WorldWideScience

Sample records for reactor internal irradiation

  1. Mathematical models in Slowpoke reactor internal irradiation site

    International Nuclear Information System (INIS)

    Raza, J.

    2007-01-01

    The main objective is to build representative mathematical models of neutron activation analysis in a Slowpoke internal irradiation site. Another significant objective is to correct various elements neutron activation analysis measured mass using these models. The neutron flux perturbation is responsible for the measured under-estimation of real masses. We supposed that neutron flux perturbation measurements taken during the Ecole Polytechnique de Montreal Slowpoke reactor first fuel loading were still valid after the second fuelling. .We also supposed that the thermal neutrons spatial and kinetic energies distributions as well as the absorption microscopic cross section dependence on the neutrons kinetic energies were important factors to satisfactorily represent neutron activation analysis results. In addition, we assumed that the neutron flux is isotropic in the laboratory system. We used experimental results from the Slowpoke reactor internal irradiation sites, in order to validate our mathematical models. Our models results are in close agreement with these experimental results..We established an accurate global mathematical correlation of the neutron flux perturbation in function of samples volumes and macroscopic neutron absorption cross sections. It is applicable to sample volumes ranging from 0,1 to 1,3 ml and macroscopic neutron absorption cross section up to 5 moles-b for seven (7) elements with atomic numbers (Z) ranging from 5 to 79. We first came up with a heuristic neutron transport mathematical semi-analytical model, in order to better understand neutrons behaviour in presence of one of several different nuclei samples volumes and mass. In order to well represent the neutron flux perturbation, we combined a neutron transport solution obtained from the spherical harmonics method of a finite cylinder and a mathematical expression combining two cylindrical harmonic functions..With the help of this model and the least squares method, we made extensive

  2. Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

    2017-06-02

    This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

  3. Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

    International Nuclear Information System (INIS)

    Was, Gary

    2017-01-01

    This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

  4. Influence of geometrical parameters of the VVER-1000 reactor construction elements to internals irradiation conditions

    Directory of Open Access Journals (Sweden)

    О. M. Pugach

    2015-07-01

    Full Text Available Investigations to determine the influences of geometrical parameters of the calculational VVER-1000 reactor model to the results of internal irradiation condition determination are carried out. It is shown that the values of appropriate sensitivity matrix elements are not dependent on a height coordinate for any core level, but there is their azimuthal dependence. Maximum possible relative biases of neutron fluence due to inexact knowledge of internal geometrical parameters are obtained for the baffle and the barrel.

  5. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    Radiation effects on metals or alloys used in fast reactor core components are examined in the papers presented at this conference, the accent being put on swelling and irradiation creep of steels and nickel alloys

  6. Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors

    International Nuclear Information System (INIS)

    Pokor, C.

    2003-01-01

    The core internals of Pressurized Water Reactors (PWR) are composed of SA 304 stainless steel plates and CW 316 stainless steel bolts. These internals undergo a neutron flux at a temperature between 280 deg C and 380 deg C which modifies their mechanical properties. These modifications are due to the changes in the microstructure of these materials under irradiation which depend on flux, dose and irradiation temperature. We have studied, by Transmission Electron Microscopy, the microstructure of stainless steels SA 304, CW 316 and CW 316Ti irradiated in a mixed flux reactor (OSIRIS at 330 deg C between 0,8 dpa et 3,4 dpa) and in a fast breeder reactor at 330 deg C (BOR-60) up to doses of 40 dpa. Moreover, samples have been irradiated at 375 deg C in a fast breeder reactor (EBR-II) up to doses of 10 dpa. The microstructure of the irradiated stainless steels consists in faulted Frank dislocation loops in the [111] planes of austenitic, with a Burgers vector of [111]. It is possible to find some voids in the solution annealed samples irradiated at 375 deg C. The evolution of the dislocations loops and voids has been simulated with a 'cluster dynamic' model. The fit of the model parameters has allowed us to have a quantitative description of our experimental results. This description of the microstructure after irradiation was coupled together with a hardening model by Frank loops that has permitted us to make a quantitative description of the hardening of SA 304, CW 316 and CW 316Ti stainless steels after irradiation at a certain dose, flux and temperature. The irradiation doses studied grow up to 90 dpa, dose of the end of life of PWR internals. (author)

  7. The future Jules Horowitz material testing reactor: An opportunity for developing international collaborations on a major European irradiation infrastructure

    International Nuclear Information System (INIS)

    Parrat, D.; Bignan, G.; Maugard, B.; Gonnier, C.; Blandin, C.

