the handling, transportation, storage, conditioning and reprocessing or final disposal of the spent fuel. In the case of spent fuel management from power reactors the shortage of available reprocessing capacity and the rising economic interest in the direct disposal of spent fuel have led to an increasing interest in the long term storage and management of spent fuel. The IAEA has played a major role in coordinating the national activities of the Member States in this area. It was against this background that the Technical Committee Meeting on ''Safe Management of Spent Fuel From Research Reactors, Prototype Power Reactors and Fuel From Commercial Power Reactors That Has Been Subjected to PIE (Post Irradiated Examination)'' (28th November - 1st December 1988) was organised. The aims of the current meeting have been to: 1. Review the state-of-the-art in the field of management of spent fuel from research and prototype power reactors, as well as the residues from post irradiation examination of commercial power reactor fuel. The emphasis was to be on the safe handling, conditioning, transportation, storage and/or disposal of the spent fuel during operation and final decommissioning of the reactors. Information was sought on design details, including shielding, criticality and radionuclide release prevention, heat removal, automation and remote control, planning and staff training; licensing and operational practices during each of the phases of spent fuel management. 2. Identify areas where additional research and development are needed. 3. Recommend areas for future international cooperation in this field. Refs, figs and tabs

  3. Reactor safety research and development in Chalk River Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Atomic Energy of Canada Limited's Chalk River Laboratories provides three different services to stakeholders and customers. The first service provided by the laboratory is the implementation of Research and Development (R&D) programs to provide the underlying technological basis of safe nuclear power reactor designs. A significant portion of the Canadian R&D capability in reactor safety resides at Atomic Energy of Canada Limited's Chalk River Laboratories, and this capability was instrumental in providing the science and technology required to aid in the safety design of CANDU power reactors. The second role of the laboratory has been in supporting nuclear facility licensees to ensure the continued safe operation of nuclear facilities, and to develop safety cases to justify continued operation. The licensing of plant life extension is a key industry objective, requiring extensive research on degradation mechanisms, such that safety cases are based on the original safety design data and valid and realistic assumptions regarding the effect of ageing and management of plant life. Recently, Chalk River Laboratories has been engaged in a third role in research to provide the technical basis and improved understanding for decision making by regulatory bodies. The state-of-the-art test facilities in Chalk River Laboratories have been contributing to the R&D needs of all three roles, not only in Canada but also in the international community, thorough Canada's participation in cooperative programs lead by International Atomic Energy Agency and the OECD's Nuclear Energy Agency. (author)

  4. Passive safe small reactor for distributed energy supply system sited in water filled pit at seaside

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Imayoshi, Shou

    2003-01-01

    Japan Atomic Energy Research Institute has developed a Passive Safe Small Reactor for Distributed Energy Supply System (PSRD) concept. The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down device. For improvement of economy, simplification of the reactor system and long operation of the core over five years without refueling with low enriched UO 2 fuel rods are achieved. To avoid releasing the radioactive materials to the circumstance even if a hypothetical accident, the containment is submerged in a pit filled with seawater at a seaside. Refueling or maintenance of the reactor can be conducted using an exclusive barge instead of the reactor building. (author)

  5. Development of Digital MMIS for Research Reactors: Graded Approaches

    Energy Technology Data Exchange (ETDEWEB)

    Khalil ur, Rahman; Shin, Jin Soo; Heo, Gyun Young [Kyunghee University, Yongin (Korea, Republic of); Son, Han Seong [Joongbu University, Geumsan (Korea, Republic of); Kim, Young Ki; Park, Jae Kwan; Seo, Sang Mun; Kim, Yong Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Though research reactors are small in size yet they are important in terms of industrial applications and R and D, educational purposes. Keeping the eye on its importance, Korean government has intention to upgrade and extend this industry. Presently, Korea is operating only HANARO at Korea Atomic Energy Research Institute (KAERI) and AGN-201K at Kyung Hee University (KHU), which are not sufficient to meet the current requirements of research and education. In addition, we need self-sufficiency in design and selfreliance in design and operation, as we are installing research reactors in domestic as well as foreign territories for instance Jordan. Based on these demands, KAERI and universities initiated a 5 year research project since December 2011 collaboratly, for the deep study of reactor core, thermal hydraulics, materials and instrumentation and control (I and C). This particular study is being carried out to develop highly reliable advanced digital I and C systems using a grading approach. It is worth mentioning that next generation research reactor should be equipped with advance state of the art digital I and C for safe and reliable operation and impermeable cyber security system that is needed to be devised. Moreover, human error is one of important area which should be linked with I and C in terms of Man Machine Interface System (MMIS) and development of I and C should cover human factor engineering. Presently, the digital I and C and MMIS are well developed for commercial power stations whereas such level of development does not exist for research reactors in Korea. Since the functional and safety requirements of research reactors are not so strict as commercial power plants, the design of digital I and C systems for research reactors seems to be graded based on the stringency of regulatory requirements. This paper was motivated for the introduction of those missions, so it is going to describe the general overview of digital I and C systems, the graded

  6. Development of Digital MMIS for Research Reactors: Graded Approaches

    International Nuclear Information System (INIS)

    Khalil ur, Rahman; Shin, Jin Soo; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jae Kwan; Seo, Sang Mun; Kim, Yong Jun

    2012-01-01

    Though research reactors are small in size yet they are important in terms of industrial applications and R and D, educational purposes. Keeping the eye on its importance, Korean government has intention to upgrade and extend this industry. Presently, Korea is operating only HANARO at Korea Atomic Energy Research Institute (KAERI) and AGN-201K at Kyung Hee University (KHU), which are not sufficient to meet the current requirements of research and education. In addition, we need self-sufficiency in design and selfreliance in design and operation, as we are installing research reactors in domestic as well as foreign territories for instance Jordan. Based on these demands, KAERI and universities initiated a 5 year research project since December 2011 collaboratly, for the deep study of reactor core, thermal hydraulics, materials and instrumentation and control (I and C). This particular study is being carried out to develop highly reliable advanced digital I and C systems using a grading approach. It is worth mentioning that next generation research reactor should be equipped with advance state of the art digital I and C for safe and reliable operation and impermeable cyber security system that is needed to be devised. Moreover, human error is one of important area which should be linked with I and C in terms of Man Machine Interface System (MMIS) and development of I and C should cover human factor engineering. Presently, the digital I and C and MMIS are well developed for commercial power stations whereas such level of development does not exist for research reactors in Korea. Since the functional and safety requirements of research reactors are not so strict as commercial power plants, the design of digital I and C systems for research reactors seems to be graded based on the stringency of regulatory requirements. This paper was motivated for the introduction of those missions, so it is going to describe the general overview of digital I and C systems, the graded

  7. Proceedings of the conference on the Safety in Reactor Operations - TopSafe 2012 Transactions

    International Nuclear Information System (INIS)

    2012-01-01

    TopSafe 2012 provides a forum for addressing the current status and future perspectives with regards to safety at nuclear installations worldwide. In view of the on-going discussions and initiatives that have been taken over the last months the European Nuclear Society (ENS) decided organising this edition of this topical conference from 22 to 26 April 2012 in Helsinki, Finland. TopSafe 2012 focus on three main subjects: Safety and related analyses in operating nuclear power plants and other nuclear installations; Safety and Risk Assessment; Trends in nuclear safety for existing and future installations. The conference is directed at a broad range of experts in the area of nuclear safety, including professionals from the different disciplines involved in the safety of nuclear power plants, fuel cycle installations and research reactors. It is aimed at professionals coming from the research organisations, universities, vendors, operators, regulatory bodies as well as policy makers. Top level representatives of the Countries that are constructing new nuclear power plants are invited. Regulators of all individual Countries with nuclear programme are expected to contribute the Conference. (authors)

  8. The application of probabilistic risk assessment to inherently safe reactors

    International Nuclear Information System (INIS)

    Cave, L.; Kastenberg, W.E.

    1987-01-01

    In the development of safety goals and design criteria for 'inherently safe' reactors a question which arises is 'To what extent is PRA relevant.' To answer this question it is necessary to consider both the risk to the public and the investment risk to the utility. In this paper the factors which are likely to determine safety objectives and their allocation are presented. (orig.)

  9. IAEA Coordinated Research Project on the Establishment of a Material Properties Database for Irradiated Core Structural Components for Continued Safe Operation and Lifetime Extension of Ageing Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Borio Di Tigliole, A.; Schaaf, Van Der; Barnea, Y.; Bradley, E.; Morris, C.; Rao, D. V. H. [Research Reactor Section, Vianna (Australia); Shokr, A. [Research Reactor Safety Section, Vienna (Australia); Zeman, A. [International Atomic Energy Agency, Vienna (Australia)

    2013-07-01

    Today more than 50% of operating Research Reactors (RRs) are over 45 years old. Thus, ageing management is one of the most important issues to face in order to ensure availability (including life extension), reliability and safe operation of these facilities for the future. Management of the ageing process requires, amongst others, the predictions for the behavior of structural materials of primary components subjected to irradiation such as reactor vessel and core support structures, many of which are extremely difficult or impossible to replace. In fact, age-related material degradation mechanisms resulted in high profile, unplanned and lengthy shutdowns and unique regulatory processes of relicensing the facilities in recent years. These could likely have been prevented by utilizing available data for the implementation of appropriate maintenance and surveillance programmes. This IAEA Coordinated Research Project (CRP) will provide an international forum to establish a material properties Database for irradiated core structural materials and components. It is expected that this Database will be used by research reactor operators and regulators to help predict ageing related degradation. This would be useful to minimize unpredicted outages due to ageing processes of primary components and to mitigate lengthy and costly shutdowns. The Database will be a compilation of data from RRs operators' inputs, comprehensive literature reviews and experimental data from RRs. Moreover, the CRP will specify further activities needed to be addressed in order to bridge the gaps in the new created Database, for potential follow-on activities. As per today, 13 Member States (MS) confirmed their agreement to contribute to the development of the Database, covering a wide number of materials and properties. The present publication incorporates two parts: the first part includes details on the pre-CRP Questionnaire, including the conclusions drawn from the answers received from

  10. IAEA Coordinated Research Project on the Establishment of a Material Properties Database for Irradiated Core Structural Components for Continued Safe Operation and Lifetime Extension of Ageing Research Reactors

    International Nuclear Information System (INIS)

    Borio Di Tigliole, A.; Schaaf, Van Der; Barnea, Y.; Bradley, E.; Morris, C.; Rao, D. V. H.; Shokr, A.; Zeman, A.

    2013-01-01

    Today more than 50% of operating Research Reactors (RRs) are over 45 years old. Thus, ageing management is one of the most important issues to face in order to ensure availability (including life extension), reliability and safe operation of these facilities for the future. Management of the ageing process requires, amongst others, the predictions for the behavior of structural materials of primary components subjected to irradiation such as reactor vessel and core support structures, many of which are extremely difficult or impossible to replace. In fact, age-related material degradation mechanisms resulted in high profile, unplanned and lengthy shutdowns and unique regulatory processes of relicensing the facilities in recent years. These could likely have been prevented by utilizing available data for the implementation of appropriate maintenance and surveillance programmes. This IAEA Coordinated Research Project (CRP) will provide an international forum to establish a material properties Database for irradiated core structural materials and components. It is expected that this Database will be used by research reactor operators and regulators to help predict ageing related degradation. This would be useful to minimize unpredicted outages due to ageing processes of primary components and to mitigate lengthy and costly shutdowns. The Database will be a compilation of data from RRs operators' inputs, comprehensive literature reviews and experimental data from RRs. Moreover, the CRP will specify further activities needed to be addressed in order to bridge the gaps in the new created Database, for potential follow-on activities. As per today, 13 Member States (MS) confirmed their agreement to contribute to the development of the Database, covering a wide number of materials and properties. The present publication incorporates two parts: the first part includes details on the pre-CRP Questionnaire, including the conclusions drawn from the answers received from the MS

  11. HEU to LEU conversion experience at the UMass-Lowell research reactor

    International Nuclear Information System (INIS)

    White, John R.; Bobek, Leo M.

    2005-01-01

    The UMass-Lowell Research Reactor (UMLRR) operated safely with high-enriched uranium (HEU) fuel for over 25 years. Having reached the end of core lifetime and due to proliferation concerns, the reactor was recently converted to low-enriched uranium silicide (LEU) fuel. The actual process for converting the UMLRR from HEU to LEU fuel covered a period of over 15 years. The conversion effort - from the initial conceptual design studies in the late 1980s to the final offsite shipment of the spent HEU fuel in August 2004 - was a unique experience for the faculty and staff of a small university research reactor. This paper gives a historical view of the process and it highlights several key milestones along the road to successful completion of this project. (author)

  12. Research reactor standards and their impact on the TRIGA reactor community

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

    The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

  13. Integrated Management System, Configuration and Document Control for Research Reactors

    International Nuclear Information System (INIS)

    Steynberg, B.J.; Bruyn, J.F. du

    2017-01-01

    . Document control helps ensure that: 1. Important facility documents are properly stored; 2. Revisions to documents are controlled, tracked, and completed in a timely manner; 3. Revised documents are formally distributed to designated users; and 4. Information concerning pending revisions is made available. Configuration and document management within an integrated management system are essential requirements for the safe operation, utilisation and modification of any research reactor. (author)

  14. Recent progress on developments of tritium safe handling techniques in Japan

    International Nuclear Information System (INIS)

    Watanabe, Kuniaki; Matsuyama, Masao

    1993-01-01

    Vast amounts of tritium will be used for thermonuclear fusion reactors. Without establishing safe handling techniques for large amounts of tritium, undoubtedly the fusion reactors will not be accepted. Japanese activity on tritium related research has considerably developed in the last 10 years. This review paper gives a brief summary of safe handling techniques developed by Japanese research groups. (author)

  15. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  16. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

    Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

  17. Supply of low enriched (LEU) and highly enriched uranium (HEU) for research reactors

    International Nuclear Information System (INIS)

    Mueller, H.

    1997-01-01

    Enriched uranium for research reactors in the form of LEU /= low enriched uranium at 19.75% U-235) and HEU (= highly enriched uranium at 90 to 93% U-235) was and is - due to its high U-235 enrichment - a political fuel other than enriched uranium for power reactors. The sufficient availability of LEU and HEU is a vital question for research reactors, especially in Europe, in order to perform their peaceful research reactor programs. In the past the USA were in the Western hemisphere sole supplier of LEU and HEU. Today the USA have de facto stopped the supply of LEU and HEU, for HEU mainly due to political reasons. This paper deals, among others, with the present availability of LEU and HEU for European research reactors and touches the following topics: - historical US supplies, - influence of the RERTR-program, - characteristics of LEU and HEU, - military HEU enters the civil market, -what is the supply situation for LEU and HEU today? - outlook for safe supplies of LEU and HEU. (author)

  18. Multi purpose research reactor

    International Nuclear Information System (INIS)

    Raina, V.K.; Sasidharan, K.; Sengupta, Samiran; Singh, Tej

    2006-01-01

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor

  19. Proceedings of the 1999 workshop on the utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-01

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

  20. Concept of passive safe small reactor for distributed energy supply system

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Nakajima, Nobuya; Sawada, Ken-ichi; Yoritsune, Tsutomu; Shimada, Shoichiro; Nakano, Yoshihiro; Yonomoto, Taisuke; Takahashi, Hiroki

    2003-01-01

    This paper presents a concept of a Passive Safe Small Reactor for Distributed energy supply system (PSRD). The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down. To comply with a severe operation condition of PSRD, material of the ball bearing with graphite retainer has been selected by test. For improvement of economy, simplification of the reactor system and long operation of the core are achieved. Optimization of core design concerning the burnable poison ensures the burn-up of 28 GWd/t for low enriched UO 2 fuel rods. (author)

  1. Review of inservice inspection and nondestructive examination practices at DOE Category A test and research reactors

    International Nuclear Information System (INIS)

    Anderson, M.T.; Aldrich, D.A.

    1990-09-01

    In-service inspection (ISI) programs are used at commercial nuclear power plants for monitoring the pressure boundary integrity of various systems and components to ensure their continued safe operation. The Department of Energy (DOE) operates several test and research reactors. This report represents an evaluation of the ISI and nondestructive examination (NDE) practices at five DOE Category A (> 20 MW thermal) reactors as compared, where applicable, to the current ISI activities of commercial nuclear power facilities. The purpose of an inservice inspection (ISI) program is to establish regular surveillance of safety-related components to ensure their safe and reliable operation. The integrity of materials comprising these components is generally monitored by means of periodic nondestructive examinations (NDE), which, if appropriately performed, provide methods for identifying degradation that could render components unable to perform their intended safety functions. The reactors evaluated during this review were the Experimental Breeder Reactor 2 and the Fast Flux Test Facility (liquid-metal cooled plants), the Advanced Test Reactor and the High Flux Isotopes Reactor (light-water cooled reactors), and the High Flux Beam Reactor (a heavy-water cooled facility). Although these facilities are extremely diverse in design and operation, they all have less stored energy, smaller inventories of radionuclides, and generally, more remote locations than commercial reactors. However, all DOE test and research facilities contain components similar to those of commercial reactors for which continued integrity is important to maintain plant safety. 10 refs., 6 tabs

  2. Ordeals of Chernobyl and the rejustification of the inherently safe reactors

    International Nuclear Information System (INIS)

    Lu, Y.

    1989-01-01

    This paper presents the necessity of developing inherently safe economic reactors (ISERs). Two characteristics which define inherent safety are discussed on the basis of various applications of such a principle in practice. Different design concepts of ISERs are then evaluated and their possible role in the future nuclear program of PRC discussed. A three-stage development strategy of ISERs in PRC is proposed

  3. Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

    International Nuclear Information System (INIS)

    1975-06-01

    The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

  4. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  5. CER. Research reactors in France

    International Nuclear Information System (INIS)

    Estrade, Jerome

    2012-01-01

    Networking and the establishment of coalitions between research reactors are important to guarantee a high technical quality of the facility, to assure well educated and trained personnel, to harmonize the codes of standards and the know-ledge of the personnel as well as to enhance research reactor utilization. In addition to the European co-operation, country-specific working groups have been established for many years, such as the French research reactor Club d'Exploitants des Reacteurs (CER). It is the association of French research reactors representing all types of research reactors from zero power up to high flux reactors. CER was founded in 1990 and today a number of 14 research reactors meet twice a year for an exchange of experience. (orig.)

  6. A reload and startup plan for conversion of the NIST research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-03-31

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  7. Radiation protection at the RA Reactor in 1993, RA research reactor, Part

    International Nuclear Information System (INIS)

    Ninkovic, M.; Pavlovic, R.; Mandic, M.; Sipka, V.; Grsic, Z.

    1993-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry and radiation protection at the RA reactor; (2) decontamination, collecting and treatment of fluid effluents and solid wastes; (3) Radioactivity control in the vicinity of the reactor and (4)meteorology measurements; (3). Each of the category is described as a separate annex of this report [sr

  8. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  9. Computerized reactor monitor and control for research reactors

    International Nuclear Information System (INIS)

    Buerger, L.; Vegh, E.

    1981-09-01

    The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

  10. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  11. Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

    International Nuclear Information System (INIS)

    Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

    1993-01-01

    Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

  12. A novel concept for CRIEC-driven subcritical research reactors

    International Nuclear Information System (INIS)

    Nieto, M.; Miley, G.H.

    2001-01-01

    A novel scheme is proposed to drive a low-power subcritical fuel assembly by means of a long Cylindrical Radially-convergent Inertial Electrostatic Confinement (CRIEC) used as a neutron source. The concept is inherently safe in the sense that the fuel assembly remains subcritical at all times. Previous work has been done for the possible implementation of CRIEC as a subcritical assembly driver for power reactors. However, it has been found that the present technology and stage of development of IEC-based neutron sources can not meet the neutron flux requirements to drive a system as big as a power reactor. Nevertheless, smaller systems, such as research and training reactors, could be successfully driven with levels of neutron flux that seem more reasonable to be achieved in the near future by IEC devices. The need for custom-made expensive nuclear fission fuel, as in the case of the TRIGA reactors, is eliminated, and the CRIEC presents substantial advantages with respect to the accelerator-driven subcritical reactors in terms of simplicity and cost. In the present paper, a conceptual design for a research/training CRIEC-driven subcritical assembly is presented, emphasizing the description, principle of operation and performance of the CRIEC neutron source, highlighting its advantages and discussing some key issues that require study for the implementation of this concept. (author)

  13. Strategic planning for research reactors. Guidance for reactor managers

    International Nuclear Information System (INIS)

    2001-04-01

    The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

  14. Engineering and planning for reactor 105-C interim safe storage project subcontract no. 0100C-SC-G0001 conceptual design report. Volume 1

    International Nuclear Information System (INIS)

    1996-04-01

    The 105-C Reactor, one of eight surplus production reactors at the Hanford Site, has been proposed by the U.S. Department of Energy, Richland, Operations Office to be the first large-scale technology demonstration project in the decontamination and decommissioning (D ampersand D) focus area as part of the project for dismantlement and interim safe storage. The 105-C Reactor will be placed in an interim safe storage condition, then undergo the decontamination and decommissioning phase. After D ampersand D, the reactor will be placed in long- term safe storage. This report provides the conceptual design for these activities

  15. Transportation of failed or damaged foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Messick, C.E.; Mustin, T.P.; Massey, C.D.

    1998-01-01

    Since resuming the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy (DOE) and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues associated with the transport of Materials Testing Reactor (MTR)-type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be important to foreign research reactor operators, shippers, and cask vendors, so that appropriate amendments to the Certificate of Compliance for spent fuel casks can be submitted in a timely manner to facilitate the safe and scheduled transport of FRR SNF

  16. The Research on Operation Strategy of Nuclear Power Plant with Multi-reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Maoyao; Peng, Minjun; Cheng Shouyu [Harbin Engineering University, Harbin (China)

    2014-08-15

    In this paper, the operation characteristics and control strategy of nuclear power plant (NPP) with multi-modular pressurized water reactors (PWR) were researched through simulation. The main objective of this research was to ensure the coordinated operation and satisfy the convenience of turbine-generator and reactor's load adjustment in NPP with multi-reactors (MR). According to the operation characteristics of MR-NPP, the operation and control strategy was proposed, which was 'he average allocation of load for each reactor and maintaining average temperature of coolant at a constant? The control system was designed based the operation and control strategy. In order to research the operation characteristics and control strategy of MR-NPP, the paper established the transient analysis model which included the reactors and thermal hydraulic models, turbine model, could simulate and analyze on different operating conditions such as load reducing, load rising. Based on the proposed operation and control strategy and simulation models, the paper verified and validated the operation strategy and control system through load reducing, load rising. The results of research simulation showed that the operation strategy was feasible and can make the MR-NPP running safely as well as steadily on different operating conditions.

  17. The Research on Operation Strategy of Nuclear Power Plant with Multi-reactors

    International Nuclear Information System (INIS)

    Fang, Maoyao; Peng, Minjun; Cheng Shouyu

    2014-01-01

    In this paper, the operation characteristics and control strategy of nuclear power plant (NPP) with multi-modular pressurized water reactors (PWR) were researched through simulation. The main objective of this research was to ensure the coordinated operation and satisfy the convenience of turbine-generator and reactor's load adjustment in NPP with multi-reactors (MR). According to the operation characteristics of MR-NPP, the operation and control strategy was proposed, which was 'he average allocation of load for each reactor and maintaining average temperature of coolant at a constant? The control system was designed based the operation and control strategy. In order to research the operation characteristics and control strategy of MR-NPP, the paper established the transient analysis model which included the reactors and thermal hydraulic models, turbine model, could simulate and analyze on different operating conditions such as load reducing, load rising. Based on the proposed operation and control strategy and simulation models, the paper verified and validated the operation strategy and control system through load reducing, load rising. The results of research simulation showed that the operation strategy was feasible and can make the MR-NPP running safely as well as steadily on different operating conditions

  18. The Jules Horowitz Reactor project, a driver for revival of the research reactor community

    International Nuclear Information System (INIS)

    Pere, P.; Cavailler, C.; Pascal, C.

    2010-01-01

    The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first-rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10 14 n.cm -2 /sec -1 E>1 MeV in core and 3,6x10 14 n.cm -2 /sec -1 E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960s, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items:accommodation of a broad experimental domain; involvement by international partners making in-kind contributions to the project; ? development of components critical to safety and performance; the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and; tools to carry out the project, including computer codes

  19. Digital control of research reactors

    International Nuclear Information System (INIS)

    Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

    1991-01-01

    Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind

  20. Man-machine systems research at the OECD Halden reactor project

    International Nuclear Information System (INIS)

    Owre, F.; Bjorlo, T.J.; Haugset, K.

    1994-01-01

    The OECD Halden Reactor Project is a jointly financed research programme under the auspices of the OECD - Nuclear Energy Agency with fifteen participating countries. One of the main research topics focuses on man-machine systems. Particular attention is paid to the operator's tasks in the reactor control room environment. The overall objective of the research in this field is to provide a basis for improving today's control rooms through the introduction of computer-based solutions for the effective and safe execution of surveillance and control functions in normal as well as off-normal plant situations. The programme comprises four main activities: the verification and validation of safety critical software systems; man-machine interaction research emphasizing improvements in man-machine interfaces on the basis of human factors studies; computerised operator support systems assisting the operator in fault detection/diagnosis and planning of control actions; and control room development providing a basis for retrofitting of existing control rooms and for the design of advanced concepts

  1. Model development for the dynamic analysis of the OSU inherently safe reactor. Part 1

    International Nuclear Information System (INIS)

    Aybar, H.S.

    1992-01-01

    Faculty and students in the Nuclear Engineering Program at the Ohio State University (OSU) have proposed a conceptual design for an inherently safe 340 MWe power reactor. The design is based on the state-of-the-art technology of LWRs and the High Temperature Gas- cooled Reactors (HTGRs). The OSU Inherently Safe Reactor (OSU-ISR) concept uses shorter than standard BWR fuel elements in the reactor core. All the fluid on the primary side is contained within a Prestressed Concrete Reactor Vessel (PCRV). This important feature significantly reduces the probability of a LOCA. A new feature of the OSU-ISR is an operator independent steam driven Emergency Core Cooling System (ECCS) housed within the PCRV. In accident conditions where the steam generators are incapacitated, steam from the core drives a jet injector, which takes water from the suppression pool and pumps it into the core cavity to maintain core coverability. The preliminary analysis of the concept was performed as a design project in the Nuclear Engineering Program at the OSU during the Spring of 1985, and published in ''Nuclear Technology.'' The use of a PCRV for ducting and containment and the replacement of forced recirculation with natural circulation on the primary side significantly improve the inherent safety of the plant. Currently, work is in progress for the refinement of the OSU-ISR concept, partially supported by a grant from the U.S. Department of Energy

  2. Making the world more secure, one research reactor at a time

    International Nuclear Information System (INIS)

    Mutluer, Adem; Rickwood, Peter

    2015-01-01

    During the night of 29 September 2014, a heavy transport plane took off from an air base in Kazakhstan after an operation to remove fuel and increase the security of a research reactor. In its cargo bay sat four massive containers, provided by the IAEA, that had been filled with a total of 10.2 kilograms of highly enriched uranium (HEU), on its way to be diluted to a safe substance or securely stored at the flight’s destination in Russia. The operation represented the latest achievement in a global programme involving the IAEA, the Russian Federation and the United States to assist several countries, including Kazakhstan, in eliminating the risks associated with HEU, while still maintaining the important scientific research conducted at the reactor. HEU is a security risk, as it is an ingredient that can be used to create a nuclear device intended for malicious use. It is not encouraged to use HEU in a research reactor as safer low enriched uranium (LEU) can be used instead (see box). In the 1960s and 1970s, when many of the world’s research reactors were built, technology using LEU was not yet available, so in order to perform experiments HEU fuel was required. As of next year, less proliferation sensitive LEU will be used to fuel the light water research reactor in Alatau near Almaty, the largest city in Kazakhstan.

  3. Making the world more secure, one research reactor at a time

    International Nuclear Information System (INIS)

    Mutluer, Adem; Rickwood, Peter

    2015-01-01

    During the night of 29 September 2014, a heavy transport plane took off from an air base in Kazakhstan after an operation to remove fuel and increase the security of a research reactor. In its cargo bay sat four massive containers, provided by the IAEA, that had been filled with a total of 10.2 kilograms of highly enriched uranium (HEU), on its way to be diluted to a safe substance or securely stored at the flight’s destination in Russia. The operation represented the latest achievement in a global programme involving the IAEA, the Russian Federation and the United States to assist several countries, including Kazakhstan, in eliminating the risks associated with HEU, while still maintaining the important scientific research conducted at the reactor. HEU is a security risk, as it is an ingredient that can be used to create a nuclear device intended for malicious use. It is not encouraged to use HEU in a research reactor as safer low enriched uranium (LEU) can be used instead (see box). In the 1960s and 1970s, when many of the world’s research reactors were built, technology using LEU was not yet available, so in order to perform experiments HEU fuel was required. As of next year, less proliferation sensitive LEU will be used to fuel the light water research reactor in Alatau near Almaty, the largest city in Kazakhstan. .

  4. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  5. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

  6. Research reactors and materials testing

    International Nuclear Information System (INIS)

    Vidal, H.

    1986-01-01

    Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

  7. Directory of Nuclear Research Reactors 1994

    International Nuclear Information System (INIS)

    1995-08-01

    The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

  8. Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

    International Nuclear Information System (INIS)

    Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

    2010-01-01

    The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

  9. Spanish collaboration in the OECD Halden Reactor Project research on Gadolinia Fuel

    International Nuclear Information System (INIS)

    Horvath, M. I.; Jenssen, H. K.; Munoz-Reja, C.; Tverberg, T.

    2011-01-01

    Safe and reliable operation of nuclear power plants benefit from research and development advances and related technical solutions. One research platform is the OECD Halden Reactor Project (HRP), HRP is a joint undertaking of national organisations in 18 countries sponsoring a jointly financed programme under the auspices of the OECD-Nuclear Energy Agency (NEA). As a member state, Spain is participating HRP research programs with ENUSA as partner in the fuel research programs. Various experiments are developed and performed also by providing materials, ENUSA collaborates with HRP on various experiments investigating the fuel behaviour, especially on Gd-bearing fuel. 20 years of successful collaboration between HRP and ENUSA is continuing with promising and results to ensure and enhance the safe operation of the Spanish and all other NPPs in the world. (Author) 12 refs.