    2015-01-01

    Development process of a fuel product or a nuclear material before using at an industrial scale in a power reactor ranges from characterization of the material itself under neutronic flux up to its qualification in accidental conditions. Irradiations in Material Testing Reactors (MTRs) are in practice the basis of the whole process, in complement of prediction capabilities gained by modelling. Dedicated experimental reactors play also an important complementary role for some specific integral tests (e.g. RIA tests). Irradiations of precursors in power reactors are often limited to products which present a slight design evolution compare to the standard product or are implemented for further tests when a statistical approach is useful for defining a safety criterion. However European MTR park status is characterized by ageing infrastructures, which could cause operational issues in coming years, either on technological or on safety point of views. Moreover some specific supplies related to the public demand could be strongly affected (e.g. radiopharmaceutical targets). To avoid a lack in irradiation capacity offer at European level, CEA launched the Jules Horowitz Material Testing Reactor (JHR) international program, in the frame of a Consortium gathering also EDF (FR), AREVA (FR), European Commission (EU), SCK.CEN (BE), VTT (FI), CIEMAT (SP), STUDSVIK (SE), UJV (CZ), NNL (UK), IAEC (IL), DAE (IN) and as associated partnership: JAEA (JP). Some institutions in this list are themselves the flagship of a national Consortium. Discussions for enlarging participation are on-going with other countries, as JHR Consortium is open to new member entrance until JHR completion. The Jules Horowitz Material Testing Reactor (JHR MTR) is under construction at CEA Cadarache in southern France and will be an important international User Facility for R&D in support to the nuclear industry, research centres, regulatory bodies and TSO, and academic institutions. It represents a unique

  8. Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors; Caracterisation microstructurale et modelisation du durcissement des aciers austenitiques irradies des structures internes des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Pokor, C

    2003-07-01

    The core internals of Pressurized Water Reactors (PWR) are composed of SA 304 stainless steel plates and CW 316 stainless steel bolts. These internals undergo a neutron flux at a temperature between 280 deg C and 380 deg C which modifies their mechanical properties. These modifications are due to the changes in the microstructure of these materials under irradiation which depend on flux, dose and irradiation temperature. We have studied, by Transmission Electron Microscopy, the microstructure of stainless steels SA 304, CW 316 and CW 316Ti irradiated in a mixed flux reactor (OSIRIS at 330 deg C between 0,8 dpa et 3,4 dpa) and in a fast breeder reactor at 330 deg C (BOR-60) up to doses of 40 dpa. Moreover, samples have been irradiated at 375 deg C in a fast breeder reactor (EBR-II) up to doses of 10 dpa. The microstructure of the irradiated stainless steels consists in faulted Frank dislocation loops in the [111] planes of austenitic, with a Burgers vector of [111]. It is possible to find some voids in the solution annealed samples irradiated at 375 deg C. The evolution of the dislocations loops and voids has been simulated with a 'cluster dynamic' model. The fit of the model parameters has allowed us to have a quantitative description of our experimental results. This description of the microstructure after irradiation was coupled together with a hardening model by Frank loops that has permitted us to make a quantitative description of the hardening of SA 304, CW 316 and CW 316Ti stainless steels after irradiation at a certain dose, flux and temperature. The irradiation doses studied grow up to 90 dpa, dose of the end of life of PWR internals. (author)

  9. Thermophysical instruments for non-destructive examination of tightness and internal gas pressure or irradiated power reactor fuel rods

    International Nuclear Information System (INIS)

    Pastoushin, V.V.; Novikov, A.Yu.; Bibilashvili, Yu.K.

    1998-01-01

    The developed thermophysical method and technical instruments for non-destructive leak-tightness and gas pressure inspection inside irradiated power reactor fuel rods and FAs under poolside and hot cell conditions are described. The method of gas pressure measuring based on the examination of parameters of thermal convection that aroused in gas volume of rod plenum by special technical instruments. The developed method and technique allows accurate value determination of not only one of the main critical rod parameters, namely total internal gas pressure, that forms rod mean life in the reactor core, but also the partial pressure of every main constituent of gaseous mixture inside irradiated fuel rod, that provides the feasibility of authentic and reliable leak-tightness detection. The described techniques were experimentally checked during the examination of all types power reactor fuel rods existing in Russia (WWER, BN, RBMK) and could form the basis for new technique development for non-destructive examination of PWR (and other) type rods and FAs having gas plenum filled with spring or another elements of design. (author)

  10. The future Jules Horowitz material test reactor: A major European research infrastructure for sustaining the international irradiation capacity

    International Nuclear Information System (INIS)

    Parrat, D.; Bignan, G.; Chauvin, J.; Gonnier, C.

    2011-01-01

    Multipurpose experimental reactors are now key infrastructures, in complement of prediction capabilities gained thanks to progresses in the modelling, for supporting nuclear energy in terms of safety, ageing management, innovation capacity, economical performances and training. However the European situation in this field is characterized by ageing large infrastructures, which could face to operational issues in the coming years and could jeopardize the knowledge acquisition and the nuclear product qualification. Moreover some specific supplies related to the public demand could be strongly affected (e.g. radiopharmaceutical targets). To avoid a lack in the experimental capacity offer at the European level, the CEA has launched the Jules Horowitz material test reactor (JHR) international program, in the frame of a Consortium gathering EDF (FR), AREVA (FR), the European Commission (EU), SCK.CEN (BE), VTT (FI), CIEMAT (SP), VATTENFALL (SE), UJV (CZ), JAEA (JP) and the DAE (IN). The JHR will be a 100 MW tank pool reactor and will have several experimental locations either inside the reactor core or outside the reactor tank in a reflector constituted by beryllium blocks. Excavation works started mid-2007 on the CEA Cadarache site in the southeast of France. After the construction permit delivery gained in September 2007, building construction began at the beginning of 2009. Reactor start-up is scheduled in 2016. The JHR is designed to offer up-to-date irradiation experimental capabilities for studying nuclear material and fuel behaviour under irradiation in a modern safety frame, mainly due to: 1) High values of fast and thermal neutron fluxes in the core and high thermal neutron flux in the reflector (producing typically twice more material damages per year than available today in European MTRs); 2) A large variety of experimental devices capable to reproduce environment conditions of mainly light water reactors (LWRs) and sodium fast reactors; 3) Several equipment

  11. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    In this conference are presented papers dealing with swelling of metals and alloys, (and specially ferritic steels), structural evolution and stability under irradiation, modifications of mechanical properties, consequences on the behaviour of fuel elements and the optimization of materials selection, and irradiation creep [fr

  12. 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.)

  13. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    International Nuclear Information System (INIS)

    MH Lane

    2006-01-01

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations

  14. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    Energy Technology Data Exchange (ETDEWEB)

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  15. Irradiated topaz in the reactor

    International Nuclear Information System (INIS)

    Helal, A.I.; Zahran, N.F.; Gomaac, M.A.M.; Salama, S.

    2007-01-01

    Gem stones are those stones which have beauty that can be based on its color, transparency, brilliance and the crystalline perfection . Topaz is used mainly as gemstones, It is the most common irradiated gem on the market. High energy such as neutrons, have enough energy to produce color centers . Irradiation is most often carried out in nuclear reactors (high-energy neutrons). Irradiation of topaz in the Egyptian research reactor (ETRR-2) by neutrons changes its cloudy white color to a reddish pink which could be changed to blue by heating

  16. International thermal reactor development

    International Nuclear Information System (INIS)

    Zebroski, E.L.

    1977-01-01

    The worldwide development of nuclear power plants is reviewed. Charts are presented which show the commitment to light-water reactor capacity construction with breakdown by region and country. Additional charts show the major nuclear research centers which have substantial scope in light water reactor development and extensive international activities

  17. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  18. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Blevins, J.D.; Stasko, R.R.

    1989-09-01

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  19. Intense neutron irradiation facility for fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

  20. International breeder reactor development

    International Nuclear Information System (INIS)

    Traube, K.