  10. Code of Conduct on the Safety of Research Reactors

    International Nuclear Information System (INIS)

    2006-09-01

    with the advice of an expert Working Group. The revised draft was again circulated to all Member States for comment, along with the report of the Chairman of the expert Working Group. In March 2004, a final draft of the Code was submitted to the Board of Governors for adoption. When the Board of Governors adopted the Code of Conduct, it requested the Director General to circulate the approved Code of Conduct to all Member States and relevant international organizations and transmit it to the General Conference with a recommendation that the Conference endorse it and call for its wide application. In resolution GC(48)/RES/10.A.8, the 2004 General Conference welcomed the adoption of the Code by the Board, and endorsed the guidance for the safe operation of research reactors set out in the Code. It also encouraged Member States to apply the guidance in the Code to the management of research reactors. Finally, it requested the Secretariat to continue to assist Member States in the implementation of the Code and associated safety guidance within available resources. While the Code of Conduct on the Safety of Research Reactors is non-binding, it should serve as guidance on the development and harmonization of laws, regulations and policies on the safety of research reactors. It provides ''best practice'' guidance to the State, the regulatory body and the operating organization for management of research reactor safety. In accordance with resolution GC(48)/RES/10, Member States are encouraged to use this Code of Conduct as the basis upon which to regulate and conduct research reactor activities.

  11. Physical security at research reactors

    International Nuclear Information System (INIS)

    Clark, R.A.

    1977-01-01

    Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

  12. Thermal hydraulic analyses of LVR-15 research reactor with IRT-M fuel

    International Nuclear Information System (INIS)

    Macek, J.

    1997-01-01

    The LVR-15 pool-type research reactor has been in operation at the Nuclear Research Institute at Rez since 1955. Following a number of reconstructions and redesigning, the current reactor power is 15 MW. Thermal hydraulic analyses to demonstrate that the core heat will be safely removed during operation as well as in accident situations were performed based on methodology which had been specifically developed for the LVR-15 research reactor. This methodology was applied to stationary thermal hydraulic computations, as well as to transients, particularly with reactivity failure and loss of circulation pumps emergencies. The applied methodology and the core configuration as used in the Safety Report are described. The initial and boundary conditions are then considered and the summary of the calculated failures with regard to the defined safety limits is presented. The results of the core configuration analyses are also discussed with respect to meeting the safety limits and to the applicability of the methodology to this purpose

  13. Research and materials irradiation reactors

    International Nuclear Information System (INIS)

    Ballagny, A.; Guigon, B.

    2004-01-01

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

  14. RA Research nuclear reactor, Part II: radiation protection at the RA reactor in 1987

    International Nuclear Information System (INIS)

    Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

    1987-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

  15. Safe use of the Institute of Nuclear Physics reactor with low enriched fuel

    International Nuclear Information System (INIS)

    Baytelesov, S.A.; Dosimbaev, A.A.; Koblik, Yu.N.; Salikhbaev, U.S.; Khalikov, U.A.; Yuldashev, B.S.

    2006-01-01

    Full text: The requirements for safe exploitation of reactor do not accept boiling of water on the surface of fuel elements. At determination of safe thermal regime of reactor (permissible level of power) the regime of the most heat-stressed fuel assembly (FA) in the active core was analyzed. By using ASTRA code [1] the heat-stressed sector is determined by most heat-stressed FA. In calculations the power of reactor was selected so that stock factor prior to the water boiling on the FA surface was not less than 1.45. Besides, in calculations the value of maximal energy density in examined FA is decreased by 10 %. As the part of the energy generated in the FA cores will be lost in constructional materials of the active zone and on the reflector. The stocks of safety before occurrence of instability of flow in gaps between of FA and before crisis of heat exchange are also analyzed. Further, by using the MCNP-4C code [2], densities of fast (E > 0,821 MeV) and thermal flows (E < 0,625 eV) of neutrons were calculated for those experimental channels where the irradiation of samples would be carried out. (author)

  16. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2001-01-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  17. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  18. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

    2000-01-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  19. Core conversion anaylses for the Portuguese Research Reactor

    International Nuclear Information System (INIS)

    Matos, J.E.; Stevens, J.G.; Feldman, E.E.; Stillman, J.A.; Dunn, F.E.; Kalimullah, K.; Marques, J.G.; Barradas, N.P.; Ramos, A.R.; Kling, A.

    2006-01-01

    Design and safety analyses are presented for conversion of the Portuguese Research Reactor (RPI) from the use of HEU fuel to the use of LEU fuel. The analyses were performed jointly by the RERTR Program at the Argonne National Laboratory (ANL) and the Instituto Tecnologico e Nuclear (ITN). The LEU fuel assembly design uses U 3 Si 2 -Al dispersion fuel with 4.8 g U/cm 3 and is very similar to the HEU fuel design. The results of neutronic studies, steady-state thermal-hydraulic analyses, accident analyses, and revisions to the Operating Limits and Conditions demonstrate that the RPI reactor can be operated safely with the new LEU fuel assemblies. Delivery of the LEU fuel is expected around the end of 2006, with conversion in early 2007. The HEU fuel is planned to be returned to the US in 2008.

  20. World-wide French experience in research reactor fuel cycle transportation

    International Nuclear Information System (INIS)

    Raisonnier, D.

    1997-01-01

    Since 1963 Transnucleaire has safely performed a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied suitable packagings for all types of nuclear fuel cycle radioactive material from front-end and back-end products and for power or for research reactors. Transportation of the nuclear fuel material for power reactors is made on a regular and industrial basis. The transportation of material for the research reactor fuel cycle is quite different due to the small quantities involved, the categorisation of material and the numerous places of delivery world-wide. Adapted solutions exist, which require a reactive organisation dealing with all the transportation issues for LEU and HEU products as metal, oxide, fresh fuel elements, spent fuel elements including supply of necessary transport packaging and equipment. This presentation will: - explain the choices made by Transnucleaire and its associates to provide and optimise the corresponding services, - demonstrate the capability to achieve, through reliable partnership, transport operations involving new routes, specific equipment and new political constraints while respecting sophisticated safety and security regulations. (author)

  1. Control system design for a 100 MW(th) research reactor

    International Nuclear Information System (INIS)

    Seshadri, S.N.; Ranganath, M.V.; Singh, Manjit.

    1983-01-01

    This paper presents the computer simulation carried out to evolve a suitable analog controller for a 100 MW(th) heavy water moderated research reactor under construction at Trombay. The control action is based on the average neutron flux in the reactor core and the reactivity is controlled by adjusting the moderator level in the calandria. A dual control scheme controlling the inflow as well as the outflow was adopted in order to fully exploit the capabilities of control elements. For reasons of reliability, the system consists of three identical channels enabling safe operation even under one channel failure. Based on the simulation studies a suitable compensation network was incorporated to achieve satisfactory system response. (author)

  2. Research reactors; Les piles de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Kowarski, L. [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires]|[Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

    1955-07-01

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

  3. Research reactors; Les piles de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Kowarski, L [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires; [Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

    1955-07-01

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

  4. Planning the Decommissioning of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Podlaha, J., E-mail: pod@ujv.cz [Nuclear Research Institute Rez, 25068 Rez (Czech Republic)

    2013-08-15

    In the Czech Republic, three research nuclear reactors are in operation. According to the valid legislation, preliminary decommissioning plans have been prepared for all research reactors in the Czech Republic. The decommissioning plans shall be updated at least every 5 years. Decommissioning funds have been established and financial resources are regularly deposited. Current situation in planning of decommissioning of research reactors in the Czech Republic, especially planning of decommissioning of the LVR-15 research reactor is described in this paper. There appeared new circumstances having wide impact on the decommissioning planning of the LVR-15 research reactor: (1) Shipment of spent fuel to the Russian Federation for reprocessing and (2) preparation of processing of radioactive waste from reconstruction of the VVR-S research reactor (now LVR-15 research reactor). The experience from spent fuel shipment to the Russian Federation and from the process of radiological characterization and processing of radioactive waste from reconstruction of the VVR-S research reactor (now the LVR-15 research reactor) and the impact on the decommissioning planning is described in this paper. (author)

  5. Power control of SAFE reactor using fuzzy logic

    International Nuclear Information System (INIS)

    Irvine, Claude

    2002-01-01

    Controlling the 100 kW SAFE (Safe Affordable Fission Engine) reactor consists of design and implementation of a fuzzy logic process control system to regulate dynamic variables related to nuclear system power. The first phase of development concentrates primarily on system power startup and regulation, maintaining core temperature equilibrium, and power profile matching. This paper discusses the experimental work performed in those areas. Nuclear core power from the fuel elements is simulated using resistive heating elements while heat rejection is processed by a series of heat pipes. Both axial and radial nuclear power distributions are determined from neuronic modeling codes. The axial temperature profile of the simulated core is matched to the nuclear power profile by varying the resistance of the heating elements. The SAFE model establishes radial temperature profile equivalence by establishing 32 control zones as the nodal coordinates. Control features also allow for slow warm up, since complete shutoff can occur in the heat pipes if heat-source temperatures drop/rise below a certain minimum value, depending on the specific fluid and gas combination in the heat pipe. The entire system is expected to be self-adaptive, i.e., capable of responding to long-range changes in the space environment. Particular attention in the development of the fuzzy logic algorithm shall ensure that the system process remains at set point, virtually eliminating overshoot on start-up and during in-process disturbances. The controller design will withstand harsh environments and applications where it might come in contact with water, corrosive chemicals, radiation fields, etc

  6. Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

    International Nuclear Information System (INIS)

    ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

    2017-01-01

    The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

  7. Assessment for the potential of Stuxnet attack on research reactor in Indonesia

    International Nuclear Information System (INIS)

    Eko H Riyadi

    2013-01-01

    Since discovered in June of 2010, the malware that named Stuxnet was directly electrify the world. The malware is interested the attentions of the media and security experts since infect instrumentation and control systems in industrial and caused significant damage. Stuxnet is a sophisticated kind of computer malware designed to sabotage industrial processes controlled by Siemens SIMATIC WinCC and PCS7. Even the malware developed to attack Iran's nuclear facilities have very specific goals and designed to infect Siemens SCADA systems are set up to control and monitor industrial processes. In Indonesia there are 3 research reactors i.e. Kartini in Jog jakarta, TRIGA in Bandung and RSG GA Siwabessy in Serpong. Last reactor has the largest capacity that was using the instrumentation and control system of the Siemens made. This study aims to determine the potential of the Stuxnet attack on research reactor with use methods and analysis the learning to the malware. With the configuration of the control system on research reactors, it is not impossible will be target of the Stuxnet attacks like as Iran's nuclear facility in Natanz. Although some security experts and anti virus have develop system patch and removal to fix this malware, but still needed the role of regulatory bodies to ensure that instrumentation and control system in research reactors are totally safe and secure from malware attacks. (author)

  8. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  9. Nuclear research reactors in Brazil

    International Nuclear Information System (INIS)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias

    2011-01-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  10. Industrial structure at research reactor suppliers

    International Nuclear Information System (INIS)

    Roegler, H.-J.; Bogusch, E.; Friebe, T.

    2001-01-01

    Due to the recent joining of the forces of Framatome S. A. from France and the Nuclear Division of Siemens AG Power Generation (KWU) from Germany to a Joint Venture named Framatome Advanced Nuclear Power S.A.S., the issue of the necessary and of the optimal industrial structure for nuclear projects as a research reactor is, was discussed internally often and intensively. That discussion took place also in the other technical fields such as Services for NPPs but also in the field of interest here, i. e. Research Reactors. In summarizing the statements of this presentation one can about state that: Research Reactors are easier to build than NPPs, but not standardised; Research Reactors need a wide spectrum of skills and experiences; to design and build Research Reactors needs an experienced team especially in terms of management and interfaces; Research Reactors need background from built reference plants more than from operating plants; Research Reactors need knowledge of suitable experienced subsuppliers. Two more essential conclusions as industry involved in constructing and upgrading research reactors are: Research Reactors by far are more than a suitable core that generates a high neutron flux; every institution that designs and builds a Research Reactor lacks quality or causes safety problems, damages the reputation of the entire community

  11. Research reactor`s role in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Choi, C-O [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1996-12-31

    After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960`s in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs.

  12. Australian research reactor studies

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1978-01-01

    The Australian AEC has two research reactors at the Lucas Heights Research Establishment, a 10 HW DIDO class materials testing reactor, HIFAR, and a smaller 100kW reactor MOATA, which was recently upgraded from 10kW power level. Because of the HIFAR being some 20 years old, major renewal and repair programmes are necessary to keep it operational. To enable meeting projected increases in demand for radioisotopes, plans for a new reactor to replace the HIFAR have been made and the design criteria are described in the paper. (author)

  13. 2012 review of French research reactors

    International Nuclear Information System (INIS)

    Estrade, Jerome

    2013-01-01

    Proposed by the French Reactor Operators' Club (CER), the meeting and discussion forum for operators of French research reactors, this report first gives a brief presentation of these reactors and of their scope of application, and a summary of highlights in 2012 for each of them. Then, it proposes more detailed presentations and reviews of characteristics, activities, highlights, objectives and results for the different types of reactors: neutron beam reactors (Orphee, High flux reactor-Laue-Langevin Institute or HFR-ILL), technological irradiation reactors (Osiris and Phenix), training reactors (Isis and Azur), reactors for safety research purposes (Cabri and Phebus), reactors for neutronic studies (Caliban, Prospero, Eole, Minerve and Masurca), and new research reactors (the RES facility and the Jules Horowitz reactor or JHR)

  14. Inherent Safety Feature of Hybrid Low Power Research Reactor during Reactivity Induced Accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, DongHyun; Yum, Soo Been; Hong, Sung Teak; Lim, In-Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Hybrid low power research reactor(H-LPRR) is the new design concept of low power research reactor for critical facility as well as education and training. In the case of typical low power research reactor, the purposes of utilization are the experiments for education of nuclear engineering students, Neutron Activation Analysis(NAA) and radio-isotope production for research purpose. H-LPRR is a light-water cooled and moderated research reactor that uses rod-type LEU UO{sub 2} fuels same as those for commercial power plants. The maximum core thermal power is 70kW and, the core is placed in the bottom of open pool. There are 1 control rod and 2 shutdown rods in the core. It is designed to cool the core by natural convection, retain negative feedback coefficient for entire fuel periods and operate for 20 years without refueling. Inherent safety in H-LPRR is achieved by passive design features such as negative temperature feedback coefficient and core cooling by natural convection during normal and emergency conditions. The purpose of this study is to find out that the inherent safety characteristics of H-LPRR is able to control the power and protect the reactor from the RIA(Reactivity induced accident). RIA analysis was performed to investigate the inherent safety feature of H-LPRR. As a result, it was found that the reactor controls its power without fuel damage in the event and that the reactor remains safe states inherently. Therefore, it is believed that high degree of safety inheres in H-LPRR.

  15. The Jules Horowitz reactor project, a driver for revival of the research reactor community

    Energy Technology Data Exchange (ETDEWEB)

    Pere, P.; Cavailler, C.; Pascal, C. [AREVA TA, CEA Cadarache - Etablissement d' AREVA TA - Chantier RJH - MOE - BV2 - BP no. 9 - 13115 Saint Paul lez Durance (France); CS 50497 - 1100, rue JR Gauthier de la Lauziere, 13593 Aix en Provence cedex 3 (France)

    2010-07-01

    The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first -rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10{sup 14} n.cm{sup -2}/sec{sup -1} E> 1 MeV in core and 3,6x10{sup 14} n.cm{sup -2}/sec{sup -1} E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960's, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items: - accommodation of a broad experimental domain, - involvement by international partners making in-kind contributions to the project, - development of components critical to safety and performance, - the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and

  16. Feasibility study for Tehran Research Reactor power upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem [Nuclear Research Center, Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)], E-mail: kfarhadi@aeoi.org.ir; Khakshournia, Samad [Nuclear Research Center, Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)

    2008-07-15

    The present work is concerned with a power upgrading study of Tehran Research Reactor (TRR). The upgrading study is aimed at investigating the possibility of raising power of the TRR from the current level of 5 MW{sub th} to a higher level without violating the original thermal-hydraulic safety criteria. The existing core, comprising 22 standard fuel elements and five control fuel elements, is used for the analyses. Different reactor thermal powers (5-11 MW) and different core coolant flow rates (500-921 m{sup 3}/h) are considered. It is shown that, for the present core, this goal could be achieved safely by gradually opening the butterfly control valve until the desired coolant flow rate is reached. The TRR power could be upgraded up to around 7.5 MW{sub th} with the total power peaking factor maintained at less than or equal to 3.0.

  17. Feasibility study for Tehran Research Reactor power upgrading

    International Nuclear Information System (INIS)

    Farhadi, Kazem; Khakshournia, Samad

    2008-01-01

    The present work is concerned with a power upgrading study of Tehran Research Reactor (TRR). The upgrading study is aimed at investigating the possibility of raising power of the TRR from the current level of 5 MW th to a higher level without violating the original thermal-hydraulic safety criteria. The existing core, comprising 22 standard fuel elements and five control fuel elements, is used for the analyses. Different reactor thermal powers (5-11 MW) and different core coolant flow rates (500-921 m 3 /h) are considered. It is shown that, for the present core, this goal could be achieved safely by gradually opening the butterfly control valve until the desired coolant flow rate is reached. The TRR power could be upgraded up to around 7.5 MW th with the total power peaking factor maintained at less than or equal to 3.0

  18. A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Varuttamaseni, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2017-09-30

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  19. An Overview of Ageing Management and Refurbishment of Research Reactors at Trombay

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, R. C.; Raina, V. K. [Bhabha Atomic Research Centre, Mumbai (India)

    2014-08-15

    Three nuclear research reactors have been in operation at Bhabha Atomic Research Centre, Mumbai, India. India has a rich experience of about 120 research reactor operating years including ageing management. A well structured programme is in force for plant life management, refurbishment and upgrading reactors in operation. Apsara, commissioned in August 1956, was the first research reactor. Apsara is a 1 MW{sub th} swimming pool type of reactor with a movable core loaded with enriched uranium fuel and immersed in demineralized light water pool, which serves as coolant, moderator and reflector besides providing radiation shielding. Apsara was shut down during May 2009 for partial decommissioning and upgrading to a 2 MW reactor with several safety upgrades, e.g. a LEU based reactor core with higher neutron flux, a new reactor building meeting seismic qualification criteria and two independent shutdown devices. Cirus, a 40 MW{sub th} tank type reactor utilizing heavy water as moderator, graphite as reflector, demineralized light water as primary coolant and natural uranium metal as fuel; has been in operation since 1960. After about three decade of operation, the availability factor started declining mainly due to outage of equipment exhibiting signs of ageing. After ageing studies and performance review, refurbishment requirements were identified. A programme for refurbishment was drawn that included safety upgrades like civil repairs to the emergency storage reservoir to meet seismic qualification criteria and a new iodine removal system for better efficiency. The reactor was shut down during 1997 for execution of this refurbishment programme. After completion of refurbishment, the reactor was brought back into operation during 2003. It has completed about seven years of safe operation after refurbishment with a significant increase in availability factor from 70% to about 90%. The reactor was permanently shut down during December 2010. The reactor core was unloaded

  20. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  1. Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    Balogun, G.I.; Jonah, S.A.; Umar, I.M.

    2005-01-01

    Safety culture has been defined as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. This paper briefly highlights efforts being made at the Centre for Energy Research and Training (CERT) towards realizing this broad objective as far as possible. To this end CERT realizes the need for instituted safety measures to reflect significant, site-specific peculiar characteristics of any generic reactor types. Consequently, standard procedures for pre-startup, startup and shutdown of NIRR-1 (a miniature neutron source reactor - MNSR) have been reviewed to reflect our local conditions and peculiarities. The review has revealed the need to incorporate important steps that impact on overall safety of the facility. For instance an interlocking system is being considered between NIRR-1 startup on the one hand and mandatory pre-startup measures on the other. Also a procedure has been put in place that would facilitate rapid response in the event of a rod-stuck-at-full-withdrawal incident. Furthermore, a program of automation of important analysis and design calculations of MNSRs is going on. Emphases are also placed, and deliberate efforts are being made, to ensure that a working atmosphere prevails that would foster the correct attitudinal approach to matters of reactor safety. A regime of constant dialogue and discussions amongst operating personnel has been factored into the overall operational program. (author)

  2. Replacement research reactor for Australia

    International Nuclear Information System (INIS)

    Miller, Ross

    1998-01-01

    In 1992, the Australian Government commissioned a review into the need for a replacement research reactor. That review concluded that in about years, if certain conditions were met, the Government could make a decision in favour of a replacement reactor. A major milestone was achieved when, on 3 September 1997, the Australian Government announced the construction of a replacement research reactor at the site of Australia's existing research reactor HIFAR, subject to the satisfactory outcome of an environmental assessment process. The reactor will be have the dual purpose of providing a first class facility for neutron beam research as well as providing irradiation facilities for both medical isotope production and commercial irradiations. The project is scheduled for completion before the end of 2005. (author)

  3. International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

  4. Comparative Study on Cyber Securities between Power Reactor and Research Reactor with Bayesian Update

    International Nuclear Information System (INIS)

    Shin, Jinsoo; Heo, Gyunyoung; Son, Han Seong

    2016-01-01

    The Stuxnet has shown that nuclear facilities are no more safe from cyber-attack. Due to practical experiences and concerns on increasing of digital system application, cyber security has become the important issue in nuclear industry. Korea Institute of Nuclear Nonproliferation and control (KINAC) published a regulatory standard (KINAC/RS-015) to establish cyber security framework for nuclear facilities. However, it is difficult to research about cyber security. It is hard to quantify cyber-attack which has malicious activity which is different from existing design basis accidents (DBAs). We previously proposed a methodology on development of a cyber security risk model with BBN. However, the methodology had a limitation in which the input data as prior information was solely on expert opinions. In this study, we propose a cyber security risk model for instrumentation and control (I and C) system of nuclear facilities with some equation for quantification by using Bayesian Belief Network (BBN) in order to overcome the limitation of previous research. The proposed model has been used for comparative study on cyber securities between large-sized nuclear power plants (NPPs) and small-sized Research Reactors (RR). In this study, we proposed the cyber security risk evaluation model with BBN. It includes I and C architecture, which is a target system of cyber-attack, malicious activity, which causes cyber-attack from attacker, and mitigation measure, which mitigates the cyber-attack risk. Likelihood and consequence as prior information are evaluated by considering characteristics of I and C architecture and malicious activity. The BBN model provides posterior information with Bayesian update by adding any of assumed cyber-attack scenarios as evidence. Cyber security risk for nuclear facilities is analyzed by comparing between prior information and posterior information of each node. In this study, we conducted comparative study on cyber securities between power reactor

  5. Comparative Study on Cyber Securities between Power Reactor and Research Reactor with Bayesian Update

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jinsoo; Heo, Gyunyoung [Kyung Hee University, Yongin (Korea, Republic of); Son, Han Seong [Joongbu Univiersity, Geumsan (Korea, Republic of)

    2016-10-15

    The Stuxnet has shown that nuclear facilities are no more safe from cyber-attack. Due to practical experiences and concerns on increasing of digital system application, cyber security has become the important issue in nuclear industry. Korea Institute of Nuclear Nonproliferation and control (KINAC) published a regulatory standard (KINAC/RS-015) to establish cyber security framework for nuclear facilities. However, it is difficult to research about cyber security. It is hard to quantify cyber-attack which has malicious activity which is different from existing design basis accidents (DBAs). We previously proposed a methodology on development of a cyber security risk model with BBN. However, the methodology had a limitation in which the input data as prior information was solely on expert opinions. In this study, we propose a cyber security risk model for instrumentation and control (I and C) system of nuclear facilities with some equation for quantification by using Bayesian Belief Network (BBN) in order to overcome the limitation of previous research. The proposed model has been used for comparative study on cyber securities between large-sized nuclear power plants (NPPs) and small-sized Research Reactors (RR). In this study, we proposed the cyber security risk evaluation model with BBN. It includes I and C architecture, which is a target system of cyber-attack, malicious activity, which causes cyber-attack from attacker, and mitigation measure, which mitigates the cyber-attack risk. Likelihood and consequence as prior information are evaluated by considering characteristics of I and C architecture and malicious activity. The BBN model provides posterior information with Bayesian update by adding any of assumed cyber-attack scenarios as evidence. Cyber security risk for nuclear facilities is analyzed by comparing between prior information and posterior information of each node. In this study, we conducted comparative study on cyber securities between power reactor

  6. Research reactors: design, safety requirements and applications

    International Nuclear Information System (INIS)

    Hassan, Abobaker Mohammed Rahmtalla

    2014-09-01

    There are two types of reactors: research reactors or power reactors. The difference between the research reactor and energy reactor is that the research reactor has working temperature and fuel less than the power reactor. The research reactors cooling uses light or heavy water and also research reactors need reflector of graphite or beryllium to reduce the loss of neutrons from the reactor core. Research reactors are used for research training as well as testing of materials and the production of radioisotopes for medical uses and for industrial application. The difference is also that the research reactor smaller in terms of capacity than that of power plant. Research reactors produce radioactive isotopes are not used for energy production, the power plant generates electrical energy. In the world there are more than 284 reactor research in 56 countries, operates as source of neutron for scientific research. Among the incidents related to nuclear reactors leak radiation partial reactor which took place in three mile island nuclear near pennsylvania in 1979, due to result of the loss of control of the fission reaction, which led to the explosion emitting hug amounts of radiation. However, there was control of radiation inside the building, and so no occurred then, another accident that lead to radiation leakage similar in nuclear power plant Chernobyl in Russia in 1986, has led to deaths of 4000 people and exposing hundreds of thousands to radiation, and can continue to be effect of harmful radiation to affect future generations. (author)

  7. New research possibilities at the Budapest research reactor

    International Nuclear Information System (INIS)

    Hargitai, T.; Vidovszky, I.

    2001-01-01

    The Budapest Research Reactor is the first nuclear facility of Hungary. It was commissioned in 1959, reconstructed and upgraded in 1967 and 1986-92. The main purpose of the reactor is to serve neutron research. The reactor was extended by a liquid hydrogen type cold neutron source in 2000. The research possibilities are much improved by the CNS both in neutron scattering and neutron activation. (author)

  8. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    International Nuclear Information System (INIS)

    Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

    2014-01-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

  9. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

    2014-07-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

  10. Research and development issues for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Ando, Masanori; Morishita, Masaki

    2003-01-01

    For realization of safe and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts and validate their structural integrity, the research and development of 'Fast Reactor Structural Design Standard (FDS)' is recognized as an essential theme. FDS considers general characteristics of FRs and design needs for their rationalization. Three main subjects were settled in research and development issues of FDS. One is rationalization of failure criteria' taking characteristic design conditions into account. Next is development of 'a guideline on inelastic analysis for design' in order to predict elastic plastic and creep behaviours of high temperature components. Furthermore, efforts are being made toward preparing a guideline on thermal loads modeling' for FR component design where thermal loads are dominant. (author)

  11. Final hazard classification and auditable safety analysis for the 105-C Reactor Interim Safe Storage Project

    International Nuclear Information System (INIS)

    Rodovsky, T.J.; Larson, A.R.; Dexheimer, D.

    1996-12-01

    This document summarizes the inventories of radioactive and hazardous materials present in the 105-C Reactor Facility and the operations associated with the Interim Safe Storage Project which includes decontamination and demolition and interim safe storage of the remaining facility. This document also establishes a final hazard classification and verifies that appropriate and adequate safety functions and controls are in place to reduce or mitigate the risk associated with those operations

  12. Safety evaluation for instrumentation and control system upgrading project of Malaysian TRIGA MARK II PUSPATI Research reactor

    International Nuclear Information System (INIS)

    Ridha Roslan; Nik Mohd Faiz Khairuddin

    2013-01-01

    Full-text: Malaysian TRIGA MARK II research reactor has been in safe operation since its first criticality in 1982. The reactor is licensed to be operated by Malaysian Nuclear Agency to perform training and research development related activities. Due to its extensive operation since last three decades, the option of modifications for safety and safety-related item and component become a necessary to replace the outdated equipment to a stat-of-art, reliable technologies. This paper will present the current regulatory activities performed by Atomic Energy Licensing Board (AELB) to ensure the upgrading of analogue to digital instrumentation and control system is implemented in safe manner. The review activity includes documentation review, manufacturer quality audit and on-site inspection for commissioning. The review performed by AELB is based on The International Atomic Energy Agency (IAEA) Safety Requirements NS-R-4, entitled Safety of Research Reactors. During this endeavour, AELB seeks technical cooperation from Korea Institute of Nuclear Safety (KINS), the nuclear experts organization of the country of origin of the instrumentation and control technology. The regulatory activity is still on-going and is expected to be completed by issuance of Authorization for Restart on December 2013. (author)

  13. Gas Cooled Fast Reactor Research and Development in the European Union

    Directory of Open Access Journals (Sweden)

    Richard Stainsby

    2009-01-01

    Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

  14. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Lai Shigang; Sun Libin; Zhang Zhengming

    2013-01-01

    Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)

  15. Reactor safety research - results and perspectives

    International Nuclear Information System (INIS)

    Banaschik, M.

    1989-01-01

    The work performed so far is an essential contribution to the determination of the safety margins of nuclear facilities and their systems and to the further development of safety engineering. The further development of safety engineering involves a shift of emphasis in reactor safety research towards event sequences beyond the design basis. The aim of this shift in emphasis is the further development of the preventive level. This is based on the fact that the conservative design of the operating and safety systems involves and essential safety potential. The R and D work is intended to help develop accident management measures and to take the plant back into the safe state even after severe accidents. In this context, it is necessary to make full use of the safety margins of the plant and to include the operating systems for coping with accidents. As a result of the aims, the research work approaches operating and plant-specific processes. (orig./DG) [de

  16. Utilization of nuclear research reactors

    International Nuclear Information System (INIS)

    1980-01-01

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

  17. Studies on capacity management for factories of nuclear fuel for research reactors

    International Nuclear Information System (INIS)

    Negro, Miguel Luiz Miotto; Durazzo, Michelangelo; Mesquita, Marco Aurélio de; Carvalho, Elita Fontenele Urano de; Andrade, Delvonei Alves de; Universidade de São Paulo

    2017-01-01

    The use and the power of nuclear reactors for research and materials testing is increasing worldwide. That implies the demand for nuclear fuel for this kind of reactors is rising. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently, safely and keeping good quality. Focus is given to factories that produce plate type fuel elements loaded with LEU U_3Si_2-Al fuel, which are typically used in nuclear research reactors. Of the various production routes for this kind of fuel, we chose the route which uses hydrolysis of uranium hexafluoride. Raising the capacity of this kind of plants faces several problems, especially regarding safety against nuclear criticality. Some of these problems are briefly addressed. The new issue of the paper is the application of knowledge from the area of production administration to the fabrication of nuclear fuel for research reactors. A specific method for the increase in production capacity is proposed. That method was tested by means of discrete event simulation. The data were collected from the nuclear fuel factory at IPEN. The results indicated the proposed method achieved its goal as well as ways of raising production capacity in up to 50%. (author). (author)

  18. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    Lu, Hong; Miller, D.W.