    1976-01-01

    For more than a decade, sodium cooled breeder reactors have now been in the focus of advanced nuclear power development in the major industrialized countries. In the sixties, a total of seven small experimental nuclear power stations were commissioned. Two of these have been shut down in the meantime, the others continue to work satisfactorily, their main purpose being the development of fuel elements. The years 1972-1974 saw the commissioning of the prototype power stations in the 300 MWe power category in France, the United Kingdom and the Soviet Union. Presently, other experimental reactors are under construction in the Federal Republic of Germany, Italy, Japan, the United States, plus another Soviet 600 MWe prototype reactor and the SNR 300 DeBeNeLux prototype at Kalkar. A comparison of the technological features either implemented or planned in the prototype and experimental power plants and of their fuel elements reveals a remarkable similarity in the basic concepts pursued in different countries. The two types of breeder reactors, viz. the loop and the pool types, show a closer resemblance to each other than do pressurized and boilling water reactors. The growing awareness of administrative problems emerging in the approaching phase of the introduction of large breeder power stations in a number of European countries has recently led to a streamlining effort in the structure of industries and to tentative steps towards international cooperation on a broad basis. (orig.) [de

  1. Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C

    Science.gov (United States)

    Yang, Yong; Chen, Yiren; Huang, Yina; Allen, Todd; Rao, Appajosula

    Reactor internal components are subjected to neutron irradiation in light water reactors, and with the aging of nuclear power plants around the world, irradiation-induced material degradations are of concern for reactor internals. Irradiation-induced defects resulting from displacement damage are critical for understanding degradation in structural materials. In the present work, microstructural changes due to irradiation in austenitic stainless steels and cast steels were characterized using transmission electron microscopy. The specimens were irradiated in the BOR-60 reactor, a fast breeder reactor, up to 40 dpa at 320°C. The dose rate was approximately 9.4x10-7 dpa/s. Void swelling and irradiation defects were analyzed for these specimens. A high density of faulted loops dominated the irradiated-altered microstructures. Along with previous TEM results, a dose dependence of the defect structure was established at 320°C.

  2. International document on food irradiation

    International Nuclear Information System (INIS)

    1990-06-01

    This international document highlights the major issues related to the acceptance of irradiated food by consumers, governmental and intergovernmental activities, the control of the process, and trade. The conference recognized that: Food irradiation has the potential to reduce the incidence of foodborne diseases. It can reduce post-harvest food losses and make available a larger quantity and a wider variety of foodstuffs for consumers. Regulatory control by competent authorities is a necessary prerequisite for introduction of the process. International trade in irradiated foods would be facilitated by harmonization of national procedures based on internationally recognized standards for the control of food irradiation. Acceptance of irradiated food by the consumer is a vital factor in the successful commercialization of the irradiation process, and information dissemination can contribute to this acceptance

  3. Role of irradiation reactor mock-ups

    International Nuclear Information System (INIS)

    Casali, F.; Cerles, J.M.; Debrue, J.

    1977-01-01

    A survey is given of the utilization of low power facilities in support to irradiation reactor experiments. The BRO2, ISIS and RB3 facilities are described as neutronic mock-ups of the BR2, OSIRIS and ESSOR reactors respectively

  4. Irradiation techniques at BR2 reactor

    International Nuclear Information System (INIS)

    Hebel, W.

    1978-01-01

    Since 1963 the material testing reactor BR2 at Mol is operated for the realisation of numerous research programs and experiments on the behavior of materials under nuclear radiation and in particular under intensive neutron exposure. During this period special irradiation techniques and experimental devices were developed according to the desiderata of the different experiments and to the irradiation possibilities offered at BR2. The design and the operating characteristics of quite a number of those irradiation rigs of proven reliability may be used or can be made available for new irradiation experiments. A brief description is given of some typical irradiation devices designed and constructed by CEN/SCK, Technology and Energy Dpt. They are compiled according to their main use for the different research and development programs realized at BR2. Their eventual application however for different objectives could be possible. A final chapter summarizes the principal irradiation conditions offered by BR2 reactor. (author)

  5. International Developments of Food Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Loaharanu, P. [Head, Food Preservation Section, Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture, Wagramerstr. 5, A-1400, Vienna (Austria)