    1991-01-01

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  19. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  20. IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci [Nuclear and Energy Research Institute, IPEN-CNEN/SP, Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242 Cid Universitaria CEP: 05508-000- Sao Paulo-SP (Brazil)

    2015-07-01

    IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

  1. The PRISM concept for a safe, economic and testable liquid metal fast reactor plant

    International Nuclear Information System (INIS)

    Berglund, R.C.; Salerno, L.N.; Tippets, F.E.

    1987-01-01

    The PRISM project is underway at General Electric as part of an advanced reactor conceptual design program sponsored by the US Department of Energy. The PRISM concept emphasizes inherent safety, modular construction, and factory fabrication. These features are intended to improve the basis for public acceptance, reduce cost,improve licensability, and reduce the risk of schedule delays and cost increases during construction. A PRISM power plant comprises a number of reactor modules. The relatively small size of the reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features for safe accommodation of accidents. These inherent safety features permit simplification and reduction of conventional safety-related systems in the plant. Testing of a full-size prototype reactor module is planned in the late 1990's to demonstrate these inherent safety characteristics. It is intended that the results of the test be used to obtain certification of the design by the US Nuclear Regulatory Commission preparatory to use of reactor modules built to this standard design in licensed commercial plants

  2. Nuclear Capacity Building through Research Reactors

    International Nuclear Information System (INIS)

    2017-01-01

    Four Instruments: •The IAEA has recently developed a specific scheme of services for Nuclear Capacity Building in support of the Member States cooperating research reactors (RR) willing to use RRs as a primary facility to develop nuclear competences as a supporting step to embark into a national nuclear programme. •The scheme is composed of four complementary instruments, each of them being targeted to specific objective and audience: Distance Training: Internet Reactor Laboratory (IRL); Basic Training: Regional Research Reactor Schools; Intermediate Training: East European Research Reactor Initiative (EERRI); Group Fellowship Course Advanced Training: International Centres based on Research Reactors (ICERR)

  3. Development of system design and seismic performance evaluation for reactor pool working platform of a research reactor

    International Nuclear Information System (INIS)

    Kwag, Shinyoung; Lee, Jong-Min; Oh, Jinho; Ryu, Jeong-Soo

    2014-01-01

    Highlights: • Design of reactor pool working platform (RPWP) is newly proposed for an open-tank-in-pool type research reactor. • Main concept of RPWP is to minimize the pool top radiation level. • Framework for seismic performance evaluation of nuclear SSCs in a deterministic and a probabilistic manner is proposed. • Structural integrity, serviceability, and seismic margin of the RPWP are evaluated during and after seismic events. -- Abstract: The reactor pool working platform (RPWP) has been newly designed for an open-tank-in-pool type research reactor, and its seismic response, structural integrity, serviceability, and seismic margin have been evaluated during and after seismic events in this paper. The main important concept of the RPWP is to minimize the pool top radiation level by physically covering the reactor pool of the open-tank-in-pool type research reactor and suppressing the rise of flow induced by the primary cooling system. It is also to provide easy handling of the irradiated objects under the pool water by providing guide tubes and refueling cover to make the radioisotopes irradiated and protect the reactor structure assembly. For this concept, the new three dimensional design model of the RPWP is established for manufacturing, installation and operation, and the analytical model is developed to analyze the seismic performance. Since it is submerged under and influenced by water, the hydrodynamic effect is taken into account by using the hydrodynamic added mass method. To investigate the dynamic characteristics of the RPWP, a modal analysis of the developed analytical model is performed. To evaluate the structural integrity and serviceability of the RPWP, the response spectrum analysis and response time history analysis have been performed under the static load and the seismic load of a safe shutdown earthquake (SSE). Their stresses are analyzed for the structural integrity. The possibility of an impact between the RPWP and the most

  4. Gas cooled fast reactor research in Europe

    International Nuclear Information System (INIS)

    Stainsby, Richard; Peers, Karen; Mitchell, Colin; Poette, Christian; Mikityuk, Konstantin; Somers, Joe

    2011-01-01

    Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce. GFR research within Europe is performed directly by those states who have signed the 'System Arrangement' document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with

  5. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

    2017-10-02

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  6. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    Dirar, H. M.

    2012-06-01

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  7. Commissioning of research reactors. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

  8. IAEA activities on research reactor safety

    International Nuclear Information System (INIS)

    Alcala-Ruiz, F.

    1995-01-01

    Since its inception in 1957, the International Atomic Energy Agency (IAEA) has included activities in its programme to address aspects of research reactors such as safety, utilization and fuel cycle considerations. These activities were based on statutory functions and responsibilities, and on the current situation of research reactors in operation around the world; they responded to IAEA Member States' general or specific demands. At present, the IAEA activities on research reactors cover the above aspects and respond to specific and current issues, amongst which safety-related are of major concern to Member States. The present IAEA Research Reactor Safety Programme (RRSP) is a response to the current situation of about 300 research reactors in operation in 59 countries around the world. (orig.)

  9. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  10. Broad-Application Test Reactor

    International Nuclear Information System (INIS)

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

  11. Technology development program for safe shipment of spent fuel from liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Freedman, J.M.; Humphreys, J.R.

    1975-10-01

    A comprehensive plan to develop shipping cask technology is described. Technical programs in the disciplines of heat transfer, structures and containment, spent fuel characterization, hot laboratory verification, shielding, and hazards analysis are discussed. Both short- and long-term goals in each discipline are delineated and how the disciplines interrelate is shown. The technologies developed will be used in the design, fabrication, and testing of truck-mounted and rail-car casks. These casks will be used for safely transporting short-cooled, high-burnup Liquid Metal Fast Breeder Reactor (LMFBR) spent fuel from reactors to reprocessing plants

  12. Da Lat Nuclear Research Reactor. Role and perspective in the development of radioisotope and nuclear technique application in Vietnam

    International Nuclear Information System (INIS)

    Tran Ha Anh; Tran Khac An; Ngo Phu Khang; Nguyen Mong Sinh

    1995-01-01

    The Da Lat Nuclear Research Reactor is playing a central role in the development of both the Nuclear Research Institute and nuclear application in our country. Thanks to this main scientific tool, the Nuclear Research Institute nearly 10 years after the completion of its renovation from the previous American-made TRIGA MARK 2 reactor is being able to implement numerous scientific and technological research projects and to develop significant applications of radioisotopes and various nuclear techniques. A general overview of the research and development activities of the Institute based on the Da Lat Nuclear Research Reactor is given as well as those aiming at ensuring its safe, reliable and efficient operation and at enlarging the perspectives of its utilisation in the future. (authors). 5 refs., 1 fig., 1 tab

  13. Design of a multipurpose research reactor

    International Nuclear Information System (INIS)

    Sanchez Rios, A.A.

    1990-01-01

    The availability of a research reactor is essential in any endeavor to improve the execution of a nuclear programme, since it is a very versatile tool which can make a decisive contribution to a country's scientific and technological development. Because of their design, however, many existing research reactors are poorly adapted to certain uses. In some nuclear research centres, especially in the advanced countries, changes have been made in the original designs or new research prototypes have been designed for specific purposes. These modifications have proven very costly and therefore beyond the reach of developing countries. For this reason, what the research institutes in such countries need is a single sufficiently versatile nuclear plant capable of meeting the requirements of a nuclear research programme at a reasonable cost. This is precisely what a multipurpose reactor does. The Mexican National Nuclear Research Institute (ININ) plans to design and build a multipurpose research reactor capable at the same time of being used for the development of reactor design skills and for testing nuclear materials and fuels, for radioisotopes production, for nuclear power studies and basic scientific research, for specialized training, and so on. For this design work on the ININ Multipurpose Research Reactor, collaborative relations have been established with various international organizations possessing experience in nuclear reactor design: Atomehnergoeksport of the USSR: Atomic Energy of Canada Limited (AECL); General Atomics (GA) of the USA; and Japan Atomic Energy Research Institute

  14. The IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    According to the research reactor database of IAEA (RRDB), 250 reactors are operating worldwide, 248 have been shut down and 170 have been decommissioned. Among the 248 reactors that do not run, some will resume their activities, others will be dismantled and the rest do not face a clear future. The analysis of reported incidents shows that the ageing process is a major cause of failures, more than two thirds of operating reactors are over 30 years old. It also appears that the lack of adequate regulations or safety standards for research reactors is an important issue concerning reactor safety particularly when reactors are facing re-starting or upgrading or modifications. The IAEA has launched a 4-axis program: 1) to set basic safety regulations and standards for research reactors, 2) to provide IAEA members with an efficient help for the application of these safety regulations to their reactors, 3) to foster international exchange of information on research reactor safety, and 4) to provide IAEA members with a help concerning safety issues linked to malicious acts or sabotage on research reactors

  15. IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    Alcala, F.; Di Meglio, A.F.

    1995-01-01

    This paper describes the IAEA programme on research reactor safety and includes the safety related areas of conversions to the use of low enriched uranium (LEU) fuel. The program is based on the IAEA statutory responsibilities as they apply to the requirements of over 320 research reactors operating around the world. The programme covers four major areas: (a) the development of safety documents; (b) safety missions to research reactor facilities; (c) support of research programmes on research reactor safety; (d) support of Technical Cooperation projects on research reactor safety issues. The demand for these activities by the IAEA member states has increased substantially in recent years especially in developing countries with increasing emphasis being placed on LEU conversion matters. In response to this demand, the IAEA has undertaken an extensive programme for each of the four areas above. (author)

  16. Regulatory aspects and experience with Russian research reactors

    International Nuclear Information System (INIS)

    Morozov, S.I.

    2003-01-01

    Regulatory activity of Gosatomnadzor of Russia in the field of research reactors (RR) safety implies implementing three major aspects: 1) establishing the nuclear and radiation safety standards; 2) licensing; and 3) inspection and enforcement. Relatively recently a full set of safety standards and regulations for RR has been established thus allowing Gosatomnadzor of Russia to effectively implement its designated functions in the field of RR safety. A minimum set of these documents is shown as follows: Level I: Fundamentals: Law 'On the use of nuclear energy'; Law 'On Public radiation protection' Level II: Safety Standard: 'General Provisions for Safety of Research Facilities' Level III: Safety Rules: Nuclear Safety; - Radiation Safety; Waste Management; Safe Decommissioning of RR; Safety Analysis Report; QAP Level IV: Safety Regulations: Licensing (incl. Peer Review and Safety Assessment) - Inspection Gosatomnadzor of Russia has created and regularly updates the database on nuclear research reactors based on the actual status of all facilities. According to the database many facilities have been shutdown during recent years whether temporary or permanently waiting for the final decision on their decommissioning. For example, in 2003 Gosatomnadzor of Russia has 85 nuclear research reactors under its supervision (compared to 113 in 1998). This fact can be explained by three main reasons: 1) experimental program finished and no other programmes in place; 2) lack of resources (financial and human); 3) safety problems (physical obsolescence and ageing of equipment). One of the main difficulties in regulating RR safety is a variety of operating organizations - 21, with different financial and human resource capabilities. Ministries responsible for supporting their operation are of a little help. It becomes obvious that a unified governmental program for RR utilization is urgently needed to decide what number of RR and for what needed purposes is required to support the

  17. Aging Management Plan for a Typical Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimi, Mahsa; Nazififard, Mohammad; Suh, Kune Y. [Seoul National University, Seoul (Korea, Republic of)

    2012-05-15

    Development of an aging management plan (AMP) is a crucial contributor to maintaining the reactor safety and controlling the risk of degradation of the concrete reactor building of a nuclear power plant. The design, operation and utilization of a research reactor (RR) fundamentally differ from those of power reactors. The AMP should nonetheless be present on account of radioactive materials and radiation risks involved. This is mainly because the RR is deemed to be used as an experiment itself or to conduct separate experiments during its operation. The AMP aims to determine the requisites for specific structural concrete components of the reactor building that entail regular inspections and maintenance to ensure safe and reliable operation of the plant. The safety of a RR necessitates the provision which is made in its design to facilitate aging management. Aging management of RR's structures is one of the vital factors to safety, to ensure continued adequacy of the safety level, reliable operation of the reactor, and compliance with the operational limits and conditions.Moreover, engineering systems should be qualified to meet the functional requirements for which they were designed with aging and environmental conditions for all situations and at all times taken into account. This study aims to present an integrated methodology for the application of an AMP for the concrete of the reactor building of a typical RR. For the purpose of safety analysis, geometry and ambient conditions were taken from a 5 MW pool-type, light-water moderated, heterogeneous, solid fuel RR in which the water is also used for cooling and shielding (Fig. 1). The reactor core is immersed in either section of a two-section concrete pool filled with water. This paper makes available background information regarding the document and the strategy developed to manage potential degradation of the reactor building concrete as well as specific programs and preventive and corrective

  18. Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl

    2010-01-01

    The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)

  19. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

    The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

  20. Nuclear data usage for research reactors

    International Nuclear Information System (INIS)

    Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

    1996-01-01

    In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)

  1. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To

  2. European Research Reactor Conference (RRFM) 2015: Conference Proceedings

    International Nuclear Information System (INIS)

    2015-01-01

    In 2015 the European Research Reactor Conference, RRFM, took place in Bucharest, Romania. The conference programme resolved around a series of plenary sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions focused on all areas of the fuel cycle of research reactors, their utilisation, operation and management as well as new research reactor projects and Innovative methods in reactor physics and thermo-hydraulics. The European Research Reactor Conference also gave special attention to safety and security of research reactors

  3. European Research Reactor Conference (RRFM) 2016: Conference Proceedings

    International Nuclear Information System (INIS)

    2016-01-01

    The 2016 European Research Reactor Conference, RRFM, took place in Berlin, Germany. The conference programme resolved around a series of plenary sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions focused on all areas of the fuel cycle of research reactors, their utilisation, operation and management as well as new research reactor projects and Innovative methods in reactor physics and thermo-hydraulics. The European Research Reactor Conference also gave special attention to safety and security of research reactors.

  4. Thermal-hydraulic modeling of flow inversion in a research reactor

    International Nuclear Information System (INIS)

    Kazeminejad, H.

    2008-01-01

    The course of loss of flow accident and flow inversion in a pool type research reactor, with scram enabled under natural circulation condition is numerically investigated. The analyses were performed by a lumped parameters approach for the coupled kinetic-thermal-hydraulics, with continuous feedback due to coolant and fuel temperature effects. A modified Runge-Kutta method was adopted for a better solution to the set of stiff differential equations. Transient thermal-hydraulics during the process of flow inversion and establishment of natural circulation were considered for a 10-MW IAEA research reactor. Some important parameters such as the peak temperatures for the hot channel were obtained for both high-enriched and low enriched fuel. The model prediction is also verified through comparison with other computer code results reported in the literature for detailed simulations of loss of flow accidents (LOFA) and the agreement between the results for the peak clad temperatures and key parameters has been satisfactory. It was found that the flow inversion and subsequent establishment of natural circulation keep the peak cladding surface temperature below the saturation temperature to avoid the escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation to ensure the safe operation of the reactor

  5. Production of Radioisotopes and Radiopharmaceuticals at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Duong Van Dong; Pham Ngoc Dien; Bui Van Cuong; Mai Phuoc Tho; Nguyen Thi Thu; Vo Thi Cam Hoa

    2014-01-01

    After reconstruction, the Dalat Nuclear Research Reactor (DNRR) was inaugurated on March 20th, 1984 with the nominal power of 500 kW. Since then the production of radioisotopes and labelled compounds for medical use was started. Up to now, DNRR is still the unique one in Vietnam. The reactor has been operated safely and effectively with the total of about 37,800 hrs (approximately 1,300 hours per year). More than 90% of its operation time and over 80% of its irradiation capacity have been exploited for research and production of radioisotopes. This paper gives an outline of the radioisotope production programme using the DNRR. The production laboratory and facilities including the nuclear reactor with its irradiation positions and characteristics, hot cells, production lines and equipment for the production of Kits for labelling with 99m Tc and for quality control, as well as the production rate are mentioned. The methods used for production of 131 I, 99m Tc, 51 Cr, 32 P, etc. and the procedures for preparation of radiopharmaceuticals are described briefly. Status of utilization of domestic radioisotopes and radiopharmaceuticals in Vietnam is also reported. (author)

  6. Backfitting of research reactors

    International Nuclear Information System (INIS)

    Delrue, R.; Noesen, T.

    1985-01-01

    The backfitting of research reactors covers a variety of activities. 1. Instrumentation and control: Control systems have developed rapidly and many reactor operators wish to replace obsolete equipment by new systems. 2. Pool liners: Some pools are lined internally with ceramic tiles. These may become pervious with time necessitating replacement, e.g. by a new stainless steel liner. 3. Heat removal system: Deficiencies can occur in one or more of the cooling system components. Upgrading may require modifications of the system such as addition of primary loops, introduction of deactivation tanks, pump replacement. Recent experience in such work has shown that renewal, backfitting and upgrading of an existing reactor is economically attractive since the related costs and delivery times are substantially lower than those required to install a new research reactor

  7. Studies on capacity management for factories of nuclear fuel for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Negro, Miguel Luiz Miotto; Durazzo, Michelangelo; Mesquita, Marco Aurélio de; Carvalho, Elita Fontenele Urano de; Andrade, Delvonei Alves de, E-mail: mlnegro@ipen.br, E-mail: mdurazzo@ipen.br, E-mail: elitaucf@ipen.br, E-mail: delvonei@ipen.br, E-mail: mamesqui@usp.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Escola Politécnica. Departamento de Engenharia de Produção

    2017-11-01

    The use and the power of nuclear reactors for research and materials testing is increasing worldwide. That implies the demand for nuclear fuel for this kind of reactors is rising. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently, safely and keeping good quality. Focus is given to factories that produce plate type fuel elements loaded with LEU U{sub 3}Si{sub 2}-Al fuel, which are typically used in nuclear research reactors. Of the various production routes for this kind of fuel, we chose the route which uses hydrolysis of uranium hexafluoride. Raising the capacity of this kind of plants faces several problems, especially regarding safety against nuclear criticality. Some of these problems are briefly addressed. The new issue of the paper is the application of knowledge from the area of production administration to the fabrication of nuclear fuel for research reactors. A specific method for the increase in production capacity is proposed. That method was tested by means of discrete event simulation. The data were collected from the nuclear fuel factory at IPEN. The results indicated the proposed method achieved its goal as well as ways of raising production capacity in up to 50%. (author). (author)

  8. A lumped parameter core dynamics model for MTR type research reactors under natural convection regime

    International Nuclear Information System (INIS)

    Ardaneh, Kazem; Zaferanlouei, Salman

    2013-01-01

    Highlights: ► A model is presented to simulate the reactivity insertion transient in MTR reactors. ► Transient dynamics of IAEA 10 MW MTR type research reactor are evaluated. ► Maximum unprotected reactivity insertion for safe condition is calculated. ► The model predictions are validated with corresponding results in the literature. - Abstract: On the basis of lumped parameter modeling of both the kinetic and thermal–hydraulic effects, a reasonably accurate simplified model has been developed to predict the dynamic response of MTR reactors following to an unprotected reactivity insertion under natural convection regime. By this model the reactor transient behavior at a given initial steady-state can be solved by a set of ordinary differential equations. The model predictions have an acceptable consent with corresponding results of reactivity insertion transients analyzed in the literature. The inherent safety characteristics of MTR research reactors utilizing natural convection is clearly demonstrated by the expanded model. The safety margin of reactor operating is selected ONB condition and thereby the proposed model determines that any slight increase in the value of $0.73 for inserted reactivity will cause the maximum cladding surface temperature to exceed the ONB condition

  9. Development of a protection system for research reactor based in Field Programmable Gate Array - FPGA

    International Nuclear Information System (INIS)

    Martins, Roque Hudson da Silva

    2016-01-01

    This study presents a implementation purpose of a protection system for research nuclear reactors by using a programed device FPGA (Field Programmable Gate Array). As well as logic protection method involved on an automatic shutdown (TRIP) of a reactor, that ensure the security on such systems. These new control and operation mechanics are developed to guarantee that the security limits of a power plant are not exceeded, these mechanics can work isolated or in groups to safe guard the security levels. For this implementation to be completed, there will be presented the main aspects and concepts referred to protection systems, mostly about research nuclear reactors, with some applications terms exposed. The system proposed at this paper was developed following the VHDL (Very High Speed Integrated Circuits) hardware describing language, and the Modelsim software from Altera Software to program the automatic turning off routines, and hypothetical simulations for such. The results show that for every software application for supporting nuclear reactors, like security devices, they have to meet the IEC 60880 criteria. This paper have great importance, seeing that nuclear reactor security systems, are a basic element for ensure the reactor security. (author)

  10. The role of research reactor and its future

    International Nuclear Information System (INIS)

    Nakagome, Yoshihiro

    2005-01-01

    About a half century passed since the start of operation of research reactors. Many research reactors were stopped their operation or decommissioned. With the practical use of nuclear energy, the meaning of research reactor has been buried in oblivion in the developed countries. Furthermore, under the nuclear weapons nonproliferation policy, the use of high enriched uranium fuel in research reactors is obliged to change to the use of low enriched uranium fuel. In such severe situation, this paper refers to the role of the research reactor once more through the operation experience of university-owned research reactor KUR (Kyoto University Reactor, Japan) and describes that research reactor is indispensable for the preparation to the second coming nuclear age. (author)

  11. 'Experience with decommissioning of research and test reactors at Argonne National Laboratory'

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Yule, T.J.; Fellhauer, C.R.; Boing, L.E.

    2002-01-01

    A large number of research reactors around the world have reached the end of their useful operational life. Many of these are kept in a controlled storage mode awaiting decontamination and decommissioning (D and D). At Argonne National Laboratory located near Chicago in the United States of America, significant experience has been gained in the D and D of research and test reactors. These experiences span the entire range of activities in D and D - from planning and characterization of the facilities to the eventual disposition of all waste. A multifaceted D nd D program has been in progress at the Argonne National Laboratory - East site for nearly a decade. The program consists of three elements: - D and D of nuclear facilities on the site that have reached the end of their useful life; - Development and demonstrations of technologies that help in safe and cost effective D and D; - Presentation of training courses in D and D practices. Nuclear reactor facilities have been constructed and operated at the ANL-E site since the earliest days of nuclear power. As a result, a number of these early reactors reached end-of-life long before reactors on other sites and were ready for D and D earlier. They presented an excellent set of test beds on which D and D practices and technologies could be demonstrated in environments that were similar to commercial reactors, but considerably less hazardous. As shown, four reactor facilities, plutonium contaminated glove boxes and hot cells, a cyclotron facility and assorted other nuclear related facilities have been decommissioned in this program. The overall cost of the program has been modest relative to the cost of comparable projects undertaken both in the U.S. and abroad. The safety record throughout the program was excellent. Complementing the actual operations, a set of D and D technologies are being developed. These include robotic methods of tool handling and operation, chemical and laser decontamination techniques, sensors

  12. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1975-01-01

    Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

  13. Utilization of research reactors - A global perspective

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1988-01-01

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

  14. SIR (Safe Integral Reactor) - reducing size can reduce cost

    International Nuclear Information System (INIS)

    Hayns, M.R.

    1991-01-01

    Traditional engineering economics have favoured the advantages of larger size as a means of reducing specific capital costs and hence unit generating costs. For large and small plants utilising the same concept, e.g. a small four-loop PWR vs a large four-loop PWR with the same number of components, economies of scale are well established. If, however, a smaller plant is sized to take advantage of features which are only feasible at smaller outputs, is of simpler design, with the advantage taken of the simplified design to produce the most cost-effective layout, and incorporates fewer, more easily replaceable components with minimal assembly on site, it is possible to produce a plant which is competitive with larger plant of more traditional design. When 'system' effects, such as better matching of installed capacity to the growth in demand and the fact that a smaller total capacity will be needed to meet a given demand with a specified level of confidence, are taken into account, it can be shown that a utility's overall cash-flow position can be improved with lower associated absolute financial risks. The UK/US Safe Integral Reactor (SIR) is an integral pressurized water reactor in the 300-400 MW(e) range which utilises conventional water reactor technology in a way not feasible at the very large, sizes of recent years. The SIR concept is briefly explained and its technical and economic advantages in terms of simplicity, construction, maintenance, availability, decommissioning, safety and siting described. The results of system analyses which demonstrate the overall financial advantages to a utility are presented. (author)

  15. The market for research reactors

    International Nuclear Information System (INIS)

    Roegler, H.J.

    1986-01-01

    The assay deals with some basic questions if there is an international market for research reactors at all, which influencing factors affect this market, and if research reactors have any effects on the future market for nuclear engineering. (UA) [de

  16. Research reactor job analysis - A project description

    International Nuclear Information System (INIS)

    Yoder, John; Bessler, Nancy J.

    1988-01-01

    Addressing the need of the improved training in nuclear industry, nuclear utilities established training program guidelines based on Performance-Based Training (PBT) concepts. The comparison of commercial nuclear power facilities with research and test reactors owned by the U.S. Department of Energy (DOE), made in an independent review of personnel selection, training, and qualification requirements for DOE-owned reactors pointed out that the complexity of the most critical tasks in research reactors is less than that in power reactors. The U.S. Department of Energy (DOE) started a project by commissioning Oak Ridge Associated Universities (ORAU) to conduct a job analysis survey of representative research reactor facilities. The output of the project consists of two publications: Volume 1 - Research Reactor Job Analysis: Overview, which contains an Introduction, Project Description, Project Methodology,, and. An Overview of Performance-Based Training (PBT); and Volume 2 - Research Reactor Job Analysis: Implementation, which contains Guidelines for Application of Preliminary Task Lists and Preliminary Task Lists for Reactor Operators and Supervisory Reactor Operators

  17. Guidelines for the review research reactor safety. Reference document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

    International Nuclear Information System (INIS)

    1997-01-01

    In 1992, the IAEA published new safety standards for research reactors as part of the set of publications considered by its Research Reactor Safety Programme (RRSP). This set also includes publications giving guidance for all safety aspects related to the lifetime of a research reactor. In addition, the IAEA has also revised the Safety Standards for radiation protection. Consequently, it was considered advisable to revise the Integrated Safety Assessment of Research Reactors (INSARR) procedures to incorporate the new requirements and guidance as well as to extend the scope of the safety reviews to currently operating research reactors. The present report is the result of this revision. The purpose of this report is to give guidance on the preparation, execution, reporting and follow-up of safety review mission to research reactors as conducted by the IAEA under its INSARR missions safety service. However, it will also be of assistance to operators and regulators in conducting: (a) ad hoc safety assessments of research reactors to address individual issues such as ageing or safety culture; and (b) other types of safety reviews such as internal and peer reviews and regulatory inspections

  18. Research reactor preparations for the air shipment of highly enriched uranium from Romania

    International Nuclear Information System (INIS)

    Bolshinsky, I.; Allen, K.J.; Biro, L.L.; Budu, M.E.; Zamfir, N.V.; Dragusin, M.; Paunoiu, C.; Ciocanescu, M.

    2010-01-01

    In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation (RF) for conversion to low enriched uranium (LEU). The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR-S research reactor at Magurele, Romania, to Ozersk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation for Atomic Energy Rosatom and the International Atomic Energy Agency (IAEA). Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel. (author)

  19. Effect of reactivity insertion rate on peak power and temperatures in swimming pool type research reactor

    International Nuclear Information System (INIS)

    Khan, L.A.; Jabbar, A.; Anwar, A.R.; Ahmad, N.

    1998-01-01

    It is essential to study the reactor behavior under different accidental conditions and take proper measures for its safe operation. We have studied the effect of reactivity insertion, with and without scram conditions, on peak power and temperatures of fuel, cladding and coolant in typical swimming pool type research reactor. The reactivity ranging from 1 $ to 2 $ and insertion times from 0.25 to 1 second have been considered. The computer code PARET has been used and results are presented in this article. (author)

  20. Design and construction of multi research reactor

    International Nuclear Information System (INIS)

    1985-05-01

    This is the report about design and construction of multi research reactor, which introduces the purpose and necessity of the project, business contents, plan of progress of project and budget for the project. There are three appendixes about status of research reactor in other country, a characteristic of research reactor, three charts about evaluation, process and budget for the multi research reactor and three drawings for the project.

  1. Guidelines for the Review of Research Reactor Safety: Revised Edition. Reference Document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

    International Nuclear Information System (INIS)

    2013-01-01

    The Integrated Safety Assessment of Research Reactors (INSARR) is an IAEA safety review service available to Member States with the objective of supporting them in ensuring and enhancing the safety of their research reactors. This service consists of performing a comprehensive peer review and an assessment of the safety of the respective research reactor. The reviews are based on IAEA safety standards and on the provisions of the Code of Conduct on the Safety of Research Reactors. The INSARR can benefit both the operating organizations and the regulatory bodies of the requesting Member States, and can include new research reactors under design or operating research reactors, including those which are under a Project and Supply Agreement with the IAEA. The first IAEA safety evaluation of a research reactor operated by a Member State was completed in October 1959 and involved the Swiss 20 MW DIORIT research reactor. Since then, and in accordance with its programme on research reactor safety, the IAEA has conducted safety review missions in its Member States to enhance the safety of their research reactor facilities through the application of the Code of Conduct on the Safety of Research Reactors and the relevant IAEA safety standards. About 320 missions in 51 Member States were undertaken between 1972 and 2012. The INSARR missions and other limited scope safety review missions are conducted following the guidelines presented in this publication, which is a revision of Guidelines for the Review of Research Reactor Safety (IAEA Services Series No. 1), published in December 1997. This publication details those IAEA safety standards and guidance publications relevant to the safety of research reactors that have been revised or published since 1997. The purpose of this publication is to give guidance on the preparation, implementation, reporting and follow-up of safety review missions. It is also intended to be of assistance to operators and regulators in conducting

  2. Safety of research reactors - A regulator's perspective

    International Nuclear Information System (INIS)

    Rahman, M.S.

    2001-01-01

    Due to historical reasons research reactors have received less regulatory attention in the world than nuclear power plants. This has given rise to several safety issues which, if not addressed immediately, may result in an undesirable situation. However, in Pakistan, research reactors and power reactors have received due attention from the regulatory authority. The Pakistan Research Reactor-1 has been under regulatory surveillance since 1965, the year of its commissioning. The second reactor has also undergone all the safety reviews and checks mandated by the licensing procedures. A brief description of the regulatory framework, the several safety reviews carried out have been briefly described in this paper. Significant activities of the regulatory authority have also been described in verifying the safety of research reactors in Pakistan along with the future activities. The views of the Pakistani regulatory authority on the specific issues identified by the IAEA have been presented along with specific recommendations to the IAEA. We are of the opinion that there are more Member States operating nuclear research reactors than nuclear power plants. Therefore, there should be more emphasis on the research reactor safety, which somehow has not been the case. In several recommendations made to the IAEA on the specific safety issues the emphasis has been, in general, to have a similar documentation and approach for maintaining and verifying operational safety at research reactors as is currently available for nuclear power reactors and may be planned for nuclear fuel cycle facilities. (author)

  3. Fuel burnup analysis for the Moroccan TRIGA research reactor

    International Nuclear Information System (INIS)

    El Bakkari, B.; El Bardouni, T.; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Zoubair, M.

    2013-01-01

    Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of

  4. Nuclear reactor safety research in Idaho

    International Nuclear Information System (INIS)

    Zeile, H.J.

    1983-01-01

    Detailed information about the performance of nuclear reactor systems, and especially about the nuclear fuel, is vital in determining the consequences of a reactor accident. Fission products released from the fuel during accidents are the ultimate safety concern to the general public living in the vicinity of a nuclear reactor plant. Safety research conducted at the Idaho National Engineering Laboratory (INEL) in support of the U.S. Nuclear Regulatory Commission (NRC) has provided the NRC with detailed data relating to most of the postulated nuclear reactor accidents. Engineers and scientists at the INEL are now in the process of gathering data related to the most severe nuclear reactor accident - the core melt accident. This paper describes the focus of the nuclear reactor safety research at the INEL. The key results expected from the severe core damage safety research program are discussed

  5. Physics and safety of advanced research reactors

    International Nuclear Information System (INIS)

    Boening, K.; Hardt, P. von der

    1987-01-01

    Advanced research reactor concepts are presently being developed in order to meet the neutron-based research needs of the nineties. Among these research reactors, which are characterized by an average power density of 1-10 MW per liter, highest priority is now generally given to the 'beam tube reactors'. These provide very high values of the thermal neutron flux (10 14 -10 16 cm -2 s -1 ) in a large volume outside of the reactor core, which can be used for sample irradiations and, in particular, for neutron scattering experiments. The paper first discusses the 'inverse flux trap concept' and the main physical aspects of the design and optimization of beam tube reactors. After that two examples of advanced research reactor projects are described which may be considered as two opposite extremes with respect to the physical optimization principle just mentioned. The present situation concerning cross section libraries and neutronic computer codes is more or less satisfactory. The safety analyses of advanced research reactors can largely be updated from those of current new designs, partially taking advantage of the immense volume of work done for power reactors. The paper indicates a few areas where generic problems for advanced research reactor safety are to be solved. (orig.)

  6. The future role of research reactors

    International Nuclear Information System (INIS)

    Glaeser, W.

    2001-01-01

    The decline of neutron source capacity in the next decades urges for the planning and construction of new neutron sources for basic and applied research with neutrons. Modern safety precautions of research reactors make them competitive with other ways of neutron production using non-chain reactions for many applications. Research reactors consequently optimized offer a very broad range of possible applications in basic and applied research. Research reactors at universities also in the future have to play an important role in education and training in basic and applied nuclear science. (orig.)

  7. IRSN research programs concerning reactor safety

    International Nuclear Information System (INIS)

    Bardelay, J.

    2005-01-01

    This paper is made up of 3 parts. The first part briefly presents the missions of IRSN (French research institute on nuclear safety), the second part reviews the research works currently led by IRSN in the following fields : -) the assessment of safety computer codes, -) thermohydraulics, -) reactor ageing, -) reactivity accidents, -) loss of coolant, -) reactor pool dewatering, -) core meltdown, -) vapor explosion, and -) fission product release. In the third part, IRSN is shown to give a major importance to experimental programs led on research or test reactors for collecting valid data because of the complexity of the physical processes that are involved. IRSN plans to develop a research program concerning the safety of high or very high temperature reactors. (A.C.)

  8. The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    Science.gov (United States)

    Cole, Christopher J. P.

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas

  9. The near boiling reactor : conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    International Nuclear Information System (INIS)

    Cole, C.J.P.

    2005-01-01

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the 'Victoria' Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96 o C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional

  10. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  11. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2002-01-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

  12. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.

    1999-01-01

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  13. Licensing activities for the partial decommissioning of IRT-2000 research reactor in Sofia

    International Nuclear Information System (INIS)

    Apostolov, T.; Ilieva, Kr.; Papukchiev, A.; Kalchev, B.

    2001-01-01

    The project for refurbishment of IRT-2000 research reactor in Sofia into low-power reactor (200 kW) is based on the retention of some IRT-2000 buildings, facilities and equipment. The activities, which determine the partial decommissioning should be realized in accordance with preliminary developed licensing documents as General Plan, Safety Analysis Report and Environment Impact assessment Report. The goal of these documents is to provide and guarantee safe and effective activities with radioactive materials, to define strictly the dismantling procedures, and in the same time to minimize their influence on the environment. The Technical Tasks for General Plan, Safety Analysis Report and Environment Impact Assessment Report have been prepared and will be presented as preliminary licensing documents to the National Regulatory Body for approval before their application. A Quality Management system is being developed nowadays at INRNE. After its certification some requirements of the regulatory body will be completed. This certified QA system is a major part of the licensing procedure for the reconstruction of IRT-2000 research reactor. (author)

  14. Introduction to Safety Analysis Approach for Research Reactors

    International Nuclear Information System (INIS)

    Park, Suki

    2016-01-01

    The research reactors have a wide variety in terms of thermal powers, coolants, moderators, reflectors, fuels, reactor tanks and pools, flow direction in the core, and the operating pressure and temperature of the cooling system. Around 110 research reactors have a thermal power greater than 1 MW. This paper introduces a general approach to safety analysis for research reactors and deals with the experience of safety analysis on a 10 MW research reactor with an open-pool and open-tank reactor and a downward flow in the reactor core during normal operation. The general approach to safety analysis for research reactors is described and the design features of a typical open-pool and open-tank type reactor are discussed. The representative events expected in research reactors are investigated. The reactor responses and the thermal hydraulic behavior to the events are presented and discussed. From the minimum CHFR and the maximum fuel temperature calculated, it is ensured that the fuel is not damaged in the step insertion of reactivity by 1.8 mk and the failure of all primary pumps for the reactor with a 10 MW thermal power and downward core flow

  15. New research reactor proposed for Australia

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    A new research reactor has been proposed for construction within the next ten years, to replace the HIFAR reactor which operating capabilities have been over taken by later designs. This paper outlines the main research applications of the new reactor design and briefly examines issues related to its cost, economic benefits, safety and location

  16. Preliminary risk assessment of the Integral Inherently-Safe Light Water Reactor

    International Nuclear Information System (INIS)

    McCarroll, Kellen R.; Lee, John C.; Manera, Annalisa; Memmott, Matthew J.; Ferroni, Paolo

    2017-01-01

    The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) concept seeks to significantly increase nuclear power plant safety. The project implements a safety-by-design philosophy, eliminating several initiating events and providing novel, passive safety systems at the conceptual phase. Pursuit of unparalleled safety employs an integrated development process linking design with deterministic and probabilistic safety analyses. Unique aspects of the I 2 S-LWR concept and design process present challenges to the probabilistic risk assessment (PRA), particularly regarding overall flexibility, auditability and resolution of results. Useful approaches to initiating events and conditional failures are presented. To exemplify the risk-informed design process using PRA, a trade-off study of two safety system configurations is presented. Although further optimization is required, preliminary results indicate that the I 2 S-LWR can achieve a core damage frequency (CDF) from internal events less than 1.01 × 10 −8 /ry, including reactor vessel ruptures. Containment bypass frequency due to primary heat exchanger rupture is found to be comparable to non-vessel rupture CDF.

  17. Research reactor modernization and refurbishment

    International Nuclear Information System (INIS)

    2009-08-01

    Many recent, high profile research reactor unplanned shutdowns can be directly linked to different challenges which have evolved over time. The concept of ageing management is certainly nothing new to nuclear facilities, however, these events are highlighting the direct impact unplanned shutdowns at research reactors have on various stakeholders who depend on research reactor goods and services. Provided the demand for these goods and services remains strong, large capital projects are anticipated to continue in order to sustain future operation of many research reactors. It is within this context that the IAEA organized a Technical Workshop to launch a broader Agency activity on research reactor modernization and refurbishment (M and R). The workshop was hosted by the operating organization of the HOR Research Reactor in Delft, the Netherlands, in October 2006. Forty participants from twenty-three countries participated in the meeting: with representation from Africa, Asia Pacific, Eastern Europe, North America, South America and Western Europe. The specific objectives of this workshop were to present facility reports on completed, existing and planned M and R projects, including the project objectives, scope and main characteristics; and to specifically report on: - the project impact (planned or actual) on the primary and key supporting motivation for the M and R project; - the project impact (planned or actual) on the design basis, safety, and/or regulatory-related reports; - the project impact (planned or actual) on facility utilization; - significant lessons learned during or following the completion of M and R work. Contributions from this workshop were reviewed by experts during a consultancy meeting held in Vienna in December 2007. The experts selected final contributions for inclusion in this report. Requests were also distributed to some authors for additional detail as well as new authors for known projects not submitted during the initial 2006 workshop

  18. Light water reactor safety research project

    International Nuclear Information System (INIS)

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

    1980-07-01

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

  19. Reactor materials research as an effective instrument of nuclear reactor perfection

    International Nuclear Information System (INIS)

    Baryshnikov, M.

    2006-01-01

    The work is devoted to reactor materiology, as to the practical tool of nuclear reactor development. The work is illustrated with concrete examples from activity experience of the appropriate division of the Russian Research Centre Kurchatov Institute - Institute of Reactor Materials Research and Radiation Nanotechnologies. Besides the description of some modern potentials of the mentioned institute is given. (author)

  20. Experience in using a research reactor for the training of power reactor operators

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Arsenaut, L.J.

    1972-01-01

    A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

  1. Utilization of research reactors

    International Nuclear Information System (INIS)

    1962-01-01

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

  2. RETU. The Finnish research programme on reactor safety. Interim report 1995 - May 1997

    International Nuclear Information System (INIS)

    Vanttola, T.; Puska, E.K.

    1997-08-01

    The Finnish national research programme on Reactor Safety (RETU, 1995-1998) concentrates on the search of safe limits of nuclear fuel and the reactor core, accident management methods and risk management of the operation of nuclear power plants. The annual volume of the programme has been about 26 person years and the annual funding FIM 15 million. This report summarises the structure and objectives of the programme, research fields included and the main results obtained during the period 1995 - May 1997. In the field of operational margins of a nuclear reactor, the behaviour of high burnup nuclear fuel is studied both in normal operation and during power transients. The static and dynamic reactor analysis codes are developed and validated to cope with new fuel designs and complicated three-dimensional reactivity transients and accidents. Research on accident management aims at development and validation of calculation methods needed to plan preventive measures and to train the personnel to severe accident mitigation. Other goals are to reduce uncertainties in phenomena important in severe accidents and to study actions planned for accident management. In the field of risk management probabilistic methods are developed for safety related decision making and for complex phenomena and event sequences. Effects of maintenance on nuclear power plant safety are studied and more effective methods for the assessment of human reliability and safety critical organisations are searched

  3. RETU. The Finnish research programme on reactor safety. Interim report 1995 - May 1997

    Energy Technology Data Exchange (ETDEWEB)

    Vanttola, T; Puska, E K [VTT Energy, Espoo (Finland). Nuclear Energy; eds.

    1997-08-01

    The Finnish national research programme on Reactor Safety (RETU, 1995-1998) concentrates on the search of safe limits of nuclear fuel and the reactor core, accident management methods and risk management of the operation of nuclear power plants. The annual volume of the programme has been about 26 person years and the annual funding FIM 15 million. This report summarises the structure and objectives of the programme, research fields included and the main results obtained during the period 1995 - May 1997. In the field of operational margins of a nuclear reactor, the behaviour of high burnup nuclear fuel is studied both in normal operation and during power transients. The static and dynamic reactor analysis codes are developed and validated to cope with new fuel designs and complicated three-dimensional reactivity transients and accidents. Research on accident management aims at development and validation of calculation methods needed to plan preventive measures and to train the personnel to severe accident mitigation. Other goals are to reduce uncertainties in phenomena important in severe accidents and to study actions planned for accident management. In the field of risk management probabilistic methods are developed for safety related decision making and for complex phenomena and event sequences. Effects of maintenance on nuclear power plant safety are studied and more effective methods for the assessment of human reliability and safety critical organisations are searched. 135 refs.

  4. Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2011-01-01

    Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

  5. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

    1974-07-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  6. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

    1974-01-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  7. On the research activities in reactor and neutron physics using the first egyptian research reactor

    International Nuclear Information System (INIS)

    Hassan, A.M.

    2000-01-01

    A review on the most important research activities in reactor and neutron physics using the first Egyptian Research Reactor (ET-RR-1) is given. An out look on: neutron cross-sections, neutron flux, neutron capture gamma-ray spectroscopy, neutron activation analysis, neutron diffraction and radiation shielding experiments, is presented

  8. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2011-01-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  9. Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2015-01-01

    Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

  10. Current status and prospects of research reactors

    International Nuclear Information System (INIS)

    Gabaraev, A.B.; Cherepnin, Yu.S.; Tretyakov, I.T.; Khmelshikov, V.V.; Dollezhal, N.A.

    2009-01-01

    Full text: The first nuclear research reactors (RR) appeared in the 1940s. Their initial purpose was to provide knowledge of the main processes associated with neutron-induced nuclear reactions. Later, the rang of problems addressed expanded substantially. Besides fundamental research in the properties of matter, such reactors are successfully used for dealing with problems in the fields of materials science, nuclear engineering, medicine, isotope production, education, etc. Over the whole period of RR fleet growth, more than six hundred nuclear research facilities were built in 70 countries of the world. As of the end of 2008, the number of Russian research reactors in service was about 20% of the globally operating RR fleet. This paper discusses the current status of the world's RR fleet and describes the capabilities of the experimental reactor facilities existing in Russia. In the 21st century, research reactors will remain in demand to solve scientific and technological problems for innovative development of society. The emerging renaissance of nuclear power, the expanding RR uses for production of isotopes and other applications, the increase in the number of countries willing to use nuclear technologies in energy production, industry and science - all contribute to a rebirth of interest in research reactors. One of the ways to improve the experimental capabilities lies in radical upgrading of the reactor facilities with qualitative changes in the main neutronic characteristics of the core. The associated design approaches are illustrated with the example of the IBR-2M reactor at the JNRI in Dubna. The imperative need restricting the spread of nuclear threat leads us to give up using highly enriched uranium in most research reactors. Development of RR fuel with reduced enrichment in uranium has been one of the priority objectives of NIKIET for many years. This paper presents the latest results obtained along these lines, as applied to pool-type research

  11. A new advanced safe nuclear reactor concept

    International Nuclear Information System (INIS)

    Sefidvash, Farhang

    1999-01-01

    The reactor design is based on fluidized bed concept and utilizes pressurized water reactor technology. The fuel is automatically removed from the reactor by gravity under any accident condition. The reactor demonstrates the characteristics of inherent safety and passive cooling. Here two options for modification to the original design are proposed in order to increase the stability and thermal efficiency of the reactor. A modified version of the reactor involves the choice of supercritical steam as the coolant to produce a plant thermal efficiency of about 40%. Another is to modify the shape of the reactor core to produce a non-fluctuating bed and consequently guarantee the dynamic stability of the reactor. The mixing of Tantalum in the fuel is also proposed as an additional inhibition to power excursion. The spent fuel pellets may not be considered nuclear waste since they are in the shape and size that can easily be used as a a radioactive source for food irradiation and industrial applications. The reactor can easily operate with any desired spectrum by varying the porosity in order to be a plutonium burner or utilize a thorium fuel cycle. (author)

  12. Research reactor safety - an overview of crucial aspects

    International Nuclear Information System (INIS)

    Laverie, M.

    1998-01-01

    Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

  13. Meeting on reactor safety research

    International Nuclear Information System (INIS)

    1982-09-01

    The meeting 'Reactor Safety Research' organized for the second time by the GRS by order of the BMFT gave a review of research activities on the safety of light water reactors in the Federal Repulbic of Germany, international co-operation in this field and latest results of this research institution. The central fields of interest were subjects of man/machine-interaction, operational reliability accident sequences, and risk. (orig.) [de

  14. Nuclear research reactors in the world. June 1988 ed.

    International Nuclear Information System (INIS)

    1988-01-01

    This is the third edition of Reference Data Series No. 3, Nuclear Research Reactors in the World, which replaces the Agency's publications Power and Research Reactors in Member States and Research Reactors in Member States. This booklet contains general information, as of the end of June 1988, on research reactors in operation, under construction, planned, and shut down. The information is collected by the Agency through questionnaires sent to the Member States through the designated national correspondents. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the IAEA Research Reactor Data Base (RRDB) system. This system contains all the information and data previously published in the Agency's publication Power and Research Reactors in Member States as well as additional information. 12 figs, 19 tabs

  15. Current status of the world's research reactors

    International Nuclear Information System (INIS)

    Dodd, B.

    1999-01-01

    Data from the IAEA's Research Reactor Database (RRDB) provides information with respect to the status of the world's research reactors. Some summary data are given. Recent initiatives by the IAEA regarding communications and information flow with respect to research reactors are discussed. Future plans and perspectives are also introduced. (author)

  16. Research reactor records in the INIS database

    International Nuclear Information System (INIS)

    Marinkovic, N.

    2001-01-01

    This report presents a statistical analysis of more than 13,000 records of publications concerned with research and technology in the field of research and experimental reactors which are included in the INIS Bibliographic Database for the period from 1970 to 2001. The main objectives of this bibliometric study were: to make an inventory of research reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of research reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in research reactors research and technology. (author)

  17. Effective utilization and management of research reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-06-01

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

  18. Effective utilization and management of research reactors

    International Nuclear Information System (INIS)

    Muranaka, R.

    1984-01-01

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

  19. A continuing success - The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Mustin, Tracy P.; Clapper, Maureen; Reilly, Jill E.

    2000-01-01

    The United States Department of Energy, in consultation with the Department of State, adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program, established under this policy, has completed 16 spent fuel shipments. 2,651 material test reactor (MTR) assemblies, one Slowpoke core containing less than 1 kilogram of U.S.-origin enriched uranium, 824 Training, Research, Isotope, General Atomic (TRIGA) rods, and 267 TRIGA pins from research reactors around the world have been shipped to the United States so far under this program. As the FRR SNF Acceptance Program progresses into the fifth year of implementation, a second U.S. cross country shipment has been completed, as well as a second overland truck shipment from Canada. Both the cross country shipment and the Canadian shipment were safely and successfully completed, increasing our knowledge and experience in these types of shipments. In addition, two other shipments were completed since last year's RERTR meeting. Other program activities since the last meeting included: taking pre-emptive steps to avoid license amendment pitfalls/showstoppers for spent fuel casks, publication of a revision to the Record of Decision allowing up to 16 casks per ocean going vessel, and the issuance of a cable to 16 of the 41 eligible countries reminding their governments and the reactor operators that the U.S.-origin uranium in their research reactors may be eligible for return to the United States under the Acceptance Program and urging them to begin discussions on shipping schedules. The FRR SNF program has also supported the Department's implementation of the competitive pricing policy for uranium and resumption of shipments of fresh uranium for fabrication into assemblies for research reactors. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues

  20. Present status and future prospect of research reactors

    International Nuclear Information System (INIS)

    Takemi, Hirokatsu

    1996-01-01

    The present status of research reactors more than MW class reactor in JAERI and the Kyoto University and the small reactors in the Musashi Institute of Technology, the Rikkyo University, the Tokyo University, the Kinki University and other countries are explained in the paper. The present status of researches are reported by the topics in each field. The future researches of the beam reactor and the irradiation reactor are reviewed. On various kinds of use of research reactor and demands of neutron field of a high order, new type research reactors under investigation are explained. Recently, the reactors are used in many fields such as the basic science: the basic physics, the material science, the nuclear physics, and the nuclear chemistry and the applied science; the earth and environmental science, the biology and the medical science. (S.Y.)

  1. Overview of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Nguyen Thai Sinh; Luong Ba Vien

    2016-01-01

    The present reactor called Dalat Nuclear Research Reactor (DNRR) has been reconstructed from the former TRIGA Mark II reactor which was designed by General Atomic (GA, San Diego, California, USA), started building in early 1960s, put into operation in 1963 and operated until 1968 at nominal power of 250 kW. In 1975, all fuel elements of the reactor were unloaded and shipped back to the USA. The DNRR is a 500-kW pool-type research reactor using light water as both moderator and coolant. The reactor is used as a neutron source for the purposes of: (1) radioactive isotope production; (2) neutron activation analysis; and (3) research and training

  2. Research reactor safety - an overview of crucial aspects

    Energy Technology Data Exchange (ETDEWEB)

    Laverie, M. [Atomic Energy Commission, Saclay, F-91191 Gif sur Yvette (France)

    1998-07-01

    Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

  3. New research reactor for Australia

    International Nuclear Information System (INIS)

    Miller, R.

    1992-01-01

    HIFAR, Australia's major research reactor was commissioned in 1958 to test materials for an envisaged indigenous nuclear power industry. HIFAR is a Dido type reactor which is operated at 10 MW. With the decision in the early 1970's not to proceed to nuclear power, HIFAR was adapted to other uses and has served Australia well as a base for national nuclear competence; as a national facility for neutron scattering/beam research; as a source of radioisotopes for medical diagnosis and treatment; and as a source of export revenue from the neutron transmutation doping of silicon for the semiconductor industry. However, all of HIFAR's capabilities are becoming less than optimum by world and regional standards. Neutron beam facilities have been overtaken on the world scene by research reactors with increased neutron fluxes, cold sources, and improved beams and neutron guides. Radioisotope production capabilities, while adequate to meet Australia's needs, cannot be easily expanded to tap the growing world market in radiopharmaceuticals. Similarly, neutron transmutation doped silicon production, and export income from it, is limited at a time when the world market for this material is expanding. ANSTO has therefore embarked on a program to replace HIFAR with a new multi-purpose national facility for nuclear research and technology in the form of a reactor: a) for neutron beam research, - with a peak thermal flux of the order of three times higher than that from HIFAR, - with a cold neutron source, guides and beam hall, b) that has radioisotope production facilities that are as good as, or better than, those in HIFAR, c) that maximizes the potential for commercial irradiations to offset facility operating costs, d) that maximizes flexibility to accommodate variations in user requirements during the life of the facility. ANSTO's case for the new research reactor received significant support earlier this month with the tabling in Parliament of a report by the Australian Science

  4. Experience in utilizing research reactors in Yugoslavia

    Energy Technology Data Exchange (ETDEWEB)

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

    1972-07-01

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

  5. Experience in utilizing research reactors in Yugoslavia

    International Nuclear Information System (INIS)

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

    1972-01-01

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

  6. An overview-probabilistic safety analysis for research reactors

    International Nuclear Information System (INIS)

    Liu Jinlin; Peng Changhong

    2015-01-01

    For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, 'Nuclear safety and radioactive pollution prevention 'Twelfth Five Year Plan' and the 2020 vision' raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs. (author)

  7. Research reactor fuel - an update

    International Nuclear Information System (INIS)

    Finlay, M.R.; Ripley, M.I.

    2003-01-01

    In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

  8. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

    The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

  9. A Review of Previous Research in Direct Energy Conversion Fission Reactors

    International Nuclear Information System (INIS)

    DUONG, HENRY; POLANSKY, GARY F.; SANDERS, THOMAS L.; SIEGEL, MALCOLM D.

    1999-01-01

    From the earliest days of power reactor development, direct energy conversion was an obvious choice to produce high efficiency electric power generation. Directly capturing the energy of the fission fragments produced during nuclear fission avoids the intermediate conversion to thermal energy and the efficiency limitations of classical thermodynamics. Efficiencies of more than 80% are possible, independent of operational temperature. Direct energy conversion fission reactors would possess a number of unique characteristics that would make them very attractive for commercial power generation. These reactors would be modular in design with integral power conversion and operate at low pressures and temperatures. They would operate at high efficiency and produce power well suited for long distance transmission. They would feature large safety margins and passively safe design. Ideally suited to production by advanced manufacturing techniques, direct energy conversion fission reactors could be produced more economically than conventional reactor designs. The history of direct energy conversion can be considered as dating back to 1913 when Moseleyl demonstrated that charged particle emission could be used to buildup a voltage. Soon after the successful operation of a nuclear reactor, E.P. Wigner suggested the use of fission fragments for direct energy conversion. Over a decade after Wigner's suggestion, the first theoretical treatment of the conversion of fission fragment kinetic energy into electrical potential appeared in the literature. Over the ten years that followed, a number of researchers investigated various aspects of fission fragment direct energy conversion. Experiments were performed that validated the basic physics of the concept, but a variety of technical challenges limited the efficiencies that were achieved. Most research in direct energy conversion ceased in the US by the late 1960s. Sporadic interest in the concept appears in the literature until this

  10. Semiconductor research with reactor neutrons

    International Nuclear Information System (INIS)

    Kimura, Itsuro

    1992-01-01

    Reactor neutrons play an important role for characterization of semiconductor materials as same as other advanced materials. On the other hand reactor neutrons bring about not only malignant irradiation effects called radiation damage, but also useful effects such as neutron transmutation doping and defect formation for opto-electronics. Research works on semiconductor materials with the reactor neutrons of the Kyoto University Reactor (KUR) are briefly reviewed. In this review, a stress is laid on the present author's works. (author)

  11. Research reactors in Austria - Present situation

    International Nuclear Information System (INIS)

    Boeck, H.; Musilek, A.; Villa, M.

    2005-01-01

    In the past decades Austria operated three research reactors, the 10 MW ASTRA reactor at Seibersdorf, the 250 kW TRIGA reactor at the Atominstitut and the 1 kW Argonaut reactor at the Technical University in Graz. Since the shut down of the ASTRA on July 31th, 1999 and its immediate decommissioning reactor and the shut down of the Argonaut reactor in Graz on August 31st, 2004 only one reactor remains operational for keeping nuclear competence in Austria which is the 250 kW TRIGA Mark II reactor. (author)

  12. Growing dimensions. Spent fuel management at research reactors

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    1998-01-01

    More than 550 nuclear research reactors are operating or shout down around the world. At many of these reactors, spent fuel from their operations is stored, pending decisions on its final disposition. In recent years, problems associated with this spent fuel storage have loomed larger in the international nuclear community. In efforts to determine the overall scope of problems and to develop a database on the subject, the IAEA has surveyed research reactor operators in its Member States. Information for the Research Reactor Spent Fuel Database (RRSFDB) so far has been obtained from a limited but representative number of research reactors. It supplements data already on hand in the Agency's more established Research Reactor Database (RRDB). Drawing upon these database resources, this article presents an overall picture of spent fuel management and storage at the world's research reactors, in the context of associated national and international programmes in the field

  13. Research reactor de-fueling and fuel shipment

    International Nuclear Information System (INIS)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-01-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures

  14. Diagnostic measurement on research reactors

    International Nuclear Information System (INIS)

    Dach, K.; Zbytovsky, A.

    A comparison is made of noise experiments on zero power and power reactors. The general characteristics of noise experiments on power reactors is their ''passivity'', i.e., the experiment does not require any interruption of the normal operating regime of the reactor system. On zero power research reactors where the fission reaction constitutes the dominant noise source such conditions have to be created in the study of noise components as to make the investigated noise dominant and the noise of the fission reaction the background. The simultaneous use of both methods makes it possible to determine the spectral composition of reactivity fluctuations, which facilitates the identification of noise sources. The conditions are described of the recordability of noise components. The possibilities are listed provided for research work in Czechoslovakia and the possibility is studied of setting up an expert team to organize the respective experimental programme on an international scale. Power reactors manufactured in the GDR are considered as the suitable experimental base. (J.P.)

  15. The development of mobile melt-dilute technology for the treatment of former Soviet Union research reactor fuel

    International Nuclear Information System (INIS)

    Sell, D.A.; Howden, E.A.; Allen, K.J.; Marsden, K.; Westphal, B.R.; Peacock, H.B.; Iyer, N.C.; Fisher, D.L.; Adams, T.M.; Sindelar, R.L.

    2004-01-01

    United States Government funded national security nuclear non-proliferation projects have historically focused on power reactor spent fuel assemblies that contain weapons usable materials. More recently concern and emphasis have been focused on the spent fuel located at the many research reactor facilities spread throughout the Former Soviet Union. The need exists for a mobile system that can be deployed at these research reactors for the purpose of ensuring that the nuclear materials cannot be used for weapons development. On-site application of the Mobile Melt-Dilute (MMD) process offers an economical method for converting weapons usable Former Soviet Union high enriched uranium research reactor fuel to a safe and secure low enriched uranium ingot. The process will generate little waste and will be performed in a sealed canister that will contain all off-gas products generated during the melting process, eliminating the need for an off-gas treatment system. The process is modular, reusable, and readily portable to a desired reactor site or storage location. The storage canisters containing the melted ingot can be configured for compatibility with the fuel storage technologies currently available or returned to Russia for reprocessing under the Russian Research Reactor Fuel Return Program. The objective of the MMD Project is to develop the mobile melt and dilute technology in preparation for active deployment at Russian built and fueled research reactors. The project has just completed conceptual design and is beginning proof of principle experiments and integrated prototype design of the furnace and canister. (authors)

  16. The Canadian research reactor spent fuel situation

    International Nuclear Information System (INIS)

    Ernst, P.C.

    1996-01-01

    This paper summarizes the present research reactor spent fuel situation in Canada. The research reactors currently operating are listed along with the types of fuel that they utilize. Other shut down research reactors contributing to the storage volume are included for completeness. The spent fuel storage facilities associated with these reactors and the methods used to determine criticality safety are described. Finally the current inventory of spent fuel and where it is stored is presented along with concerns for future storage. (author). 3 figs

  17. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    International Nuclear Information System (INIS)

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

  18. German research reactor back-end provisions

    International Nuclear Information System (INIS)

    Koester, Siegfried; Gruber, Gerhard

    2002-01-01

    Germany has several types of Research Reactors in operation. These reactors use fuel containing uranium of U.S. origin. Basically all the fuel which will be spent until May 2006 will be returned to the U.S. under existing contracts with the U.S. Department of Energy. The contracts are based on the U.S. FRR SNF (Foreign Research Reactor Spent Nuclear Fuel) Program which started in May 1996 and which will last for 10 years. In 1990, the German Federal Government started a program to long-term store (approx. 40 years) and finally dispose of spent fuel in Germany after the so-called U.S. fuel return window will be closed. In order to long-term store the fuel, a special container was designed which covers all different types of spent fuel from the Research Reactors. The container called 'CASTOR MTR 2' is basically licensed and is already in use for the spent fuel of Russian origin from the 'Research Reactor Rossendorf' in the eastern part of Germany. All that fuel is expected to be stored in the existing intermediate storage facility, the so-called BZA (Brennelemente Zwischenlager Ahaus). BZA already accomodates spent fuel from the former THTR-300 high temperature reactor. A final repository does not yet exist in Germany. Alternative provisions to close the back-end of the Research Reactor fuel cycle are reprocessing at COGEMA (France) or in Russian facilities, perspectively. Waste return in a form to be agreed will be mandatory, at least in France. (author)

  19. Spanish collaboration in the OECD Halden Reactor Project research on Gadolinia Fuel

    International Nuclear Information System (INIS)

    Horvath, M.; Munoz-Reja, C.; Tverberg, T.; Jenssen, H. K.

    2010-01-01

    Safe and reliable operation of nuclear power plants benefit from research and development advances and related technical solutions. One research platform is the OECD Halden Reactor Project (HRP). HRP is a joint undertaking of national organisations in 18 countries sponsoring a jointly financed programme under the auspices of the OECD - Nuclear Energy Agency (NEA). As a member state, Spain is participating HRP research programs with ENUSA as a partner in the fuel research programs. Improving the NPP operations, fuel cycles were designed to increase fuel burnup. Higher fuel burnup reduces the number of spent fuel assemblies and thus the costs of new fuel as well as the costs of back-end management. Higher burnup is reached either by prolonging the reactor cycles or by increasing the number of reactor cycles for the fuel in the core. Both ways entail additional requirements concerning fuel enrichment and burnable absorbers as additives and adjustments on the cladding material properties, such as mechanical treatment and chemical composition of the alloys. For these demands and needs ENUSA promotes the research on high burnup effects, gadolinium doped fuels and cladding material behaviour under irradiation. Various experiments, called IFA, are developed and performed also by providing materials. ENUSA collaborates with HRP on various experiments investigating the fuel densification and swelling, fission gas release, pressure limits on UO 2 and (U,Gd)O 2 fuels (IFA-504, -515, -636, -681); the cladding creep, lift-off, corrosion and hydrides on different tubing materials (IFA-567, -610, -638); instrumentation of the experiments, especially on pre-irradiated materials (IFA-533). These experiments are combined with model calculations to improve predictions for higher burnups and to maintain safety margins (IFA-515, -636, -681). Besides these unique in-pile experiments PIEs are performed as well on fuel and structural materials to complete the scope of these studies (IFA

  20. Education and Training on ISIS Research Reactor

    International Nuclear Information System (INIS)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X.

    2013-01-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions

  1. Education and Training on ISIS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  2. The first university research reactor in India

    International Nuclear Information System (INIS)

    Murty, G.S.

    1999-01-01

    As the first university research reactor in India, the low power, pool type with fixed core and low enriched uranium fuel research reactor is under construction in the Andhra university campus, Andhra Pradesh, India. The reactor is expected to be commissioned during 2001-2002. The mission of the reactor is to play the research center as a regional research facility catering to the needs of academic institutions and industrial organizations of this region of the country. Further, to encourage interdisplinary and multidisplinary research activities, to supply radioisotope and labelled compounds to the user institutions and to create awareness towards the peaceful uses of atomic energy. This report describes its objectives, status and future plans in brief. (H. Itami)

  3. The first university research reactor in India

    Energy Technology Data Exchange (ETDEWEB)

    Murty, G.S. [Co-ordinator, Low Power Research Reactor, Andhra Univ., Visakapatnam (India)

    1999-08-01

    As the first university research reactor in India, the low power, pool type with fixed core and low enriched uranium fuel research reactor is under construction in the Andhra university campus, Andhra Pradesh, India. The reactor is expected to be commissioned during 2001-2002. The mission of the reactor is to play the research center as a regional research facility catering to the needs of academic institutions and industrial organizations of this region of the country. Further, to encourage interdisplinary and multidisplinary research activities, to supply radioisotope and labelled compounds to the user institutions and to create awareness towards the peaceful uses of atomic energy. This report describes its objectives, status and future plans in brief. (H. Itami)

  4. Cost aspects of the research reactor fuel cycle

    International Nuclear Information System (INIS)

    2010-01-01

    Research reactors have made valuable contributions to the development of nuclear power, basic science, materials development, radioisotope production for medicine and industry, and education and training. In doing so, they have provided an invaluable service to humanity. Research reactors are expected to make important contributions in the coming decades to further development of the peaceful uses of nuclear technology, in particular for advanced nuclear fission reactors and fuel cycles, fusion, high energy physics, basic research, materials science, nuclear medicine, and biological sciences. However, in the context of decreased public sector support, research reactors are increasingly faced with financial constraints. It is therefore of great importance that their operations are based on a sound understanding of the costs of the complete research reactor fuel cycle, and that they are managed according to sound financial and economic principles. This publication is targeted at individuals and organizations involved with research reactor operations, with the aim of providing both information and an analytical framework for assessing and determining the cost structure of fuel cycle related activities. Efficient management of fuel cycle expenditures is an important component in developing strategies for sustainable future operation of a research reactor. The elements of the fuel cycle are presented with a description of how they can affect the cost efficient operation of a research reactor. A systematic review of fuel cycle choices is particularly important when a new reactor is being planned or when an existing reactor is facing major changes in its fuel cycle structure, for example because of conversion of the core from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, or the changes in spent fuel management provision. Review and optimization of fuel cycle issues is also recommended for existing research reactors, even in cases where research reactor

  5. Developments in the regulation of research reactors

    International Nuclear Information System (INIS)

    Loy, J.

    2003-01-01

    The International Atomic Energy Agency (IAEA) has data on over 670 research reactors in the world. Fewer than half of them are operational and a significant number are in a shutdown but not decommissioned state. The International Nuclear Safety Advisory Group (INSAG) has expressed concerns about the safety of many research reactors and this has resulted in a process to draw up an international Code of Conduct on the Safety of Research Reactors. The IAEA is also reviewing its safety standards applying to research reactors. On the home front, regulation of the construction of the Replacement Research Reactor continues. During the construction phase, regulation has centred around the consideration of Requests for Approval (RFA) for the manufacture and installation of systems, structures and components important for safety. Quality control of construction of systems, structures and components is the central issue. The process for regulation of commissioning is under consideration

  6. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  7. History, Development and Future of TRIGA Research Reactors

    International Nuclear Information System (INIS)

    2016-01-01

    Due to its particular fuel design and resulting enhanced inherent safety features, TRIGA reactors (Training, Research, Isotopes, General Atomics) constitute a ‘class of their own’ among the large variety of research reactors built world-wide. This publication summarizes in a single document the information on the past and present of TRIGA research reactors and presents an outlook in view of potential issues to be solved by TRIGA operating organizations in the near future. It covers the historical development and basic TRIGA characteristics, followed by utilization, fuel conversion and ageing management of TRIGA research reactors. It continues with issues and challenges, introduction to the global TRIGA research reactor network and concludes with future perspectives. The publication is complemented with a CD-ROM to illustrate the historical developments of TRIGA research reactors through individual facility examples and experiences

  8. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

    International Nuclear Information System (INIS)

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

    1988-01-01

    According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  9. Research reactors: a tool for science and medicine

    International Nuclear Information System (INIS)

    Ordonez, Juan

    2001-01-01

    The types and uses of research reactors are reviewed. After an analysis of the world situation, the demand of new research reactors of about 20 MW is foreseen. The experience and competitiveness of INVAP S.E. as designer and constructor of research reactors is outlined and the general specifications of the reactors designed by INVAP for Egypt and Australia are given

  10. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    International Nuclear Information System (INIS)

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions

  11. Reactor safety research. The CEC contribution

    International Nuclear Information System (INIS)

    Krischer, W.

    1990-01-01

    The involvement of the EC Commission in the reactor safety research dates back almost to the implementation of the EURATOM Treaty and has thus lasted for thirty years. The need for close collaboration and for general consensus on some crucial problems of concern to the public, has made the role of international organizations and, as far as Europe is concerned, the role of the European Community particularly important. The areas in which the CEC has been active during the last five years are widespread. This is partly due to the fact that, after TMI and Chernobyl, the effort and the interest of the different countries in reactor safety was considerable. Reactor Safety Research represents the proceedings of a seminar held by the Commission at the end of its research programme 1984-88 on reactor safety. As such it gives a comprehensive overview of the recent activities and main results achieved in the CEC Joint Research Centre and in national laboratories throughout Europe on the basis of shared cost actions. In a concluding chapter the book reports on the opinions, expressed during a panel by a group of major exponents, on the needs for future research. The main topics addressed are, with particular reference to Light Water Reactors (LWRS): reliability and risk evaluation, inspection of steel components, primary circuit components end-of-life prediction, and abnormal behaviour of reactor cooling systems. As far as LMFBRs are concerned, the topics covered are: severe accident modelling, material properties and structural behaviour studies. There are 67 pages, all of which are indexed separately. Reactor Safety Research will be of particular interest to reliability and safety engineers, nuclear engineers and technicians, and mechanical and structural engineers. (author)

  12. Human machine interaction research experience and perspectives as seen from the OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    Oewre, F.

    1999-01-01

    In this paper a short review is given on important safety issues in the field of human machine interaction as expressed by important nuclear organisations such as USNRC, IAEA and the OECD NEA. Further on, a presentation is offered of research activities at the OECD Halden Reactor Project in the field of human machine interaction aiming to clarify some of the issues outlined by the above mentioned organisations. The OECD Halden Reactor Project is a joint undertaking of national nuclear organisations in 19 countries sponsoring a jointly financed research programme under the auspices of the OECD - Nuclear Energy Agency. One of the research areas is the man-machine systems research addressing the operator tasks in a control room environment. The overall objective is to provide a basis for improving today's control rooms through introduction of computer-based solutions for effective and safe execution of surveillance and control functions in normal as well as off-normal plant situations. (author)

  13. CSAU (code scaling, applicability and uncertainty), a tool to prioritize advanced reactor research

    International Nuclear Information System (INIS)

    Wilson, G.E.; Boyack, B.E.

    1990-01-01

    Best Estimate computer codes have been accepted by the US Nuclear Regulatory Commission as an optional tool for performing safety analysis related to the licensing and regulation of current nuclear reactors producing commercial electrical power, providing their uncertainty is quantified. In support of this policy change, the NRC and its contractors and consultants have developed and demonstrated an uncertainty quantification methodology called CSAU. At the process level, the method is generic to any application which relies on best estimate computer code simulations to determine safe operating margins. The primary use of the CSAU methodology is to quantify safety margins for existing designs; however, the methodology can also serve an equally important role in advanced reactor research for plants not yet built. Applied early, during the period when alternate designs are being evaluated, the methodology can identify the relative importance of the sources of uncertainty in the knowledge of each plant behavior and, thereby, help prioritize the research needed to bring the new designs to fruition. This paper describes the CSAU methodology, at the generic process level, and provides the general principles whereby it may be applied to evaluations of advanced reactor designs. 9 refs., 1 fig., 1 tab

  14. 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.)

  15. A compact, inherently safe liquid metal reactor plant concept for terrestrial defense power applications

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Lutz, D.E.; Palmer, R.S.

    1987-01-01

    A compact, inherently safe, liquid metal reactor concept based on the GE PRISM innovative LMR design has been developed for terrestrial defense power applications in the 2-50 MWe range. The concept uses a small, sodium-cooled, U-5%Zr metal fueled reactor contained within two redundant steel vessels. The core is designed to operate at a low power density and temperature (925 F) and can operate 30 years without refueling. One two primary coolant loops, depending upon the plant size, transport heat from the core to sodium-to-air, double-wall heat exchangers. Power is produced by a gas turbine operated in a closed ''bottoming'' cycle that employs intercoolers between the compressor stages and a recuperator. Inherent safety is provided by passive means only; operator action is not required to ensure plant safety even for events normally considered Beyond Design Basis Accidents. In addition to normal shutdown heat removal via the sodium-to-air heat exchangers, the design utilizes an inherently passive radiant vessel auxiliary cooling system similar to that designed for PRISM. The use of an air cycle gas turbine eliminates the cost and complexity of the sodium-water reactor pressure relief system required for a steam cycle sodium-cooled reactor

  16. The AFR. An approved network of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele [Mainz Univ. (Germany). Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren (AFR)

    2012-10-15

    AFR (Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren) is the German acronym for 'Association for Research Reactor Operation and Safety Issues' which was founded in 1959. Reactor managers of European research reactors mainly from the German linguistic area meet regularly for their mutual benefit to exchange experience and knowledge in all areas of operating, managing and utilization of research reactors. In the last 2 years joint meetings were held together with the French association of research reactors CER (Club d'Exploitants des Reacteurs). In this contribution the AFR, its members, work and aims as well as the French partner CER are presented. (orig.)

  17. Safety of Research Reactors. Safety Requirements

    International Nuclear Information System (INIS)

    2010-01-01

    The main objective of this Safety Requirements publication is to provide a basis for safety and a basis for safety assessment for all stages in the lifetime of a research reactor. Another objective is to establish requirements on aspects relating to regulatory control, the management of safety, site evaluation, design, operation and decommissioning. Technical and administrative requirements for the safety of research reactors are established in accordance with these objectives. This Safety Requirements publication is intended for use by organizations engaged in the site evaluation, design, manufacturing, construction, operation and decommissioning of research reactors as well as by regulatory bodies

  18. IRT-type research reactor physical calculation methodology

    International Nuclear Information System (INIS)

    Carrera, W.; Castaneda, S.; Garcia, F.; Garcia, L.; Reyes, O.

    1990-01-01

    In the present paper an established physical calculation procedure for the research reactor of the Nuclear Research Center (CIN) is described. The results obtained by the method are compared with the ones reported during the physical start up of a reactor with similar characteristics to the CIN reactor. 11 refs

  19. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  20. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  1. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  2. RMB. The new Brazilian multipurpose research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto; Soares, Adalberto Jose

    2015-01-01

    Brazil has four research reactors (RR) in operation: IEA-R1, a 5 MW pool type RR; IPR-R1, a 100 kW TRIGA type RR; ARGONAUTA, a 500 W Argonaut type RR, and IPEN/MB-01, a 100 W critical facility. The first three were constructed in the 50's and 60's, for teaching, training, and nuclear research, and for many years they were the basic infrastructure for the Brazilian nuclear developing program. The last, IPEN/MB-01, is the result of a national project developed specifically for qualification of reactor physics codes. Considering the relative low power of Brazilian research reactors, with exception of IEAR1, none of the other reactors are feasible for radioisotope production, and even IEA-R1 has a limited capacity. As a consequence, since long ago, 100% of the Mo-99 needed to attend Brazilian nuclear medicine services has been imported. Because of the high dependence on external supply, the international Moly-99 supply crisis that occurred in 2008/2009 affected significantly Brazilian nuclear medicine services, and as presented in previous IAEA events, in 2010 Brazilian government formalized the decision to build a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be a 30 MW open pool type reactor, using low enriched uranium fuel. The facility will be part of a new nuclear research centre, to be built about 100 kilometres from Sao Paulo city, in the southern part of Brazil. The new nuclear research centre will have several facilities, to use thermal and cold neutron beams; to produce radioisotopes; to perform neutron activation analysis; and to perform irradiations tests of materials and fuels of interest for the Brazilian nuclear program. An additional facility will be used to store, for at least 100 years, all the fuel used in the reactor. The paper describes the main characteristics of the new centre, emphasising the research reactor and giving a brief description of the laboratories that will be constructed, It also presents the

  3. RMB. The new Brazilian multipurpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, Jose Augusto; Soares, Adalberto Jose [Comissao Nacional de Energia Nuclear (CNEN) (Brazil)

    2015-01-15

    Brazil has four research reactors (RR) in operation: IEA-R1, a 5 MW pool type RR; IPR-R1, a 100 kW TRIGA type RR; ARGONAUTA, a 500 W Argonaut type RR, and IPEN/MB-01, a 100 W critical facility. The first three were constructed in the 50's and 60's, for teaching, training, and nuclear research, and for many years they were the basic infrastructure for the Brazilian nuclear developing program. The last, IPEN/MB-01, is the result of a national project developed specifically for qualification of reactor physics codes. Considering the relative low power of Brazilian research reactors, with exception of IEAR1, none of the other reactors are feasible for radioisotope production, and even IEA-R1 has a limited capacity. As a consequence, since long ago, 100% of the Mo-99 needed to attend Brazilian nuclear medicine services has been imported. Because of the high dependence on external supply, the international Moly-99 supply crisis that occurred in 2008/2009 affected significantly Brazilian nuclear medicine services, and as presented in previous IAEA events, in 2010 Brazilian government formalized the decision to build a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be a 30 MW open pool type reactor, using low enriched uranium fuel. The facility will be part of a new nuclear research centre, to be built about 100 kilometres from Sao Paulo city, in the southern part of Brazil. The new nuclear research centre will have several facilities, to use thermal and cold neutron beams; to produce radioisotopes; to perform neutron activation analysis; and to perform irradiations tests of materials and fuels of interest for the Brazilian nuclear program. An additional facility will be used to store, for at least 100 years, all the fuel used in the reactor. The paper describes the main characteristics of the new centre, emphasising the research reactor and giving a brief description of the laboratories that will be constructed, It also

  4. RMB: the new Brazilian Multipurpose Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, Jose Augusto, E-mail: perrotta@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2016-07-01

    Full text: The Brazilian research reactors have a limited capacity for radioisotopes production, leading to a high dependence on external supply for radioisotopes used in nuclear medicine. In order to overcome this condition and due to the old age of these research reactors, the Brazilian Nuclear Energy Commission decided, in 2008, to construct a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be part of a new nuclear research center, to be built on a site about 100 kilometers from São Paulo city, in the southern part of Brazil. The new nuclear research center will have a 30 MW open pool type research reactor using low enriched uranium fuel, and several associated laboratories in order to produce radioisotopes for medical and industrial use, to use neutron beams in scientific and technological research; to perform neutron activation analysis; and to perform materials and fuels irradiation tests. Regarding the neutron beams use, the RMB design provides thermal and cold neutron beams. From one side of the reactor, the neutron guides will extend to an experimental hall of instruments named Neutron Guide Hall where it will be installed the scattering instruments. In the initial stage of the reactor operation, the intent is to implement two neutron guides for thermal neutrons and another two for cold neutrons. The 2015 SBPMAT symposium has presented the technical overview of the RMB project and its main buildings, structures and components. At this year symposium, the RMB presentation updates some technical information and the development status of the project, discussing the negative results of the Brazilian political and economic crisis to the project development and its future perspectives. (author)

  5. Present status of research reactor and future prospects

    International Nuclear Information System (INIS)

    Nakajima, Ken

    2013-01-01

    Research reactors have been playing an important role in the research and development of the various fields, such as physics, chemistry, biology, engineering, agriculture, medicine, etc. as well as human resource development. However, the most of them are older than 40 years, and the ageing management is an important issue. In Japan, only two research reactors are operational after the Great East Japan Earthquake in 2011. JAEA's reactors suffered from the quake and they are under inspections. Kyoto University Research Reactor, one of the operational reactors, has been widely used for research and human resource development, and the additional safety measures against the station blackout were installed. Besides the affect of the quake, the disposal or treatment of spent fuel becomes an inevitable problem for research reactors. The way of spent fuel disposal or treatment should be determined with the nation-wide and/or international coalition. (author)

  6. Application of research reactors for radiation education

    International Nuclear Information System (INIS)

    Ito, Yasuo; Harasawa, Susumu; Hayashi, Shu A.; Tomura, Kenji; Matsuura, Tatsuo; Nakanishi, Tomoko M.; Yamamoto, Yusuke

    1999-01-01

    Nuclear research Reactors are, as well as being necessary for research purposes, indispensable educational tools for a country whose electric power resources are strongly dependent on nuclear energy. Both large and small research reactors are available, but small ones are highly useful from the viewpoint of radiation education. This paper oders a brief review of how small research reactors can, and must, be used for radiation education for high school students, college and graduate students, as well as for the public. (author)

  7. Application of research reactors for radiation education

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Yasuo [Tokyo Univ. (Japan). Research Center for Nuclear Science and Technology; Harasawa, Susumu; Hayashi, Shu A.; Tomura, Kenji; Matsuura, Tatsuo; Nakanishi, Tomoko M.; Yamamoto, Yusuke

    1999-09-01

    Nuclear research Reactors are, as well as being necessary for research purposes, indispensable educational tools for a country whose electric power resources are strongly dependent on nuclear energy. Both large and small research reactors are available, but small ones are highly useful from the viewpoint of radiation education. This paper oders a brief review of how small research reactors can, and must, be used for radiation education for high school students, college and graduate students, as well as for the public. (author)

  8. Commercial products and services of research reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2013-07-01

    Although the number of operational research reactors is steadily decreasing, more than half of those that remain are greatly underutilized and, in most cases, underfunded. To continue to play a key role in the development of peaceful uses of nuclear technology, the remaining research reactors will need to provide useful products and services to private, national and regional customers, in some cases with adequate revenue generation for reliable, safe and secure facility management and operation. In the light of declining governmental financial support and the need for improved physical security and conversion to low enriched uranium (LEU) fuel, many research reactors have been challenged to generate income to offset increasing operational and maintenance costs. The renewed interest in nuclear power (and therefore in nuclear education and training), the global expansion of diagnostic and therapeutic nuclear medicine, and the extensive use of semiconductors in electronics and in other areas have created new opportunities for research reactors, prominent among them, markets for products and services in regions and countries without such facilities. It is clear that such initiatives towards greater self-reliance will need to address such aspects as market surveys, marketing and business plans, and cost of delivery services. It will also be important to better inform present and future potential end users of research reactor services of the capabilities and products that can be provided. This publication is a compilation of material from an IAEA technical meeting on “Commercial Products and Services of Research Reactors”, held in Vienna, Austria, from 28 June to 2 July 2010. The overall objective of the meeting was to exchange information on good practices and to provide concrete examples, in technical presentations and brainstorming discussions, to promote and facilitate the development of commercial applications of research reactors. The meeting also aimed to

  9. The national standards program for research reactors

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1977-01-01

    In 1970 a standards committee called ANS-15 was established by the American Nuclear Society (ANS) to prepare appropriate standards for research reactors. In addition, ANS acts as Secretariat for a national standards committee N17 which is responsible to the American National Standards Institute (ANSI) for the national consensus efforts for standards related to research reactors. To date ANS-15 has completed or is working on 14 standards covering all aspects of the operation of research reactors. Of the 11 research reactor standards submitted to the ANSI N17 Committee since its inception, six have been issued as National standards, and the remaining are still in the process of review. (author)

  10. Contributions of research Reactors in science and technology

    International Nuclear Information System (INIS)

    Butt, N.M.; Bashir, J.

    1992-12-01

    In the present paper, after defining a research reactor, its basic constituents, types of reactors, their distribution in the world, some typical examples of their uses are given. Particular emphasis in placed on the contribution of PARR-I (Pakistan Research Reactor-I), the 5 MW Swimming Pool Research reactor which first became critical at the Pakistan Institute of Nuclear Science and Technology (PINSTECH) in Dec. 1965 and attained its full power in June 1966. This is still the major research facility at PINSTECH for research and development. (author)

  11. Overview of fast reactor safety research and development in the USA

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Avery, R.; Marchaterre, J.F.

    1986-01-01

    The liquid metal reactor (LMR) safety R and D program in the U.S. is presently focused on support of two modular innovative reactor concepts: PRISM - the General Electric Power Reactor Inherently Safe Module and SAFR - the Rockwell International Sodium Advanced Fast Reactor. These reactor plant concepts accommodate the use of either oxide fuel or the metal fuel which is under development in the Argonne National Laboratory (ANL) Integral Fast Reactor (IFR) program. Both concepts emphasize prevention of accidents through enhancement of inherent and passive safety characteristics. Enhancement of these characteristics is expected to be a major factor in establishing new and improved safety criteria and licensing arrangements with regulatory authorities for advanced reactors. Limited work is also continuing on the Large Scale Prototype Breeder (LSPB), a large pool plant design. Major elements of the current and restructured safety program are discussed. (author)

  12. Cost effective safety enhancements for research reactors in Uzbekistan and Kazakhstan - results of a joint program with US DOE

    International Nuclear Information System (INIS)

    Earle, O.K.; Carlson, R.B.; Rakhmanov, A.; Salikhbaev, U.S.; Chernyaev, V.; Chakrov, P.

    2004-01-01

    Full text: The US Department of Energy's Office of International Nuclear Safety and Cooperation established the Integrated Research Reactor Safety Enhancement Program (IRRSEP) in February 2002 to support U.S. nonproliferation goals by (1) implementing safety upgrades, or (2) assisting with the safe shutdown and decommissioning of foreign test and research reactors which present security concerns. IRRSEP's key program components are: Phase I: Self-evaluation by facility using provided checklists followed by prioritization to identify the 20 highest risk facilities; Phase II: Site visits with technical evaluation to finalize a list of projects that will enhance safety consistent with IAEA observations; Phase III: Corrective measures to implement the projects. Phases I, II and III are accomplished on a rolling basis, such that work is ongoing at three or four reactors per year. IRRSEP's key objective is to resolve the highest-priority nuclear safety issues at the most vulnerable foreign research reactors as quickly as possible. The prioritization methodology employed identified which research reactors fell into this category. The corrective measures mutually developed with the host facility are based on the premise of developing a sustainable infrastructure within each country to deal with its own nuclear material safety, security, and response issues in the future. IRRSEP also assists in creating an international framework of cooperation and openness between research and test reactor operators, and national and international regulators. The initial projects under IRRSEP are underway at research reactors in Kazakhstan, Uzbekistan, and Romania. This paper focuses on the projects undertaken at the WWR-K research reactor at the Institute of Nuclear Physics in Alatau, Kazakhstan and the WWR-SM research reactor at the Institute of Nuclear Physics in Ulugbek, Uzbekistan. These projects demonstrate the success and cost effectiveness of the IRRSEP program

  13. Cost effective safety enhancements for research reactors in Uzbekistan and Kazakhstan - results of a joint program with US DOE

    International Nuclear Information System (INIS)

    Earle, O.K.; Carlson, R.B.; Rakhmanov, A.; Salikhbaev, U.S.; Chernyaev, V.; Chakrov, P.

    2004-01-01

    The US Department of Energy's Office of International Nuclear Safety and Cooperation established the Integrated Research Reactor Safety Enhancement Program (IRRSEP) in February 2002 to support U.S. nonproliferation goals by implementing safety upgrades, or assisting with the safe shutdown and decommissioning of foreign test and research reactors which present security concerns. IRRSEP's key program components are: Phase I: Self-evaluation by facility using provided checklists followed by prioritization to identify the 20 highest risk facilities; Phase II: Site visits with technical evaluation to finalize a list of projects that will enhance safety consistent with IAEA observations; Phase III: Corrective measures to implement the projects. Phases I, II and III are accomplished on a rolling basis, such that work is ongoing at three or four reactors per year. IRRSEP's key objective is to resolve the highest-priority nuclear safety issues at the most vulnerable foreign research reactors as quickly as possible. The prioritization methodology employed identified which research reactors fell into this category. The corrective measures mutually developed with the host facility are based on the premise of developing a sustainable infrastructure within each country to deal with its own nuclear material safety, security, and response issues in the future. IRRSEP also assists in creating an international framework of cooperation and openness between research and test reactor operators, and national and international regulators. The initial projects under IRRSEP are underway at research reactors in Kazakhstan, Uzbekistan, and Romania. This paper focuses on the projects undertaken at the WWR-K research reactor at the Institute of Nuclear Physics in Alatau, Kazakhstan and the WWR-SM research reactor at the Institute of Nuclear Physics in Ulugbek, Uzbekistan. These projects demonstrate the success and cost effectiveness of the IRRSEP program

  14. Research reactor management. Safety improvement activities in HANARO

    International Nuclear Information System (INIS)

    Wu, Jong-Sup; Jung, Hoan-Sung; Hong, Sung Taek; Ahn, Guk-Hoon

    2012-01-01

    Safety activities in HANARO have been continuously conducted to enhance its safe operation. Great effort has been placed on a normalization and improvement of the safety attitude of the regular staff and other employees working at the reactor and other experimental facilities. This paper introduces the activities on safety improvement that were performed over the last few years. (author)

  15. Experience and prospects for developing research reactors of different types

    International Nuclear Information System (INIS)

    Kuatbekov, R.P.; Tretyakov, I.T.; Romanov, N.V.; Lukasevich, I.B.

    2015-01-01

    NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

  16. RA Research reactor, Annual report 1971

    International Nuclear Information System (INIS)

    Milosevic, D. et al.

    1971-12-01

    effects. In the introduction of this report it has been emphasised that the decision makers should have in mind the negative effects of low budget on the reactor safe and reliable operation. For the sake of reactor, decision about the future operation and financing should be done as soon as possible, either to cease operation or continue with adequate financial support [sr

  17. Main refurbishment activities on electronic and electrical equipment for the FRG-1 research reactor

    International Nuclear Information System (INIS)

    Blom, K.H.; Krull, W.

    1997-01-01

    As GKSS intends to operate the research reactor FRG-1 safely and reliably for many years to come, the plant is constantly refurbished and upgraded both in the interests of safety and operational reasons. The following electronic and electrical systems have been replaced or improved since 1990: Information and signalling systems; Emergency power plant (permit applied for); External and internal lightning protection system; Reactor protection system (in part); Safety lighting; Alarm and staff locating system; Control room telephone system; Closed-circuit television system; Beam tube controls; Storage plant for radioactive liquid waste; Ambient dose rate measuring system; Meteorological measuring system; Control and measuring system for the primary cooling circuit; Control rod drives; Control rod control system; Soft start for the secondary pumps; Control and switching devices for the emergency power plant; Trailing cable installation for the reactor bridge; Main-voltage distribution systems/cable routes. (author). 13 figs, 1 tab

  18. Research reactor's role in Korea

    International Nuclear Information System (INIS)

    Choi, C-O.

    1995-01-01

    After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960's in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs

  19. Making better use of research reactors

    International Nuclear Information System (INIS)

    1964-01-01

    Some 250 research reactors are in operation in the world today, and there are problems in putting them to the most fruitful use. The difficulties - of trained manpower, of auxiliary equipment, of satisfactory research programmes, of co-ordination, between the various disciplines - are common to all users. But as is only to be expected, they press more heavily on the newly-established centres, particularly those in the developing countries which are lacking in long experience in research and usually severely limited as to technical manpower and money. The IAEA has been turning its attention to this question for the past three or four years - ever since, in fact, its early assistance missions and other field operations brought it into close contact with the operations of numerous Member States. The task of providing assistance and advice in this matter is growing. Many centres have been building research reactors under bilateral arrangements; with the completion of their projects this form of aid usually ends, and they look to IAEA for help in operating the reactors. Although some critics consider that difficulties have been caused by premature construction of research reactors, before well-founded programmes of nuclear research had been developed in the countries concerned, several valid motives have led to the establishment of some of these centres at an early stage. A research reactor often provides an effective stimulant for scientific research in the country. It is a remarkably versatile tool for workers in many fields of science and technology. There have been instances where the establishment of a research reactor has had a great impact on the scientific education of a country and has led to a salutary reappraisal and reforms. A reactor is sometimes considered to be a particularly effective means of retaining in the country men trained in the nuclear field. This particular problem is common to most countries. In fact, it is a feature of the present age that

  20. Supervisory system to monitor the neutron flux of the IPR-R1 TRIGA research reactor at CDTN

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Tello, Cledola Cassia Oliveira

    2009-01-01

    The IPR-R1 TRIGA Mark I nuclear research reactor at the Nuclear Technology Development Center - CDTN (Belo Horizonte) is a pool type reactor. It was designed for research, training and radioisotope production. The International Atomic Energy Agency- IAEA - recommends the use of friendly interfaces for monitoring and controlling the operational parameters of nuclear reactors. This paper reports the activities for implementing a supervisory system, using LabVIEW software, with the purpose to provide the IPR-R1 TRIGA research reactor with a modern, safe and reliable system to monitor the time evolution of the power of its core. The use of the LabVIEW will introduce modern techniques, based on electronic processor and visual interface in video monitor, substituting the mechanical strip chart recorders (ink-pen drive and paper) that monitor the current neutrons flux, which is proportional to the thermal power supplied by reactor core. The main objective of the system will be to follow the evolution of the neutronic flux originated in the Linear and Logarithmic channels. A great advantage of the supervisory software nowadays, in relation to computer programs currently used in the facility, is the existence of new resources such as the data transmission and graphical interfaces by net, grid lines display in the graphs, and resources for real time reactor core video recordings. The considered system could also in the future be optimized, not only for data acquisition, but also for the total control of IPR-R1 TRIGA reactor(author)

  1. Research reactor spent nuclear fuel shipment from the Czech Republic to the Russian Federation

    International Nuclear Information System (INIS)

    Svoboda, K.; Broz, V.; Novosad, P.; Podlaha, J.; Svitak, F.

    2009-01-01

    In May 2004, the Global Threat Reduction Initiative agreement was signed by the governments of the United States and the Russian Federation. The goal of this initiative is to minimize, in cooperation with the International Atomic Energy Agency (IAEA) in Vienna, the existing threat of misuse of nuclear and radioactive materials for terrorist purposes, particularly highly enriched uranium (HEU), fresh and spent nuclear fuel (SNF), and plutonium, which have been stored in a number of countries. Within the framework of the initiative, HEU materials and SNF from research reactors of Russian origin will be transported back to the Russian Federation for reprocessing/liquidation. The program is designated as the Russian Research Reactor Fuel Return (RRRFR) Program and is similar to the U.S. Foreign Research Reactor Spent Nuclear Fuel Acceptance Program, which is underway for nuclear materials of United States origin. These RRRFR activities are carried out under the responsibilities of the respective ministries (i.e., U.S. Department of Energy (DOE) and Russian Federation Rosatom). The Czech Republic and the Nuclear Research Institute Rez, plc (NRI) joined Global Threat Reduction Initiative in 2004. During NRI's more than 50 years of existence, radioactive and nuclear materials had accumulated and had been safely stored on its grounds. In 1995, the Czech regulatory body , State Office for Nuclear Safety (SONS), instructed NRI that all ecological burdens from its past activities must be addressed and that the SNF from the research reactor LVR -15 had to be transported for reprocessing. At the end of November 2007, all these activities culminated with the unique shipment to the Russian Federation of 527 fuel assemblies of SNF type EK-10 (enrichment 10% U-235) and IRT-M (enrichment 36% and 80% U-235) and 657 irradiated fuel rods of EK-10 fuel, which were used in LVR-15 reactor. (authors)

  2. Decommissioning techniques for research reactors. Final report of a co-ordinated research project 1997-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-02-01

    in this technical publication. Operating experience in real-scale applications, lessons learned, key results in laboratory scale or pilot scale research, and validation of mathematical models, are among the most significant achievements of the CRP and have been highlighted. The objective of this CRP was to promote the exchange of information on the practical experience gained by Member States in decommissioning or operation, maintenance, and refurbishment activities which would be eventually related to the decommissioning of research reactors. Special emphasis was given to the development/adaptation of methods and approaches for optimization of the decommissioning process. The scope of the project included several technical areas of decommissioning rather than focusing on a single aspect of it. It was felt that this format would generate more awareness of the integrated approach to decommissioning. In particular, the scope included the following: design, construction and operational features to assist in final decommissioning; planning for decommissioning, including technical solution assessment; decommissioning strategies and their technological implications; radiological and physical characterization; dismantling technology; decontamination technology; remotely operated equipment; means to reduce occupational exposures; waste generation and management, including clearance of solid materials; restricted and unrestricted site release, including final surveys; costs and financial provisions; safe enclosure of shutdown reactors, including long-term integrity of buildings and systems; decommissioning experience; and ageing management and refurbishment experience.

  3. Decommissioning techniques for research reactors. Final report of a co-ordinated research project 1997-2001

    International Nuclear Information System (INIS)

    2002-02-01

    in this technical publication. Operating experience in real-scale applications, lessons learned, key results in laboratory scale or pilot scale research, and validation of mathematical models, are among the most significant achievements of the CRP and have been highlighted. The objective of this CRP was to promote the exchange of information on the practical experience gained by Member States in decommissioning or operation, maintenance, and refurbishment activities which would be eventually related to the decommissioning of research reactors. Special emphasis was given to the development/adaptation of methods and approaches for optimization of the decommissioning process. The scope of the project included several technical areas of decommissioning rather than focusing on a single aspect of it. It was felt that this format would generate more awareness of the integrated approach to decommissioning. In particular, the scope included the following: design, construction and operational features to assist in final decommissioning; planning for decommissioning, including technical solution assessment; decommissioning strategies and their technological implications; radiological and physical characterization; dismantling technology; decontamination technology; remotely operated equipment; means to reduce occupational exposures; waste generation and management, including clearance of solid materials; restricted and unrestricted site release, including final surveys; costs and financial provisions; safe enclosure of shutdown reactors, including long-term integrity of buildings and systems; decommissioning experience; and ageing management and refurbishment experience

  4. Containment design, performance criteria and research needs for advanced reactor designs

    International Nuclear Information System (INIS)

    Bagdi, G.; Ali, S.; Costello, J

    2004-01-01

    This paper points out some important shifts in the basic expectations in the performance requirements for containment structures and discusses the areas where the containment structure design requirements and acceptance criteria can be integrated with ultimate test based insights. Although there has not been any new reactor construction in the United States for over thirty years, several designs of evolutionary and advanced reactors have already been certified. Performance requirements for containment structures under design basis and severe accident conditions and explicit consideration of seismic margins have been used in the design certification process. In the United States, the containment structure design code is the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NE-Class MC for the steel containment and Section III, Division 2 for reinforced and prestressed concrete reactor vessels and containments. This containment design code was based on the early concept of applying design basis internal pressure and associated load combinations that included the operating basis and safe shutdown earthquake ground motion. These early design criteria served the nuclear industry and the regulatory authorities in maintaining public health and safety. However, these early design criteria do not incorporate the performance criteria related to containment function in an integrated fashion. Research in large scale model testing of containment structures to failure from over pressurization and shake table testing using simulated ground motion, have produced insights related to failure modes and material behavior at failure. The results of this research provide the opportunity to integrate these observations into design and acceptance criteria. This integration process would identify 'gaps' in the present knowledge and future research needs. This knowledge base is important for gleaning risk-informed insights into

  5. The Waste Management Plan integration into Decommissioning Plan of the WWR-S research reactor from Romania

    International Nuclear Information System (INIS)

    Barariu, Gheorghe; Oprescu, Theodor; Filip, Mihaela; Sociu, Florin

    2008-01-01

    The paper presents the progress of the Radioactive Waste Management Plan which accompanies the Decommissioning Plan for research reactor WWR-S located in Magurele, Ilfov, near Bucharest, Romania. The new variant of the Decommissioning Plan was elaborated taking into account the IAEA recommendation concerning radioactive waste management. A new feasibility study for WWR-S decommissioning was also developed. The preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as rehabilitation of the existing Radioactive Waste Treatment Plant and the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and reusing of cleaned reactor building is envisaged. An inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the IAEA assistance. Environmental concerns are part of the radioactive waste management strategy. (authors)

  6. Safe handling of tritium

    International Nuclear Information System (INIS)

    1991-01-01

    The main objective of this publication is to provide practical guidance and recommendations on operational radiation protection aspects related to the safe handling of tritium in laboratories, industrial-scale nuclear facilities such as heavy-water reactors, tritium removal plants and fission fuel reprocessing plants, and facilities for manufacturing commercial tritium-containing devices and radiochemicals. The requirements of nuclear fusion reactors are not addressed specifically, since there is as yet no tritium handling experience with them. However, much of the material covered is expected to be relevant to them as well. Annex III briefly addresses problems in the comparatively small-scale use of tritium at universities, medical research centres and similar establishments. However, the main subject of this publication is the handling of larger quantities of tritium. Operational aspects include designing for tritium safety, safe handling practice, the selection of tritium-compatible materials and equipment, exposure assessment, monitoring, contamination control and the design and use of personal protective equipment. This publication does not address the technologies involved in tritium control and cleanup of effluents, tritium removal, or immobilization and disposal of tritium wastes, nor does it address the environmental behaviour of tritium. Refs, figs and tabs

  7. Neutrons down-under: Australia's research reactor review

    International Nuclear Information System (INIS)

    Murray, Allan

    1995-01-01

    Australian research reactor review commenced in September 1992, the Review had the following Terms of Reference: Whether, on review of the benefits and costs for scientific, commercial, industrial and national interest reasons, Australia has a need for a new reactor; a review of the present reactor, HIFAR, to include: an assessment of national and commercial benefits and costs of operations, its likely remaining useful life and its eventual closure and decommissioning; if Australia has a need for a new nuclear research reactor, the Review will consider: possible locations for a new reactor, its environmental impact at alternative locations, recommend a preferred location, and evaluate matters associated with regulation of the facility and organisational arrangements for reactor-based research. From the Review findings the following recommendations were stated: keep HIFAR going; commission a PRA to ascertain HIFAR's remaining life and refurbishment possibilities; identify and establish a HLW repository; accept that neither HIFAR nor a new reactor can be completely commercial; any decision on a new neutron source must rest primarily on benefits to science and Australia's national interest; make a decision on a new neutron source in about five years' time (1998). Design Proposals for a New Reactor are specified

  8. Safe operation of existing radioactive waste management facilities at Dalat Nuclear Research Institute

    International Nuclear Information System (INIS)

    Pham Van Lam; Ong Van Ngoc; Nguyen Thi Nang

    2000-01-01

    The Dalat Nuclear Research Reactor was reconstructed from the former TRIGA MARK-II in 1982 and put into operation in March 1984. The combined technology for radioactive waste management was newly designed and put into operation in 1984. The system for radioactive waste management at the Dalat Nuclear Research Institute (DNRI) consists of radioactive liquid waste treatment station and disposal facilities. The treatment methods used for radioactive liquid waste are coagulation and precipitation, mechanical filtering and ion- exchange. Near-surface disposal of radioactive wastes is practiced at DNRI In the disposal facilities eight concrete pits are constructed for solidification and disposal of low level radioactive waste. Many types of waste generated in DNRI and in some Nuclear Medicine Departments in the South of Vietnam are stored in the disposal facilities. The solidification of sludge has been done by cementation. Hydraulic compactor has done volume reduction of compatible waste. This paper presents fifteen-years of safe operation of radioactive waste management facilities at DNRI. (author)

  9. Safety considerations for research reactors in extended shutdown

    International Nuclear Information System (INIS)

    2004-01-01

    According to the IAEA Research Reactor Database, in the last 20 years, 367 research reactors have been shut down. Of these, 109 have undergone decommissioning and the rest are in extended shutdown with no clear definition about their future. Still other research reactors are infrequently operated with no meaningful utilization programme. These two situations present concerns related to safety such as loss of corporate memory, personnel qualification, maintenance of components and systems and preparation and maintenance of documentation. There are many reasons to shut down a reactor; these may include: - the need to carry out modifications in the reactor systems; - the need for refurbishment to extend the lifetime of the reactor; - the need to repair reactor structures, systems, or components; - the need to remedy technical problems; - regulatory or public concerns; - local conflicts or wars; - political convenience; - the lack of resources. While any one of these reasons may lead to shutdown of a reactor, each will present unique problems to the reactor management. The large variations from one research reactor to the next also will contribute to the uniqueness of the problems. Any option that the reactor management adopts will affect the future of the facility. Options may include dealing with the cause of the shutdown and returning to normal operation, extending the shutdown period waiting a future decision, or decommissioning. Such options are carefully and properly analysed to ensure that the solution selected is the best in terms of reactor type and size, period of shutdown and legal, economic and social considerations. This publication provides information in support of the IAEA safety standards for research reactors

  10. Usage of burnable poison on research reactors

    International Nuclear Information System (INIS)

    Villarino, Eduardo Anibal

    2002-01-01

    The fuel assemblies with burnable poison are widely used on power reactors, but there are not commonly used on research reactors. This paper shows a neutronic analysis of the advantages and disadvantages of the burnable poison usage on research reactors. This paper analyses both burnable poison design used on research reactors: Boron on the lateral wall and Cadmium wires. Both designs include a parametric study on the design parameters like the amount and geometry of the burnable poison. This paper presents the design flexibility using burnable poisons, it does not find an optimal or final design, which it will strongly depend on the core characteristics and fuel management strategy. (author)

  11. Establishing a Radiation Protection Programme for a Research Reactor

    International Nuclear Information System (INIS)

    Abdallah, M. M.

    2014-04-01

    The nature and intensity of radiation from the operation of a research reactor depend on the type of reactor, its design features and its operational history. The protection of workers from the harmful effect of radiation must therefore be of paramount importance to any operating organization of a research reactor. This project report attempts to establish an operational radiation protection programme for a research reactor using the Ghana Research Reactor-1 as a case study. (au)

  12. Study on the decommissioning of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Doo Hwan; Jun, Kwan Sik; Choi, Yoon Dong; Lee, Tae Yung; Kwon, Sang Woon; Lee, Jong Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

    Currently, KAERI operates TRIGA Mark-II and TRIGA Mark-III research reactors as a general purpose research and training facility. As these are, however, situated at Seoul office site of KAERI which is scheduled to be transferred to KEPCO as well as 30 MW HANARO research reactor which is expected to reach the first criticality in 1995 is under construction at head site of KAERI, decommissioning of TRIGA reactors has become an important topic. The objective of this study is to prepare and present TRIGA facility decontamination and decommissioning plan. Estimation of the radioactive inventory in TRIGA research reactor was carried out by the use of computational method. In addition, summarized in particular were the methodologies associated with decontamination, segmenting processes for activated metallic components, disposition of wastes. Particular consideration in this study was focused available technology applicable to decommissioning of TRIGA research reactor. State-of-the-art summaries of the available technology for decommissioning presented here will serve a useful document for preparations for decommissioning in the future. 6 figs, 41 tabs, 30 refs. (Author).

  13. Ageing Management for Research Reactors. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    This Safety Guide was developed under the IAEA programme for safety standards for research reactors, which covers all the important areas of research reactor safety. It supplements and elaborates upon the safety requirements for ageing management of research reactors that are established in paras 6.68-6.70 and 7.109 of the IAEA Safety Requirements publication, Safety of Research Reactors. The safety of a research reactor requires that provisions be made in its design to facilitate ageing management. Throughout the lifetime of a research reactor, including its decommissioning, ageing management of its structures, systems and components (SSCs) important to safety is required, to ensure continued adequacy of the safety level, reliable operation of the reactor, and compliance with the operational limits and conditions. Managing the safety aspects of research reactor ageing requires implementation of an effective programme for the monitoring, prediction, and timely detection and mitigation of degradation of SSCs important to safety, and for maintaining their integrity and functional capability throughout their service lives. Ageing management is defined as engineering, operation, and maintenance strategy and actions to control within acceptable limits the ageing degradation of SSCs. Ageing management includes activities such as repair, refurbishment and replacement of SSCs, which are similar to other activities carried out at a research reactor in maintenance and testing or when a modification project takes place. However, it is important to recognize that effective management of ageing requires the use of a methodology that will detect and evaluate ageing degradation as a consequence of the service conditions, and involves the application of countermeasures for prevention and mitigation of ageing degradation. The objective of this Safety Guide is to provide recommendations on managing ageing of SSCs important to safety at research reactors on the basis of international

  14. Ageing Management for Research Reactors. Specific Safety Guide

    International Nuclear Information System (INIS)

    2010-01-01

    This Safety Guide was developed under the IAEA programme for safety standards for research reactors, which covers all the important areas of research reactor safety. It supplements and elaborates upon the safety requirements for ageing management of research reactors that are established in paras 6.68-6.70 and 7.109 of the IAEA Safety Requirements publication, Safety of Research Reactors. The safety of a research reactor requires that provisions be made in its design to facilitate ageing management. Throughout the lifetime of a research reactor, including its decommissioning, ageing management of its structures, systems and components (SSCs) important to safety is required, to ensure continued adequacy of the safety level, reliable operation of the reactor, and compliance with the operational limits and conditions. Managing the safety aspects of research reactor ageing requires implementation of an effective programme for the monitoring, prediction, and timely detection and mitigation of degradation of SSCs important to safety, and for maintaining their integrity and functional capability throughout their service lives. Ageing management is defined as engineering, operation, and maintenance strategy and actions to control within acceptable limits the ageing degradation of SSCs. Ageing management includes activities such as repair, refurbishment and replacement of SSCs, which are similar to other activities carried out at a research reactor in maintenance and testing or when a modification project takes place. However, it is important to recognize that effective management of ageing requires the use of a methodology that will detect and evaluate ageing degradation as a consequence of the service conditions, and involves the application of countermeasures for prevention and mitigation of ageing degradation. The objective of this Safety Guide is to provide recommendations on managing ageing of SSCs important to safety at research reactors on the basis of international

  15. Some considerations for assurance of reactor safety from experiences in research reactors

    International Nuclear Information System (INIS)

    Okamoto, Sunao; Nishihara, Hideaki; Shibata, Toshikazu

    1981-01-01

    For the purpose of assuring reactor safety and strengthening research in the related fields, a multi-disciplinary group was formed among university researchers, including social scientists, with a special allocation of the Grant-in-Aid from the Ministry of Education, Science and Culture. An excerpt from the first year's report (1979 -- 1980) is edited here, which contains an interpretation of Murphy's reliability engineering law, a scope of reactor diagnostic studies to be pursued at universities, and safety measures already implemented or suggested to be implemented in university research reactors. (author)

  16. Towards an ethics safe harbor for global biomedical research

    Science.gov (United States)

    Dove, Edward S.; Knoppers, Bartha M.; Zawati, Ma'n H.

    2014-01-01

    Although increasingly global, data-driven genomics and other ‘omics’-focused research hold great promise for health discoveries, current research ethics review systems around the world challenge potential improvements in human health from such research. To overcome this challenge, we propose a ‘Safe Harbor Framework for International Ethics Equivalency’ that facilitates the harmonization of ethics review of specific types of data-driven international research projects while respecting globally transposable research ethics norms and principles. The Safe Harbor would consist in part of an agency supporting an International Federation for Ethics Review (IFER), formed by a voluntary compact among countries, granting agencies, philanthropies, institutions, and healthcare, patient advocacy, and research organizations. IFER would be both a central ethics review body, and also a forum for review and follow-up of policies concerning ethics norms for international research projects. It would be built on five principle elements: (1) registration, (2) compliance review, (3) recognition, (4) monitoring and enforcement, and (5) public participation. The Safe Harbor would create many benefits for researchers, countries, and the general public, and may eventually have application beyond (gen)omics to other areas of biomedical research that increasingly engage in secondary use of data and present only negligible risks. PMID:27774154

  17. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    International Nuclear Information System (INIS)

    Salam, M. A.

    2013-01-01

    The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

  18. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Salam, M. A. [Atomic Energy Research Establishment, Dhaka (Bangladesh)

    2013-07-01

    The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

  19. Safety evaluation of the Dalat research reactor operation

    International Nuclear Information System (INIS)

    Long, V.H.; Lam, P.V.; An, T.K.

    1989-01-01

    After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

  20. Enhancement of research reactor utilization in the developing countries

    International Nuclear Information System (INIS)

    Bashir, J.; Butt, N.M.

    1994-06-01

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

  1. Research reactor utilization in chemistry programmes

    International Nuclear Information System (INIS)

    Bautista, E.

    1983-01-01

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

  2. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

    2006-04-15

    RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

  3. The first university research reactor in India

    International Nuclear Information System (INIS)

    Murthy, G.S.

    1999-01-01

    At low power research reactor is being set up in Andhra University to cater to the needs of researchers and isotope users by the Department of Atomic Energy in collaboration with Andhra University. This reactor is expected to be commissioned by 2001-02. Departments like Chemistry, Earth Sciences, Physics, Life Sciences, Pharmacy, Medicine and Engineering would be the beneficiaries of the availability of this reactor. In this paper, details of the envisaged research programme and training activities are discussed. (author)

  4. Proceedings of the European Research Reactor Conference - RRFM 2013 Transactions

    International Nuclear Information System (INIS)

    2013-01-01

    In 2013 RRFM, the European Research Reactor Conference is jointly organised by ENS and Atomexpo LLC. This time the Research Reactor community meet in St. Petersburg, Russia. The conference programme will revolve around a series of Plenary Sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions will focus on all areas of the Fuel Cycle of Research Reactors, their Utilisation, Operation and Management as well as specific research projects and innovative methods in research reactor analysis and design. In 2013 the European Research Reactor Conference will for the first time give special attention to complementary safety assessments of Research Reactors, following the Fukushima-Dai-Ichi NPP's Accident. (authors)

  5. Review of Operation and Maintenance Support Systems for Research Reactors

    International Nuclear Information System (INIS)

    Jin, Kyungho; Heo, Gyunyoung; Park, Jaekwan

    2014-01-01

    Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

  6. Testing of research reactor fuel in the high flux reactor (Petten)

    International Nuclear Information System (INIS)

    Guidez, J.; Markgraf, J.W.; Sordon, G.; Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.

    1999-01-01

    The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium ( 3 . However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

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

    International Nuclear Information System (INIS)

    1967-01-01

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

  8. Factors affecting nuclear research reactor utilization across countries

    International Nuclear Information System (INIS)

    Hien, P.D.

    2000-01-01

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

  9. Status of Dalat research reactor and progress of new reactor plan in Vietnam

    International Nuclear Information System (INIS)

    Dien, Nguyen Nhi; Vien, Luong Ba

    2005-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a 500-kW pool-type reactor loaded with the Soviet WWR-M2 Fuel Assemblies (FA), moderated and cooled by light water. The reactor was reconstructed from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The first criticality of the renovated reactor was achieved on 1 st November 1983, and then on 20 March 1984 the reactor was officially inaugurated and its activities restarted. During the last twenty years, the DNRR has played an important role as a large national research facility to implement researches and applications, and its utilization has been broadened in various fields of human life. However, due to the limitation of the neutron flux and power level, the out-of date design of the experimental facilities and the ageing of the reactor facilities, it cannot meet the increasing user's demands even in the existing utilization areas. In addition, the utilization demands of the Research Reactor (RR) will be increased along with the development of the nation's economy growth. In this aspect, it is necessary to have in Vietnam a new high performance multipurpose RR with a sufficient neutron flux and power level. According to the last draft of a national strategy for atomic energy development submitted to the Government for consideration and approval, it is expected that a new high power RR would be put into operation before 2020. The operation and utilization status of the DNRR is presented and some preliminary results of the national research project on new reactor plan for Vietnam are discussed in this paper

  10. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.; Horlock, K.

    2001-01-01

    The contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000. This was followed by the completion of the detailed design and an application for a construction licence was made in May 2001. This paper will describe the main elements of the design and their relation to the proposed applications of the reactor. The future stages in the project leading to full operation are also described

  11. Research reactors spent fuel management in the Nuclear Research Institute Rez

    International Nuclear Information System (INIS)

    Rychecky, J.

    2001-01-01

    In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15

  12. Research reactors spent fuel management in the Nuclear Research Institute Rez

    Energy Technology Data Exchange (ETDEWEB)

    Rychecky, J. [Nuclear Research Institute, 25068 Rez (Czech Republic)

    2001-07-01

    In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15.

  13. MAPLE research reactor beam-tube performance

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Gillespie, G.E.

    1989-05-01

    Atomic Energy of Canada Limited (AECL) has been developing the MAPLE (Multipurpose Applied Physics Lattice Experimental) reactor concept as a medium-flux neutron source to meet contemporary research reactor applications. This paper gives a brief description of the MAPLE reactor and presents some results of computer simulations used to analyze the neutronic performance. The computer simulations were performed to identify how the MAPLE reactor may be adapted to beam-tube applications such as neutron radiography

  14. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    International Nuclear Information System (INIS)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form

  15. Regulatory Framework for Controlling the Research Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2009-01-01

    Decommissioning is one of important stages in construction and operation of research reactors. Currently, there are three research reactors operating in Indonesia. These reactors are operated by the National Nuclear Energy Agency (BATAN). The age of the three research reactors varies from 22 to 45 years since the reactors reached their first criticality. Regulatory control of the three reactors is conducted by the Nuclear Energy Regulatory Agency (BAPETEN). Controlling the reactors is carried out based on the Act No. 10/1997 on Nuclear Energy, Government Regulations and BAPETEN Chairman Decrees concerning the nuclear safety, security and safeguards. Nevertheless, BAPETEN still lack of the regulation, especially for controlling the decommissioning project. Therefore, in the near future BAPETEN has to prepare the regulations for decommissioning, particularly to anticipate the decommissioning of the oldest research reactors, which probably will be done in the next ten years. In this papers author give a list of regulations should be prepared by BAPETEN for the decommissioning stage of research reactor in Indonesia based on the international regulatory practice

  16. Research on reactor physics data

    International Nuclear Information System (INIS)

    1961-01-01

    In the early years of nuclear reactor research, each national program tended to develop its own reactor physics information. The Government of Norway proposed to the Agency the undertaking of a joint program in reactor physics utilizing the facilities and staff of its zero power reactor NORA then under construction. Following the approval by the Board of Governors in February, the Agency invited Member States to submit the names and qualifications of scientists they wished to suggest for the project. All the results and information gained through the program, which is expected to last about three years, will be placed at the disposal of the Agency's Member States

  17. Applications of Research Reactors

    International Nuclear Information System (INIS)

    2014-01-01

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The purpose of the earlier publication, The Application of Research Reactors, IAEA-TECDOC-1234, was to present descriptions of the typical forms of research reactor use. The necessary criteria to enable an application to be performed were outlined for each one, and, in many cases, the minimum as well as the desirable requirements were given. This revision of the publication over a decade later maintains the original purpose and now specifically takes into account the changes in service requirements demanded by the relevant stakeholders. In particular, the significant improvements in

  18. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    International Nuclear Information System (INIS)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data

  19. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  20. Research report on the users' needs for next research reactor

    International Nuclear Information System (INIS)

    Takahashi, Hiroyuki; Tamura, Itaru; Hosoya, Toshiaki; Horiguchi, Hironori

    2015-03-01

    JRR-3 has been operated for more than 25 years for that it is time to investigate the role of a next research reactor. A task force under the Committee for Promotion of JRR-3 Neutron Beam Application has been organized by Department of Research Reactor and Tandem Accelerator to survey neutron beam application trends in the future. This is a report on the survey results and users' requirements for the next research reactor have been summarized in this report carried by the task force. (author)

  1. IAEA Activities supporting education and training at research reactors

    International Nuclear Information System (INIS)

    Peld, N.D.; Ridikas, D.

    2013-01-01

    Full-text: Through the provision of neutrons for experiments and their historical association with universities, research reactors have played a prominent role in nuclear education and training of students, scientists and radiation workers. Today education and training remains the foremost application of research reactors, involving close to 160 facilities out of 246 operational. As part of its mandate to facilitate and expand the contribution of atomic energy to peace, health and prosperity throughout the world, the IAEA administers a number of activities intended to promote nuclear research and enable access to nuclear technology for peaceful purposes, one of which is the support of various education and training measures involving research reactors. In the last 5 years, education and training has formed one pillar for the creation of research reactor coalitions and networks to pool their resources and offer joint programmes, such as the on-going Group Fellowship Training Course. Conducted mainly through the Eastern European Research Reactor Initiative, this programme is a periodic sic week course for young scientists and engineers on nuclear techniques and administration jointly conducted at several member research reactor institutes. Organization of similar courses is under consideration in Latin America and the Asia-Pacific Region, also with support from the IAEA. Additionally, four research reactor institutes have begun offering practical education courses through virtual reactor experiments and operation known as the Internet Reactor Laboratory. Through little more than an internet connection and projection screens, university science departments can be connected regionally or bilaterally with the control room o a research reactor for various training activities. Finally, two publications are being prepared, namely Hands-On Training Courses Using Research Reactors and Accelerators, and Compendium on Education and training Based on Research Reactors. These

  2. Operation and utilizations of Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    Hien, P.Z.

    1988-01-01

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

  3. International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13). Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The conference, which was held from 4 to 7 of March 2013 in Paris, provided a forum to exchange information on national and international programmes, and more generally new developments and experience, in the field of fast reactors and related fuel cycle technologies. A first goal was to identify and discuss strategic and technical options that have been proposed by individual countries or companies. Another goal was to promote the development of fast reactors and related fuel cycle technologies in a safe, proliferation resistant and economic way. A third goal was to identify gaps and key issues that need to be addressed in relation to the industrial deployment of fast reactors with a closed fuel cycle. A fourth goal was to engage young scientists and engineers in this field, in particular with sustainability, innovation, simulation, safety, economics and public acceptance

  4. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  5. New trends for future reactors. A research and development review

    International Nuclear Information System (INIS)

    Anzieu, P.

    2002-01-01

    Third generation reactors proposed to the market are mostly LWR, pressurized or boiling, with confirmed competitiveness. A special effort to increase the safety level is sensible and should be improved. At least, solutions are studied to better use plutonium. The development of a new generation of NPPs offers opportunity to have another step towards more safety, for example in being fail-safe, and towards a minimization of ultimate waste produced. In this field, CEA dedicates its main effort to the development of a gas cooled reactor and constraint on safety, waste minimization are indicated. At least some examples of progression in the safety level of a plant are shown from an existing one to an hypothetical future reactor

  6. Status of research reactor spent fuel world-wide

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    2004-01-01

    Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialised and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author)

  7. Digital computer operation of a nuclear reactor

    International Nuclear Information System (INIS)

    Colley, R.W.

    1984-01-01

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state

  8. Research Reactors Coalitions

    International Nuclear Information System (INIS)

    Borio, Andrea; Bradley, E.; Vyshniauskas, Judy; Ridikas, Danas

    2013-01-01

    When considering the potential role of an existing RR or possibly the construction of a new RR, it is clear that a nuclear science and technology programme (including nuclear power) could benefit provided the RR is safely and competently managed, well utilised and adequately funded. Based on MSs experience, a domestic RR may not be required to develop a nuclear power programme, provided the decision takes advantage of foreign expertise, including access to foreign RRs facilities and RRs regional/international networks. If a country decides to gain access to a foreign research reactor, it may need considering the potential risk of change in the political relationship with the host country that could compromise the achievement of its national relevant objectives. This risk may be offset by availability of many options within one or more regional/international RRs networks and coalitions. Examples include the use of existing RRs in vendor, non-vendor countries and, in some cases non-nuclear power countries, to develop human resources in support of the introduction of nuclear power elsewhere. International RR networking trends are most evident with high flux, higher capability, and more complex fuel and material testing RRs being shared through international partnerships. However, networks involving low-medium power RRs for education and training purposes are also gaining a more prominent role to support nuclear capacity building in newcomer MSs. Networking through the internet seems also to be a promising way to support, as complementary offer to direct access to RRs facilities, MSs nuclear capacity building objectives (e.g. the IRL project)

  9. Future plans on the Kyoto University Research Reactor (KUR)

    International Nuclear Information System (INIS)

    Shibata, Seiichi

    2000-01-01

    The Research Reactor Institute (RRI), Kyoto University, for aiming at performing the 'Experiments using a reactor and its related research', was established in Showa 38 (1963) as a cooperative research institute for universities and so on in allover Japan. Operation using KUR of one of main facilities in RRI was started by 1 MW of its rated output in 1964, and converted to 5 MW in 1968, after which through development , addition and modification of various research apparatus it has been proposed to the cooperative application researches with universities and so on in allover Japan, hitherto. Among these periods, its research organization is improved to six departments containing twenty divisions and two attached research facilities to progress some investigations on future plans at RRI for response to new researching trends. Here were described on present state of research on use of low concentrated uranium fuels at research reactor, and future plans on neutron factory and hybrid reactor. The former aims at establishment of a new research facility capable of alternating to KUR for future academic research on research reactor containing high quality and high degree application of neutron field and safety management and feature upgrading of nuclear energy. And, the latter aims at development on an accelerator drive uncritical reactor combined an accelerator neutron source and an uncritical reactor. (G.K.)

  10. Research reactor status for future nuclear research in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Raymond, Patrick; Bignan, Gilles; Guidez, Joel [Commissariat a l' Energie Atomique - CEA (France)

    2010-07-01

    During the 1950's and 60's, the European countries built several research reactors, partially to support their emerging nuclear-powered electricity programs. Now, over forty years later, the use and operation of these reactors have both widened and grown more specialized. The irradiation reactors test materials and fuels for power reactors, produce radio-isotopes for medicine, neutro-graphies, doping silicon, and other materials. The neutron beam reactors are crucial to science of matter and provide vital support to the development of nano-technologies. Other reactors are used for other specialized services such as teaching, safety tests, neutron physics measurements... The modifications to the operating uses and the ageing of the nuclear facilities have led to increasing closures year after year. Since last ENC, for example, we have seen, only in France, the closure of the training reactor Ulysse in 2007, the closure of the safety test dedicated reactor Phebus in 2008 and recently the Phenix reactor, last fast breeder in operation in the European Community, has been shut down after a set of 'end of life' technological and physical tests. For other research reactors, safety re-evaluations have had to take place, to enable extension of reactor life. However, in the current context of streamlining and reorganization, new European tools have emerged to optimally meet the changing demands for research. However the operation market of these reactors seems now increasing in all fields. For the neutron beams reactors (FRMII, ORPHEE, ILL, ISIS,..) the experimental needs are increasing years after years, especially for nano sciences and bio sciences new needs. The measurement of residual stress on manufactured materials is also more and more utilised. All these reactors have increasing utilizations, and their future seems promising. A new project project based on a neutron spallation is under definition in Sweden (ESSS: European Spallation Source

  11. Research reactor status for future nuclear research in Europe

    International Nuclear Information System (INIS)

    Raymond, Patrick; Bignan, Gilles; Guidez, Joel

    2010-01-01

    During the 1950's and 60's, the European countries built several research reactors, partially to support their emerging nuclear-powered electricity programs. Now, over forty years later, the use and operation of these reactors have both widened and grown more specialized. The irradiation reactors test materials and fuels for power reactors, produce radio-isotopes for medicine, neutro-graphies, doping silicon, and other materials. The neutron beam reactors are crucial to science of matter and provide vital support to the development of nano-technologies. Other reactors are used for other specialized services such as teaching, safety tests, neutron physics measurements... The modifications to the operating uses and the ageing of the nuclear facilities have led to increasing closures year after year. Since last ENC, for example, we have seen, only in France, the closure of the training reactor Ulysse in 2007, the closure of the safety test dedicated reactor Phebus in 2008 and recently the Phenix reactor, last fast breeder in operation in the European Community, has been shut down after a set of 'end of life' technological and physical tests. For other research reactors, safety re-evaluations have had to take place, to enable extension of reactor life. However, in the current context of streamlining and reorganization, new European tools have emerged to optimally meet the changing demands for research. However the operation market of these reactors seems now increasing in all fields. For the neutron beams reactors (FRMII, ORPHEE, ILL, ISIS,..) the experimental needs are increasing years after years, especially for nano sciences and bio sciences new needs. The measurement of residual stress on manufactured materials is also more and more utilised. All these reactors have increasing utilizations, and their future seems promising. A new project project based on a neutron spallation is under definition in Sweden (ESSS: European Spallation Source Scandinavia). The nuclear

  12. International collaboration between nuclear research centres and the role of research reactors

    International Nuclear Information System (INIS)

    Dodd, B.

    2001-01-01

    A research reactor is a core facility in many nuclear research centres (NRCs) of Member States and it is logical that it should be the focus of any international collaboration between such centres. There are several large and sophisticated research reactors in operation in both developed and developing Member States, such as Belgium, China, Egypt, France, Hungary, Indonesia, India, Japan, ROK, Netherlands, South Africa and the USA. There are also several new, large reactors under construction or being planned such as those in Australia, Canada, China, France, Germany, and Thailand. It is felt that the utilization of these reactors can be enhanced by international co-operation to achieve common goals in research and applications. (author)

  13. Simulation of a pool type research reactor

    International Nuclear Information System (INIS)

    Oliveira, Andre Felipe da Silva de; Moreira, Maria de Lourdes

    2011-01-01

    Computational fluid dynamic is used to simulate natural circulation condition after a research reactor shutdown. A benchmark problem was used to test the viability of usage such code to simulate the reactor model. A model which contains the core, the pool, the reflector tank, the circulation pipes and chimney was simulated. The reactor core contained in the full scale model was represented by a porous media. The parameters of porous media were obtained from a separate CFD analysis of the full core model. Results demonstrate that such studies can be carried out for research and test of reactors design. (author)

  14. Application of JAERI research reactors to education

    International Nuclear Information System (INIS)

    Ogawa, Shigeru; Morozumi, Minoru

    1987-01-01

    At the dawning of the atomic age in Japan, training on reactor operation and reactor engineering experiments has been started in 1958 using JRR-1 (a 50 kW water boiler type reactor with liquid fuel), which was the first research reactor in Japan. The role of the training has been transferred to JRR-4 (a 3500 kW swimming pool type reactor with ETR type fuel) since 1969 due to the decommission of JRR-1. The training courses which have been held are: JRR-1 Short-Term Course for Operation (1958 ∼ 1963) General Course (1961 ∼ ) Reactor Engineering Course (1976 ∼ ) Training Course in Nuclear Technology (International course)(1986 ∼ ). And individual training concerning research reactors for the participants of scientist exchange program sponsored by Science and Technology Agency and of bilateral agreement have been initiated in 1985. The graduates of these courses work as staff members in various fields in nuclear industry. (author)

  15. Status report of Indonesian research reactor

    International Nuclear Information System (INIS)

    Arbie, B.; Supadi, S.

    1992-01-01

    A general description of three Indonesian research reactor, its irradiation facilities and its future prospect are described. Since 1965 Triga Mark II 250 KW Bandung, has been in operation and in 1972 the design powers were increased to 1000 KW. Using core grid form Triga 250 KW BATAN has designed and constructed Kartini Reactor in Yogyakarta which started its operation in 1979. Both of this Triga type reactors have served a wide spectrum of utilization such as training manpower in nuclear engineering, radiochemistry, isotope production and beam research in solid state physics. Each of this reactor have strong cooperation with Bandung Institute of Technology at Bandung and Gajah Mada University at Yogyakarta which has a faculty of Nuclear Engineering. Since 1976 the idea to have high flux reactor has been foreseen appropriate to Indonesian intention to prepare infrastructure for nuclear industry for both energy and non-energy related activities. The idea come to realization with the first criticality of RSG-GAS (Multipurpose Reactor G.A. Siwabessy) in July 1987 at PUSPIPTEK Serpong area. It is expected that by early 1992 the reactor will reached its full power of 30 MW and by end 1992 its expected that irradiation facilities will be utilized in the future for nuclear scientific and engineering work. (author)

  16. Preservation of the first research nuclear reactor in Korea

    International Nuclear Information System (INIS)

    2008-06-01

    This book describes preservation of the first research nuclear reactor in Korea and necessity of building memorial hall, sale of the Institute of Atomic Energy Research in Seoul and dismantlement of the first and the second nuclear reactor, preservation of the first research nuclear reactor and activity about memorial hall of the atomic energy reactor, assignment and leaving the report, and the list of related data.

  17. Research reactor programmes at the IAEA

    International Nuclear Information System (INIS)

    Reijonen, H.

    1978-01-01

    The activities performed according to the Agency programs for research reactors in the fields of information collection and dissemination, meetings organization, publications of the proceedings and execution of technical assistance are discussed in the paper emphasizing the services that are provided for developing countries. It is intended that the programme on research reactors should be flexible and respond to the actual needs of the countries receiving assistance

  18. RA Research reactor, Annual report 1988

    International Nuclear Information System (INIS)

    Sotic, O.

    1988-12-01

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

  19. RETU The Finnish research programme on reactor safety 1995-1998. Final Symposium

    International Nuclear Information System (INIS)

    Vanttola, T.

    1998-01-01

    The Reactor Safety (RETU, 1995-1998) research programme concentrated on search of safe limits for nuclear fuel and the reactor core, accident management methods and risk management of nuclear power plants. The total volume of the programme was 100 person years and funding FIM 58 million. This symposium report summarises the research fields, the objectives and the main results obtained. In the field of operational margins of a nuclear reactor, the behaviour of high burnup nuclear fuel was studied both in normal operation and during power transients. The static and dynamic reactor analysis codes were developed and validated to cope with new fuel designs and complicated three-dimensional reactivity transients. Advanced flow models and numerical solution methods for the dynamics codes were developed and tested. Research on accident management developed and validated calculation methods needed to plan preventive measures and to train the personnel to severe accident mitigation. Efforts were made to reduce uncertainties in phenomena important in severe accidents and to study actions planned for accident management. The programme included experimental work, but also participation in large international tests. The Finnish thermal-hydraulic test facility PACTEL was used extensively for the evaluation of the VVER-440 plant accident behaviour, for the validation of the accident analysis computer codes and for the testing of passive safety system concepts for future plant designs. In risk management probabilistic methods were developed for safety related decision making and for complex event sequences. Effects of maintenance on safety were studied and effective methods for assessment of human reliability and safety critical organisations were searched. To enhance human competencies in control of complex environments, practical tools for training and continuous learning were worked out, and methods to evaluate appropriateness of control room design were developed. (orig)

  20. Enabling Technologies for Ultra-Safe and Secure Modular Nuclear Energy

    International Nuclear Information System (INIS)

    Mendez Cruz, Carmen Margarita; Rochau, Gary E.; Middleton, Bobby; Rodriguez, Salvador B.; Rodriguez, Carmelo; Schleicher, Robert

    2016-01-01

    Sandia National Laboratories and General Atomics are pleased to respond to the Advanced Research Projects Agency-Energy (ARPA-e)'s request for information on innovative developments that may overcome various current reactor-technology limitations. The RFI is particularly interested in innovations that enable ultra-safe and secure modular nuclear energy systems. Our response addresses the specific features for reactor designs called out in the RFI, including a brief assessment of the current state of the technologies that would enable each feature and the methods by which they could be best incorporated into a reactor design.

  1. Enabling Technologies for Ultra-Safe and Secure Modular Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Mendez Cruz, Carmen Margarita [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rochau, Gary E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Middleton, Bobby [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rodriguez, Salvador B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rodriguez, Carmelo [General Atomics, San Diego, CA (United States); Schleicher, Robert [General Atomics, San Diego, CA (United States)

    2016-06-01

    Sandia National Laboratories and General Atomics are pleased to respond to the Advanced Research Projects Agency-Energy (ARPA-e)’s request for information on innovative developments that may overcome various current reactor-technology limitations. The RFI is particularly interested in innovations that enable ultra-safe and secure modular nuclear energy systems. Our response addresses the specific features for reactor designs called out in the RFI, including a brief assessment of the current state of the technologies that would enable each feature and the methods by which they could be best incorporated into a reactor design.

  2. Light water ultra-safe plant concept

    International Nuclear Information System (INIS)

    Klevans, E.

    1989-01-01

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

  3. IAEA/CRP for decommissioning techniques for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Won Zin; Won, H. J.; Kim, K. N.; Lee, K. W.; Jung, C. H

    2001-03-01

    The following were studied through the project entitled 'IAEA/CRP for decommissioning techniques for research reactors 1. Decontamination technology development for TRIGA radioactive soil waste - Electrokinetic soil decontamination experimental results and its mathematical simulation 2. The 2nd IAEA/CRP for decommissioning techniques for research reactors - Meeting results and program 3. Hosting the 2001 IAEA/RCA D and D training course for research reactors and small nuclear facilities.

  4. IAEA/CRP for decommissioning techniques for research reactors

    International Nuclear Information System (INIS)

    Oh, Won Zin; Won, H. J.; Kim, K. N.; Lee, K. W.; Jung, C. H.

    2001-03-01

    The following were studied through the project entitled 'IAEA/CRP for decommissioning techniques for research reactors 1. Decontamination technology development for TRIGA radioactive soil waste - Electrokinetic soil decontamination experimental results and its mathematical simulation 2. The 2nd IAEA/CRP for decommissioning techniques for research reactors - Meeting results and program 3. Hosting the 2001 IAEA/RCA D and D training course for research reactors and small nuclear facilities

  5. Code structure for U-Mo fuel performance analysis in high performance research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Cho, Tae Won; Lee, Chul Min; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A performance analysis modeling applicable to research reactor fuel is being developed with available models describing fuel performance phenomena observed from in-pile tests. We established the calculation algorithm and scheme to best predict fuel performance using radio-thermo-mechanically coupled system to consider fuel swelling, interaction layer growth, pore formation in the fuel meat, and creep fuel deformation and mass relocation, etc. In this paper, we present a general structure of the performance analysis code for typical research reactor fuel and advanced features such as a model to predict fuel failure induced by combination of breakaway swelling and pore growth in the fuel meat. Thermo-mechanical code dedicated to the modeling of U-Mo dispersion fuel plates is being under development in Korea to satisfy a demand for advanced performance analysis and safe assessment of the plates. The major physical phenomena during irradiation are considered in the code such that interaction layer formation by fuel-matrix interdiffusion, fission induced swelling of fuel particle, mass relocation by fission induced stress, and pore formation at the interface between the reaction product and Al matrix.

  6. Ten years of IAEA cooperation with the Russian research reactor fuel return programme

    Energy Technology Data Exchange (ETDEWEB)

    Tozser, S.; Adelfang, P.; Bradley, E. [International Atomic Energy Agency, Vienna (Austria)

    2013-01-15

    The Russian Research Reactor Fuel Return (RRRFR) Programme was launched in 2001. Over the duration, the programme successfully completed 43 safe shipments of 1.6 tons of fresh and spent HEU fuel from different countries using Russian fuelled research reactors to the country of origin. The IAEA has been a very active supporter of the RRRFR Programme since its inception. Under the auspices of the RRRFR Programme, the Agency has been ensuring a broad range of technical advisory and organizational support to the HEU fuel repatriation, as well as training and advisory assistance for supporting RR conversion from HEU to LEU. The presentation gives an overview of the RRRFR programme achievements with special consideration of the IAEA contribution. These include an overview of the shipments' history in terms of fresh and spent fuel, as well as a summary of experiences gained during the shipments' preparation and termination. The presentation focuses on technical advisory support given by the IAEA during the programme implementation, captures the consolidated knowledge of the unique international programme and shares the most important lessons learned. (orig.)

  7. The heavy water accountancy for research reactors in JAERI

    International Nuclear Information System (INIS)

    Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

    1998-11-01

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

  8. Strategic Planning for Research Reactors

    International Nuclear Information System (INIS)

    2017-01-01

    This publication is a revision of IAEA-TECDOC-1212 which primarily focused on enhancing the utilization of existing research reactors. This updated version also provides guidance on how to develop and implement a strategic plan for a new research reactor project and will be of particular interest for organizations which are preparing a feasibility study to establish such a new facility. This publication will enable managers to determine more accurately the actual and potential capabilities of an existing reactor, or the intended purpose and type of a new facility. At the same time, management will be able to match these capabilities to stakeholders/users’ needs and establish the strategy of meeting such needs. In addition, several annexes are presented, including some examples as clarification to the main text and ready-to-use templates as assistance to the team drafting a strategic plan.

  9. Nuclear research reactors in the world. May 1987 ed.

    International Nuclear Information System (INIS)

    1987-01-01

    This is the second edition of Reference Data Series No.3, Nuclear Research Reactors in the World, which replaces the Agency's publications Power and Research Reactors in Member States and Research Reactors in Member States. This booklet contains general information, as of the end of May 1987, on research reactors in operation, under construction, planned, and shut down. The information is collected by the Agency through questionnaires sent to the Member States through the designated national correspondents. 11 figs, 19 tabs

  10. Self Assessment for the Safety of Research Reactor in Indonesia

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2008-01-01

    At the present Indonesia has no nuclear power plant in operation yet, although it is expected that the first nuclear power plant will be operated and commercially available in around the year of 2016 to 2017 in Muria Peninsula. National Nuclear Energy Agency (BATAN) has three research reactor; which are: Reactor Triga Mark II at Bandung, Reactor Kartini at Yogyakarta and Reactor Serbaguna (Multi Purpose Reactor) at Serpong. The Code of Conduct on the Safety of Research Reactors establishes 'best practice' guidelines for the licensing, construction and operation of research reactors. In this paper the author use the requirement in code of conduct to review the safety of research reactor in Indonesia

  11. Preparation fo nuclear research reactors operators

    International Nuclear Information System (INIS)

    Roedel, G.

    1986-01-01

    The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

  12. Preparation of nuclear research reactors operators

    International Nuclear Information System (INIS)

    Roedel, G.

    1986-01-01

    The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN, are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

  13. Future directions of small research reactors

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Rack, E.P.

    1986-01-01

    In prognosticating future perspectives, it is important to realize that the current number of small reactors throughout the world is not overly large and will undoubtedly decrease or at best remain constant in future generations. To survive and remain productive, small reactor facilities must concentrate on work that is unique and that cannot be performed as well by other instruments. Wherever possible, these facilities should develop some form of collaboration with universities and medical center investigators. Future development will continue and will flourish in neutron activation analysis and its applications for a diversity of fields. Fundamental research such as hot atom chemistry will continue to use neutrons from small research reactors. Finally, training of power reactor operators can be an important source of revenue for the small facility in addition to performing an important service to the nuclear power industry

  14. Towards intrinsically safe light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hannerz, K

    1983-07-01

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

  15. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Hannerz, K.

    1983-07-01

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

  16. Basic research using the 250 KW research reactor triga in Ljubljana, Yugoslavia

    International Nuclear Information System (INIS)

    Dimic, V.

    1983-01-01

    The 25 KW Triga Mark II reactor of J. 'Stefan Institute' was commissioned on May 1966. During the last two years, it has been operated for about 4200 hr/year. According to experience gained with the reactor, most of the cost of reactor operation will be earned through isotope production for local hospitals and industries, performing low cost applied experiments and organizing training courses. The rest was provided through the Research Communities of the Republic of Slovenia. The reactor has been operated for 15 years without major problems and many basic research programmes have been performed. The research is being conducted in the following mainfields: solid state physics, neutron dosimetry, neutron radiography and autoradiography, reactor physics, examination of nuclear fuel using gamma scanning, irradiation of semiconducting materials and neutron activation analysis. (A.J)

  17. Advanced fuel in the Budapest research reactor

    International Nuclear Information System (INIS)

    Hargitai, T.; Vidovsky, I.

    1997-01-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  18. Management of operational events in research reactor

    International Nuclear Information System (INIS)

    Zhong Heping; Yang Shuchun; Peng Xueming

    2001-01-01

    The author describes the tracing management process post-operational event in a research reactor based on nuclear safety code, under the background of the research reactor in Nuclear Power Institute of China. It presorts the definite measures to the event tracing and it up its management factors

  19. Cost estimation for decommissioning of research reactors

    International Nuclear Information System (INIS)

    Grossi, Pablo Andrade; Tello, Cledola Cassia Oliveira de; Segabinaze, Roberto de Oliveira; Daniska, Vladimir

    2013-01-01

    In the case of research reactors, the limited data that is available tends to provide only overall decommissioning costs, without any breakdown of the main cost elements. In order to address this subject, it is important to collect and analyse all available data of decommissioning costs for the research reactors. The IAEA has started the DACCORD Project focused on data analysis and costing of research reactors decommissioning. Data collection is organized in accordance with the International Structure for Decommissioning Costing (ISDC), developed jointly by the IAEA, the OECD Nuclear Energy Agency and the European Commission. The specific aims of the project include the development of representative and comparative data and datasets for preliminary costing for decommissioning. This paper will focus on presenting a technique to consider several representative input data in accordance with the ISDC structure and using the CERREX (Cost Estimation for Research Reactors in Excel) software developed by IAEA. (author)

  20. Safety upgrades to the NRU research reactor

    International Nuclear Information System (INIS)

    DeAbreu, B.; Mark, J.M.; Mutterback, E.J.

    1998-01-01

    The NRU (National Research Universal) Reactor is a 135 MW thermal research facility located at Chalk River Laboratories, and is owned and operated by Atomic Energy of Canada Limited. One of the largest and most versatile research reactors in the world, it serves as the R and D workhorse for Canada's CANDU business while at the same time filling the role as one of the world's major producers of medical radioisotopes. AECL plans to extend operation of the NRU reactor to approximately the year 2005 when a new replacement, the Irradiation Research Facility (IRF) will be available. To achieve this, AECL has undertaken a program of safety reassessment and upgrades to enhance the level of safety consistent with modem requirements. An engineering assessment/inspection of critical systems, equipment and components was completed and seven major safety upgrades are being designed and installed. These upgrades will significantly reduce the reactor's vulnerability to common mode failures and external hazards, with particular emphasis on seismic protection. The scheduled completion date for the project is 1999 December at a cost approximately twice the annual operating cost. All work on the NRU upgrade project is planned and integrated into the regular operating cycles of the reactor; no major outages are anticipated. This paper describes the safety upgrades and discusses the technical and managerial challenges involved in extending the operating life of the NRU reactor. (author)

  1. On-line Monitoring of Instrumentation in Research Reactors

    International Nuclear Information System (INIS)

    2017-12-01

    This publication is the result of a benchmarking effort undertaken under the IAEA coordinated research project on improved instrumentation and control (I&C) maintenance techniques for research reactors. It lays the foundation for implementation of on-line monitoring (OLM) techniques and establishment of the validity of those for improved maintenance practices in research reactors for a number of applications such as change to condition based calibration, performance monitoring of process instrumentation systems, detection of process anomalies and to distinguish between process problems/effects and instrumentation/sensor issues. The techniques and guidance embodied in this publication will serve the research reactor community in providing the technical foundation for implementation of OLM techniques. It is intended to be used by Member States to implement I&C maintenance and to improve performance of research reactors.

  2. The present status and the prospect of China research reactors

    International Nuclear Information System (INIS)

    Yongmao, Z.; Yizheng, C.

    1990-01-01

    A total of 100 reactor operation years' experience of research reactors has now been obtained in China. The type and principal parameters of China research reactors and their operating status are briefly introduced in this paper. Chinese research reactors have been playing an important role in nuclear power and nuclear weapon development, industrial and agricultural production, medicine, basic and applied science research and environmental protection, etc. The utilization scale, benefits and achievements will be given. There is a good safety record in the operation of these reactors. A general safety review is discussed. The important incidents and accidents happening during a hundred reactor operating years are described and analyzed. China has the capability of developing any type of research reactor. The prospective projects are briefly introduced

  3. Inherently safe technologies-chemical and nuclear

    International Nuclear Information System (INIS)

    Weinberg, A.M.

    1984-01-01

    Probabilistic risk assessments show an inverse relationship between the likelihood and the consequences of nuclear and chemical plant accidents, but the Bhopal accident has change public complacency about the safety of chemical plants to such an extent that public confidence is now at the same low level as with nuclear plants. The nuclear industry's response was to strengthen its institutions and improve its technologies, but the public may not be convinced. One solution is to develop reactors which do not depend upon the active intervention of humans of electromechanical devices to deal with emergencies, but which have physical properties that limit the possible temperature and power of a reactor. The Process Inherent Ultimately Safe and the modular High-Temperature Gas-Cooled reactors are two possibilities. the chemical industry needs to develop its own inherently safe design precepts that incorporate smallness, safe processes, and hardening against sabotage. 5 references

  4. Proceedings of first SWCR-KURRI academic seminar on research reactors and related research topics

    International Nuclear Information System (INIS)

    Kimura, Itsuro; Cong, Zhebao

    1986-01-01

    These are the proceedings of an academic seminar on research reactors and related research topics held at the Southwest Centre for Reactor Engineering Research and Design in Chengdu, Sichuan, People's Republic of China in September 24-26 in 1985. Included are the chairmen's addresses and 10 papers presented at the seminar in English. The titles of these papers are: (1) Nuclear Safety and Safeguards, (2) General Review of Thorium Research in Japanese Universities, (3) Comprehensive Utilization and Economic Analysis of the High Flux Engineering Test Reactor, (4) Present States of Applied Health Physics in Japan, (5) Neutron Radiography with Kyoto University Reactor, (6) Topics of Experimental Works with Kyoto University Reactor, (7) Integral Check of Nuclear Data for Reactor Structural Materials, (8) The Reactor Core, Physical Experiments and the Operation Safety Regulation of the Zero Energy Thermal Reactor for PWR Nuclear Power Plant, (9) HFETR Core Physical Parameters at Power, (10) Physical Consideration for Loads of Operated Ten Cycles in HFETR. (author)

  5. Present status of research reactor decommissioning programme in Indonesia

    International Nuclear Information System (INIS)

    Suripto, A.; Mulyanto, N.

    2002-01-01

    At present Indonesia has 3 research reactors, namely the 30 MW MTR-type multipurpose reactor at Serpong Site, two TRIGA-type research reactors, the first one being 1 MW located at Bandung Site and the second one a small reactor of 100 kW at Yogyakarta Site. The TRIGA Reactor at the Bandung Site reached its first criticality at 250 kW in 1964, and then was operated at 1000 kW since 1971. In October 2000 the reactor power was successfully upgraded to 2 MW. This reactor has already been operated for 38 years. There is not yet any decision for the decommissioning of this reactor. However it will surely be an object for the near future decommissioning programme and hence anticipation for the above situation becomes necessary. The regulation on decommissioning of research reactor is already issued by the independent regulatory body (BAPETEN) according to which the decommissioning permit has to be applied by the BATAN. For Indonesia, an early decommissioning strategy for research reactor dictates a restricted re-use of the site for other nuclear installation. This is based on high land price, limited availability of radwaste repository site, and other cost analysis. Spent graphite reflector from the Bandung TRIGA reactor is recommended for a direct disposal after conditioning, without any volume reduction treatment. Development of human resources, technological capability as well as information flow from and exchange with advanced countries are important factors for the future development of research reactor decommissioning programme in Indonesia. (author)

  6. Safety status of Russian research reactors

    International Nuclear Information System (INIS)

    Morozov, S.I.

    2001-01-01

    Gosatomnadzor of Russia is conducting the safety regulation and inspection activity related to nuclear and radiation safety at nuclear research facilities, including research reactors, critical assemblies and sub-critical assemblies. It implies implementing three major activities: 1) establishing the laws and safety standards in the field of research reactors nuclear and radiation safety; 2) research reactors licensing; and 3) inspections (or license conditions tracking and inspection). The database on nuclear research facilities has recently been updated based on the actual status of all facilities. It turned out that many facilities have been shutdown, whether temporary or permanently, waiting for the final decision on their decommissioning. Compared to previous years the situation has been inevitably changing. Now we have 99 nuclear research facilities in total under Gosatomnadzor of Russia supervision (compared to 113 in previous years). Their distribution by types and operating organizations is presented. The licensing and conduct of inspection processes are briefly outlined with emphasis being made on specific issues related to major incidents that happened in 2000, spent fuel management, occupational exposure, effluents and emissions, emergency preparedness and physical protection. Finally, a summary of problems at current Russian research facilities is outlined. (author)

  7. Ageing management of the BR2 research reactor

    International Nuclear Information System (INIS)

    Verpoortem, J. R.; Van Dyck, S.

    2014-01-01

    At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

  8. IDAS-RR: an incident data base system for research reactors

    International Nuclear Information System (INIS)

    Matsumoto, Kiyoshi; Kohsaka, Atsuo; Kaminaga, Masanori; Murayama, Youji; Ohnishi, Nobuaki; Maniwa, Masaki.

    1990-03-01

    An Incident Data Base System for Research Reactors, IDAS-RR, has been developed. IDAS-RR has information about abnormal incidents (failures, transients, accidents, etc.) of research reactors in the world. Data reference, input, editing and other functions of IDAS-RR are menu driven. The routine processing and data base management functions are performed by the system software and hardware. PC-9801 equipment was selected as the hardware because of its portability and popularity. IDAS-RR provides effective reference information for the following activities. 1) Analysis of abnormal incident of research reactors, 2) Detail analysis of research reactor behavior in the abnormal incident for building the knowledge base of the reactor emergency diagnostic system for research reactor, 3) Planning counter-measure for emergency situation in the research reactor. This report is a user's manual of IDAS-RR. (author)

  9. Performance of a multipurpose research electrochemical reactor

    International Nuclear Information System (INIS)

    Henquin, E.R.; Bisang, J.M.

    2011-01-01

    Highlights: → For this reactor configuration the current distribution is uniform. → For this reactor configuration with bipolar connection the leakage current is small. → The mass-transfer conditions are closely uniform along the electrode. → The fluidodynamic behaviour can be represented by the dispersion model. → This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of ±10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  10. Reactor aging research. United States Nuclear Regulatory Commission

    International Nuclear Information System (INIS)

    Vassilaros, M.G.

    1998-01-01

    The reactor ageing research activities in USA described, are focused on the research of reactor vessel integrity, including regulatory issues and technical aspects. Current emphasis are described for fracture analysis, embrittlement research, inspection capabilities, validation od annealing rule, revision of regulatory guide

  11. Study on secondary shutdown systems in Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jalali, H.R.; Fadaei, A.H., E-mail: Fadaei_amir@aut.ac.ir; Gharib, M.

    2015-09-15

    Highlights: • A study was undertaken to summarize the techniques for secondary shutdown systems (SSS). • Neutronic calculation performed for proposed systems as SSS. • Dumping the heavy water stored in the reflector vessel is capable to shut down reactor. • Neutronic and transient calculation was done for validating the selected SSS. • All calculation shown that this system has advantages in safety and neutron economy. - Abstract: One important safety aspect of any research reactor is the ability to shut down the reactor. Usually, research reactors, currently in operation, have a single shutdown system based on the simultaneous insertion of the all control rods into the reactor core through gravity. Nevertheless, the International Atomic Energy Agency currently recommends use of two shutdown systems which are fully independent from each other to guarantee secure shutdown when one of them fails. This work presents an investigative study into secondary shutdown systems, which will be an important safety component in the research reactor and will provide another alternative way to shut down the reactor emergently. As part of this project, a study was undertaken to summarize the techniques that are currently used at world-wide research reactors for recognizing available techniques to consider in research reactors. Removal of the reflector, removal of the fuels, change in critical shape of reactor core and insertion of neutron absorber between the core and reflector are selected as possible techniques in mentioned function. In the next step, a comparison is performed for these methods from neutronic aspects. Then, chosen method is studied from the transient behavior point of view. Tehran research reactor which is a 5 MW open-pool reactor selected as a case study and all calculations are carried out for it. It has 5 control rods which serve the purpose of both reactivity control and shutdown of reactor under abnormal condition. Results indicated that heavy

  12. Supply of enriched uranium for research reactors

    International Nuclear Information System (INIS)

    Mueller, Hans; Laucht, Juergen

    1996-01-01

    Since the RERTR meeting in 1990 at Newport/USA, NUKEM recommended that the research reactor community agree upon a worldwide unified technical specification for low enriched uranium (LEU) and high enriched uranium (HEU) since there existed numerous specifications both from suppliers/fabricators and research reactors. The target recommended by NUKEM is to arrive at a worldwide unified standard specification in order to facilitate supplies of LEU and HEU to fabricators for fabrication of research reactor fuel elements. In our paper presented at the RERTR meeting at Paris in September 1995, we pointed out that LEU and HEU supplied by the U.S. Department of Energy (DOE) in the past was never 'virgin' material, i.e., it was mixed with reprocessed uranium. Our recommendation was to include this fact in the proposed unified specification. Since the RERTR meeting in 1995 progress on a unified standard specification has been made and we would like to provide more specific information about that in this paper. Furthermore, we will deal with the question whether there is a secure supply of LEU for converted research reactors. We list current and potential suppliers of LEU, noting however, that the DOE has for a number of years been unable to supply any LEU due to production problems. The future availability of LEU of U.S. origin is, of course, essential for those research reactor operators which have converted their reactors from HEU to LEU and which are intending to return spent fuel of U.S. origin to the U.S.A. (author)

  13. Australia's replacement research reactor project

    International Nuclear Information System (INIS)

    Harris, K.J.

    1999-01-01

    HIFAR, a 10 MW tank type DIDO Class reactor has operated at the Lucas Heights Science and Technology Centre for 43 years. HIFAR and the 10 kW Argonaut reactor 'Moata' which is in the Care and Maintenance phase of decommissioning are Australia's only nuclear reactors. The initial purpose for HIFAR was for materials testing to support a nuclear power program. Changing community attitude through the 1970's and a Government decision not to proceed with a planned nuclear power reactor resulted in a reduction of materials testing activities and a greater emphasis being placed on neutron beam research and the production of radioisotopes, particularly for medical purposes. HIFAR is not fully capable of satisfying the expected increase in demand for medical radiopharmaceuticals beyond the next 5 years and the radial configuration of the beam tubes severely restricts the scope and efficiency of neutron beam research. In 1997 the Australian Government decided that a replacement research reactor should be built by the Australian Nuclear Science and Technology Organisation at Lucas Heights subject to favourable results of an Environmental Impact Study. The Ei identified no reasons on the grounds of safety, health, hazard or risk to prevent construction on the preferred site and it was decided in May 1999 that there were no environmental reasons why construction of the facility should not proceed. In recent years ANSTO has been reviewing the operation of HIFAR and observing international developments in reactor technology. Limitations in the flexibility and efficiency achievable in operation of a tank type reactor and the higher intrinsic safety sought in fundamental design resulted in an early decision that the replacement reactor must be a pool type having cleaner and higher intensity tangential neutron beams of wider energy range than those available from HIFAR. ANSTO has chosen to use it's own resources supported by specialised external knowledge and experience to identify

  14. Current tendencies and perspectives of development research reactors of Russia

    International Nuclear Information System (INIS)

    Gabaraev, B.A.; Kchmelschikov, V.V.

    2004-01-01

    Full text: During more than fifty years many Research Reactors were constructed under Russian projects, and that is a considerable contribution to the world reactor building. The designs of Research Reactors, constructed under Russian projects, appeared to be so successful, that permitted to raise capacity and widen the range of their application. The majority of Russian Research Reactors being middle-aged are far from having their designed resources exhausted and are kept on the intensive run still. In 2000 'Strategy of nuclear power development in Russia in the first half of XXI century' was elaborated and approved. The national nuclear power requirements and possible ways of its development determined in this document demanded to analyze the state of the research reactors base. The analysis results are presented in this report. The main conclusion consists in the following statement: on the one hand quantity and experimental potentialities of domestic Research Reactors are sufficient for the solution of reactor materials science tasks, and on the other hand the reconstruction and modernization appears to be the most preferable way of research reactors development for the near-term outlook. At present time the modernization and reconstruction works and works on extension of operational life of high-powered multipurpose MIR-M1, SM-3, IRV-1M, BOR-60, IVV-2M and others are conducted. There is support for the development of Research Reactors, intended for carrying out the fundamental investigations on the neutron beams. Toward this end the Government of Russia gives financial and professional support with a view to complete the reactor PIK construction in PINPh and the reactor IBR-2 modernization in JINR. In future prospect Research Reactors branch in Russia is to acquire the following trends: - limited number of existent scientific centers, based on the construction sites, with high flux materials testing research reactors, equipped with experimental facilities

  15. Decommissioning of the Salaspils Research Reactor

    Directory of Open Access Journals (Sweden)

    Abramenkovs Andris

    2011-01-01

    Full Text Available In May 1995, the Latvian government decided to shut down the Salaspils Research Reactor and to dispense with nuclear energy in the future. The reactor has been out of operation since July 1998. A conceptual study on the decommissioning of the Salaspils Research Reactor was drawn up by Noell-KRC-Energie- und Umwelttechnik GmbH in 1998-1999. On October 26th, 1999, the Latvian government decided to start the direct dismantling to “green-field” in 2001. The upgrading of the decommissioning and dismantling plan was carried out from 2003-2004, resulting in a change of the primary goal of decommissioning. Collecting and conditioning of “historical” radioactive wastes from different storages outside and inside the reactor hall became the primary goal. All radioactive materials (more than 96 tons were conditioned for disposal in concrete containers at the radioactive wastes depository “Radons” at the Baldone site. Protective and radiation measurement equipment of the personnel was upgraded significantly. All non-radioactive equipment and materials outside the reactor buildings were released for clearance and dismantled for reuse or conventional disposal. Contaminated materials from the reactor hall were collected and removed for clearance measurements on a weekly basis.

  16. Status report of Indonesian research reactors

    International Nuclear Information System (INIS)

    Arbie, B.; Supadi, S.

    1995-01-01

    A general description of the three Indonesia research reactors, their irradiation facilities and future prospect are given. The 250 kW Triga Mark II in Bandung has been in operation since 1965 and in 1972 its designed power was increased to 1000 kW. The core grid from the previous 250 kW Triga Mark II was then used by Batan for designing and constructing the Kartini reactor in Yogyakarta. This reactor commenced its operation in 1979. Both Triga reactors have served a wide spectrum of utilization such as for manpower training in nuclear engineering, radiochemistry, isotope production, and beam research in solid state physics. The Triga reactor management in Bandung has a strong cooperation with the Bandung Institute of Technology and the one in Yogyakarta with the Gadjah Mada University which has a Nuclear Engineering Department at its Faculty of Engineering. In 1976 there emerged an idea to have a high flux reactor appropriate for Indonesia's intention to prepare an infrastructure for both nuclear energy and non-energy industry era. Such an idea was then realized with the achievement of the first criticality of the RSG-GAS reactor at the Serpong area. It is now expected that by early 1992 the reactor will reach its full 30 MW power level and by the end of 1992 the irradiation facilities be utilizable fully for future scientific and engineering work. As a part of the national LEU fuel development program a study has been underway since early 1989 to convert the RSG-GAS reactor core from using oxide fuel to using higher loading silicide fuel. (author)

  17. Plans for use of ENDF/B in reactor research in Indonesia

    International Nuclear Information System (INIS)

    Santoso, B.; Syaukat, A.; Subki, I.; Ganesan, S.

    1989-07-01

    Nuclear data are numerical constants of nature which quantify the nuclear behaviour of all elements and isotopes which make up the reactor medium and its environment, and which are needed as input for performing design calculations for safe and reliable operation of nuclear reactors. The nuclear data are available in the form of recommended values in specially formatted computerized files such as the Evaluated Nuclear Data File-B, known as ENDF/B. The development of base technology in the scheme of original reactor design calculations involves the mastering of the art of ENDF/B data processing. This paper briefly discusses the current status of this activity in Jakarta and gives an account of the future plans, with emphasis on the role of ENDF/B in reactor calculations. (author). 15 refs, 9 figs

  18. Strategy for Sustainable Utilization of IRT-Sofia Research Reactor

    International Nuclear Information System (INIS)

    Mitev, M.; Apostolov, T.; Ilieva, K.; Belousov, S.; Nonova, T.

    2013-01-01

    The Research Reactor IRT-2000 in Sofia is in process of reconstruction into a low-power reactor of 200 kW under the decision of the Council of Ministers of Republic of Bulgaria from 2001. The reactor will be utilized for development and preservation of nuclear science, skills, and knowledge; implementation of applied methods and research; education of students and training of graduated physicists and engineers in the field of nuclear science and nuclear energy; development of radiation therapy facility. Nuclear energy has a strategic place within the structure of the country’s energy system. In that aspect, the research reactor as a material base, and its scientific and technical personnel, represent a solid basis for the development of nuclear energy in our country. The acquired scientific experience and qualification in reactor operation are a precondition for the equal in rights participation of the country in the international cooperation and the approaching to the European structures, and assurance of the national interests. Therefore, the operation and use of the research reactor brings significant economic benefits for the country. For education of students in nuclear energy, reactor physics experiments for measurements of static and kinetic reactor parameters will be carried out on the research reactor. The research reactor as a national base will support training and applied research, keep up the good practice and the preparation of specialists who are able to monitor radioactivity sources, to develop new methods for detection of low quantities of radioactive isotopes which are hard to find, for deactivation and personal protection. The reactor will be used for production of isotopes needed for medical therapy and diagnostics; it will be the neutron source in element activation analysis having a number of applications in industrial production, medicine, chemistry, criminology, etc. The reactor operation will increase the public understanding, confidence

  19. Research reactor spent fuel management in Argentina

    International Nuclear Information System (INIS)

    Audero, M.A.; Bevilacqua, A.M.; Mehlich, A.M.; Novara, O.

    2002-01-01

    The research reactor spent fuel (RRSF) management strategy will be presented as well as the interim storage experience. Currently, low-enriched uranium RRSF is in wet interim storage either at reactor site or away from reactor site in a centralized storage facility. High-enriched uranium RRSF from the centralized storage facility has been sent to the USA in the framework of the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The strategy for the management of the RRSF could implement the encapsulation for interim dry storage. As an alternative to encapsulation for dry storage some conditioning processes are being studied which include decladding, isotopic dilution, oxidation and immobilization. The immobilized material will be suitable for final disposal. (author)

  20. Sustainability management for operating organizations of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo; Aquino, Afonso Rodrigues de, E-mail: ekibrit@ipen.br, E-mail: araquino@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    Sustainable development is development that meets the needs of the present without compromising the ability of future generations to meet their own needs. In a country like Brazil, where nuclear activity is geared towards peaceful purposes, any operating organization of research reactor should emphasize its commitment to social, environmental, economic and institutional aspects. Social aspects include research and development, production and supply of radiopharmaceuticals, radiation safety and special training for the nuclear sector. Environmental aspects include control of the surroundings and knowledge directed towards environment preservation. Economic aspects include import substitution and diversification of production. Institutional aspects include technology, innovation and knowledge. These aspects, if considered in the management system of an operating organization of research reactor, will help with its long-term maintenance and success in an increasingly competitive market scenario. About this, we propose a sustainability management system approach for operating organizations of research reactors. A bibliographical review on the theme is made. A methodology for identifying indicators for measuring sustainability in nuclear research reactors processes is also described. Finally, we propose a methodology for sustainability perception assessment to be applied at operating organizations of research reactors. (author)

  1. Sustainability management for operating organizations of research reactors

    International Nuclear Information System (INIS)

    Kibrit, Eduardo; Aquino, Afonso Rodrigues de

    2017-01-01

    Sustainable development is development that meets the needs of the present without compromising the ability of future generations to meet their own needs. In a country like Brazil, where nuclear activity is geared towards peaceful purposes, any operating organization of research reactor should emphasize its commitment to social, environmental, economic and institutional aspects. Social aspects include research and development, production and supply of radiopharmaceuticals, radiation safety and special training for the nuclear sector. Environmental aspects include control of the surroundings and knowledge directed towards environment preservation. Economic aspects include import substitution and diversification of production. Institutional aspects include technology, innovation and knowledge. These aspects, if considered in the management system of an operating organization of research reactor, will help with its long-term maintenance and success in an increasingly competitive market scenario. About this, we propose a sustainability management system approach for operating organizations of research reactors. A bibliographical review on the theme is made. A methodology for identifying indicators for measuring sustainability in nuclear research reactors processes is also described. Finally, we propose a methodology for sustainability perception assessment to be applied at operating organizations of research reactors. (author)

  2. IAEA activities in the field of research reactors safety

    International Nuclear Information System (INIS)

    Ciuculescu, C.; Boado Magan, H.J.

    2004-01-01

    IAEA activities in the field of research reactor safety are included in the programme of the Division of Nuclear Installations Safety. Following the objectives of the Division, the results of the IAEA missions and the recommendations from International Advisory Groups, the IAEA has conducted in recent years a certain number of activities aiming to enhance the safety of research reactors. The following activities will be presented: (a) the new Requirements for the Safety of Research Reactors, main features and differences with previous standards (SS-35-S1 and SS-35-S2) and the grading approach for implementation; (b) new documents being developed (safety guides, safety reports and TECDOC's); (c) activities related to the Incident Reporting System for Research Reactor (IRSRR); (d) the new features implemented for the INSARR missions; (e) the Code of Conduct on the Safety of Research Reactors adopted by the Board of Governors on 8 March 2004, following the General Conference Resolution GC(45)/RES/10; and (f) the survey on the safety of research reactors published on the IAEA website on February 2003 and the results obtained. (author)

  3. Dynamic simulation platform to verify the performance of the reactor regulating system for a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-07-01

    Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of research reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)

  4. Research for enhancing reactor safety

    International Nuclear Information System (INIS)

    1989-05-01

    Recent research for enhanced reactor safety covers extensive and numerous experiments and computed modelling activities designed to verify and to improve existing design requirements. The lectures presented at the meeting report GRS research results and the current status of reactor safety research in France. The GRS experts present results concerning expert systems and their perspectives in safety engineering, large-scale experiments and their significance in the development and verification of computer codes for thermohydraulic modelling of safety-related incidents, the advanced system code ATHLET for analysis of thermohydraulic processes of incidents, the analysis simulator which is a tool for fast evaluation of accident management measures, and investigations into event sequences and the required preventive emergency measures within the German Risk Study. (DG) [de

  5. Training of research reactor personnel

    International Nuclear Information System (INIS)

    Cherruau, F.

    1980-01-01

    Research reactor personnel operate the reactor and carry out the experiments. These two types of work entail different activities, and therefore different skills and competence, the number of relevant staff being basically a function of the size, complexity and versatility of the reactor. Training problems are often reactor-specific, but the present paper considers them from three different viewpoints: the training or retraining of new staff or of personnel already employed at an existing facility, and training of personnel responsible for the start-up and operation of a new reactor, according to whether local infrastructure and experience already exist or whether they have to be built up from scratch. On-the-spot experience seems to be an essential basis for sound training, but requires teaching abilities and aids often difficult to bring together, and the availability of instructors that does not always fit in smoothly with current operational and experimental tasks. (author)

  6. On the utilization of neutron beams of research reactors in research and applications

    International Nuclear Information System (INIS)

    FAYEK, M.K.

    2000-01-01

    Nuclear research reactors are the most widely available neutron sources, and they are capable of producing very high fluxes of neutrons having a considerable range of energies, from a few MeV to 10 MeV. Therefore, these neutrons can be used in many fields of basic research and for applications in physics, chemistry, medicine, biology, etc. Experiments with research reactors over the last 50 years have laid the foundations of today's nuclear technology. In addition, research reactors continue to be utilized as facilities for testing materials and in training manpower for nuclear programs, because basic training on a research reactor provides an essential understanding of the nuclear process, and personnel become accustomed to work under the special conditions resulting from irradiation and contamination risks

  7. A model for nuclear research reactor dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Barati, Ramin, E-mail: Barati.ramin@aut.ac.ir; Setayeshi, Saeed, E-mail: setayesh@aut.ac.ir

    2013-09-15

    Highlights: • A thirty-fourth order model is used to simulate the dynamics of a research reactor. • We consider delayed neutrons fraction as a function of time. • Variable fuel and temperature reactivity coefficients are used. • WIMS, BORGES and CITVAP codes are used for initial condition calculations. • Results are in agreement with experimental data rather than common codes. -- Abstract: In this paper, a useful thirty-fourth order model is presented to simulate the kinetics and dynamics of a research reactor core. The model considers relevant physical phenomena that govern the core such as reactor kinetics, reactivity feedbacks due to coolant and fuel temperatures (Doppler effects) with variable reactivity coefficients, xenon, samarium, boron concentration, fuel burn up and thermal hydraulics. WIMS and CITVAP codes are used to extract neutron cross sections and calculate the initial neuron flux respectively. The purpose is to present a model with results similar to reality as much as possible with reducing common simplifications in reactor modeling to be used in different analyses such as reactor control, functional reliability and safety. The model predictions are qualified by comparing with experimental data, detailed simulations of reactivity insertion transients, and steady state for Tehran research reactor reported in the literature and satisfactory results have been obtained.

  8. Procedures for the medical application of research reactors (Appendix)

    International Nuclear Information System (INIS)

    Nishihara, H.; Kanda, K.

    2004-01-01

    The Kyoto University Reactor (KUR) is one of the four research reactors in Japan that are currently licensed for medical application, in addition to other research purposes. Taking the KUR as an example, legal and other procedures for using research reactors for boron neutron capture therapy (BNCT) are described, which are practiced in accordance with the 'Provisional Guideline Pertaining to Medical Irradiation by Accelerators and/or Reactors, other than defined by the Medical Service Act' of the Science Council of Japan

  9. Experts' discussion on reactor safety research

    International Nuclear Information System (INIS)

    1980-01-01

    The experts' discussion on reactor safety research deals with risk analysis, political realization, man and technics, as well as with the international state of affairs. Inspite of a controversy on individual issues and on the proportion of governmental and industrial involvment in reactor safety research, the continuation and intensification of corresponding research work is said to be necessary. Several participants demanded to consider possible 'conventional accidents' as well as a stronger financial commitment by the industry in this sector. The ratio 'man and technics' being an interface decisive for the proper functioning or failure of complex technical systems requires even more research work to be done. (GL) [de

  10. Nuclear reactor kinetics and control

    International Nuclear Information System (INIS)

    Lewins, J.

    1978-01-01

    A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

  11. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    Markoczy, G.; Hudina, M.; Richmond, R.; Wydler, P.; Stratton, R.W.; Burgsmueller, P.

    1980-03-01

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  12. Manual on reliability data collection for research reactor PSAs

    Energy Technology Data Exchange (ETDEWEB)

    1992-01-01

    The IAEA has been actively promoting performance of probabilistic safety assessment (PSA) studies for research reactors. From 1986 to 1988 the IAEA undertook a Coordinated Research Programme (CRP) on PSA for research reactors which helped promote use of PSA and foster a broad exchange of information. Although the basic methodological approach in performing a research reactor PSA is understood, some unresolved issues, data availability being among them, still exist. To address the issue on the international level, the IAEA initiated a new CRP on ``Data Acquisition for Research Reactors PSA Studies``. The aim of the CRP is to develop a data collection system and generate research reactor specific reliability data for use in PSAs. The achieve this aim a set of precise definitions should be adopted. A set of definitions developed specifically for research reactors and covering classification of equipment and failure terms, reliability parameters, failure modes and other terms necessary for data collection and processing is presented in this document which is based on discussions during the first meeting of the CRP held in Vienna in October 1989 and during the second meeting held in Beijing, China, in October 1990. Refs and figs.

  13. Manual on reliability data collection for research reactor PSAs

    International Nuclear Information System (INIS)

    1992-01-01

    The IAEA has been actively promoting performance of probabilistic safety assessment (PSA) studies for research reactors. From 1986 to 1988 the IAEA undertook a Coordinated Research Programme (CRP) on PSA for research reactors which helped promote use of PSA and foster a broad exchange of information. Although the basic methodological approach in performing a research reactor PSA is understood, some unresolved issues, data availability being among them, still exist. To address the issue on the international level, the IAEA initiated a new CRP on ''Data Acquisition for Research Reactors PSA Studies''. The aim of the CRP is to develop a data collection system and generate research reactor specific reliability data for use in PSAs. The achieve this aim a set of precise definitions should be adopted. A set of definitions developed specifically for research reactors and covering classification of equipment and failure terms, reliability parameters, failure modes and other terms necessary for data collection and processing is presented in this document which is based on discussions during the first meeting of the CRP held in Vienna in October 1989 and during the second meeting held in Beijing, China, in October 1990. Refs and figs

  14. Refurbishment and safety upgradation of research reactor Cirus

    International Nuclear Information System (INIS)

    Marik, S.K.; Rao, D.V.H.; Bhatnagar, A.; Pant, R.C.; Tikku, A.C.; Sankar, S.

    2006-01-01

    Cirus, a 40 MW t, vertical tank type research reactor, having wide range of research facilities, was commissioned in the year 1960. This research reactor, situated at Mumbai, India has been operated and utilized extensively for isotope production, material testing and neutron beam research for nearly four decades. With a view to assess the residual life of the reactor, detailed ageing studies were carried out during the early 1990s. Based on these studies, refurbishment of Cirus for its life extension was taken up. During refurbishment, additional safety features were incorporated in various systems to qualify them for the current safety standards. This paper gives the details of the operating experiences, utilization of the reactor along with methodologies followed for carrying out detailed ageing studies, refurbishment and safety upgradation for its life extension

  15. Future plans for the Imperial College CONSORT research reactor

    International Nuclear Information System (INIS)

    Franklin, S.J.

    1999-01-01

    The Imperial College (IC) research reactor was designed jointly by GEC and the IC Mechanical Engineering Department. It first went critical on 9 April 1965 and has been operating successfully for over 33 years. The reactor provides a service to both academia and industry for neutron activation analysis, reactor and applied nuclear physics training, neutron detector calibration, isotope production and irradiations. The reactor has strategic importance for the UK, as it is now the only remaining research reactor in the country. It is therefore important to put in place refurbishment programmes and to maintain and upgrade the safety case. This paper describes the current facilities, applications and users of the research reactor and outlines both the recent and the planned developments. (author)

  16. Research reactor spent fuel in Ukraine

    International Nuclear Information System (INIS)

    Trofimenko, A.P.

    1996-01-01

    This paper describes the research reactors in Ukraine, their spent fuel facilities and spent fuel management problems. Nuclear sciences, technology and industry are highly developed in Ukraine. There are 5 NPPs in the country with 14 operating reactors which have total power capacity of 12,800 MW

  17. 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)

  18. Development of Safety Review Guidance for Research and Training Reactors

    International Nuclear Information System (INIS)

    Oh, Kju-Myeng; Shin, Dae-Soo; Ahn, Sang-Kyu; Lee, Hoon-Joo

    2007-01-01

    The KINS already issued the safety review guidance for pressurized LWRs. But the safety review guidance for research and training reactors were not developed. So, the technical standard including safety review guidance for domestic research and training reactors has been applied mutates mutandis to those of nuclear power plants. It is often difficult for the staff to effectively perform the safety review of applications for the permit by the licensee, based on peculiar safety review guidance. The NRC and NSC provide the safety review guidance for test and research reactors and European countries refer to IAEA safety requirements and guides. The safety review guide (SRG) of research and training reactors was developed considering descriptions of the NUREG- 1537 Part 2, previous experiences of safety review and domestic regulations for related facilities. This study provided the safety review guidance for research and training reactors and surveyed the difference of major acceptance criteria or characteristics between the SRG of pressurized light water reactor and research and training reactors

  19. Materials research with neutron beams from a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Root, J.; Banks, D. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, Ontario (Canada)

    2015-03-15

    Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

  20. A strategy for resolving the research reactor dilemma in the US

    International Nuclear Information System (INIS)

    Kerr, H.T.

    1991-01-01

    The steadily declining number of operating research reactors in the US has been characterized as a growing dilemma that could significantly limit future opportunities for research and educational programs. An overview is presented describing the existing inventory of research reactors in the US Projections are given of potential research and other uses for the reactors. The factors which have contributed to the declining population of research reactors are discussed, and a strategy is proposed to identify and preserve those research reactor facilities needed to fulfill future national needs. The proposed strategy will focus on establishment of user-oriented research reactor centers that are affiliated with reactors at universities, national laboratories, and defense sites in the US and, where appropriate, in foreign countries