    1997-12-31

    Food irradiation is increasingly accepted and applied in many countries in the past decade. Through its use, food losses and food-borne diseases can be reduced significantly, and wider trade in many food items can be facilitated. The past five decades have witnessed a positive evolution on food irradiation according to the following: 1940`s: discovery of principles of food irradiation; 1950`s: initiation of research in advanced countries; 1960`s: research and development were intensified in some advanced and developing countries; 1970`s: proof of wholesomeness of irradiated foods; 1980`s: establishment of national regulations; 1990`s: commercialization and international trade. (Author)

  6. International Developments of Food Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Loaharanu, P [Head, Food Preservation Section, Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture, Wagramerstr. 5, A-1400, Vienna (Austria)

    1998-12-31

    Food irradiation is increasingly accepted and applied in many countries in the past decade. Through its use, food losses and food-borne diseases can be reduced significantly, and wider trade in many food items can be facilitated. The past five decades have witnessed a positive evolution on food irradiation according to the following: 1940`s: discovery of principles of food irradiation; 1950`s: initiation of research in advanced countries; 1960`s: research and development were intensified in some advanced and developing countries; 1970`s: proof of wholesomeness of irradiated foods; 1980`s: establishment of national regulations; 1990`s: commercialization and international trade. (Author)

  7. International Developments of Food Irradiation

    International Nuclear Information System (INIS)

    Loaharanu, P.

    1997-01-01

    Food irradiation is increasingly accepted and applied in many countries in the past decade. Through its use, food losses and food-borne diseases can be reduced significantly, and wider trade in many food items can be facilitated. The past five decades have witnessed a positive evolution on food irradiation according to the following: 1940's: discovery of principles of food irradiation; 1950's: initiation of research in advanced countries; 1960's: research and development were intensified in some advanced and developing countries; 1970's: proof of wholesomeness of irradiated foods; 1980's: establishment of national regulations; 1990's: commercialization and international trade. (Author)

  8. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  9. Introduction to reactor internal materials for pressurized water reactor

    International Nuclear Information System (INIS)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

  10. The international thermonuclear reactor project

    International Nuclear Information System (INIS)

    James, T.R.

    1993-01-01

    The International Thermonuclear Experimental Reactor Project is a 6-year collaborative effort involving the U.S., Europe, Japan, and the Russian Federation to produce a detailed engineering design for the next-step fusion device

  11. Generation IV reactors: international projects

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

    2003-01-01

    Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

  12. Food irradiation receives international acceptance

    Energy Technology Data Exchange (ETDEWEB)

    Beddoes, J M [Atomic Energy of Canada Ltd., Ottawa, Ontario. Commercial Products

    1982-04-01

    Irradiation has advantages as a method of preserving food, especially in the Third World. The author tabulates some examples of actual use of food irradiation with dates and tonnages, and tells the story of the gradual acceptance of food irradiation by the World Health Organization, other international bodies, and the U.S. Food and Drug Administration (USFDA). At present, the joint IAEA/FAO/WHO standard permits an energy level of up to 5 MeV for gamma rays, well above the 1.3 MeV energy level of /sup 60/Co. The USFDA permits irradiation of any food up to 10 krad, and minor constituents of a diet may be irradiated up to 5 Mrad. The final hurdle to be cleared, that of economic acceptance, depends on convincing the food processing industry that the process is technically and economically efficient.

  13. International status of food irradiation

    International Nuclear Information System (INIS)

    Roberts, P.B.

    1982-09-01

    Recent international moves that are likely to result in an increasing acceptance of irradiated foods are reviewed. Particular attention is given to the activities of the FAO, WHO, Codex Alimentarius and to attitudes in the United States and the Asian-Pacific region. In 1979, the Codex Alimentarius Commission adopted a Recommended General Standard for Irradiated Food. A resume is given of a revised version of the standard that is presently under consideration. However, remaining barriers to trade in irradiated food are briefly discussed, such as legal and regulatory problems, labelling, public acceptance and economic viability

  14. International status of food irradiation

    International Nuclear Information System (INIS)

    Diehl, J.F.

    1983-01-01

    Radiation processing of foods has been studied for over 30 years. To a considerable extent this research was carried out in the framework of various international projects. After optimistic beginnings in the 1950s and long delays, caused by uncertainty about the health safety of foods so treated, food irradiation has now reached the stage of practical application in several countries. In order to prepare the way for world-wide accceptance of the new process, the Codex Alimentarius Commission has accepted an 'International General Standard for Irradiated Foods' and an 'International Code of Practice for the Operation of Irradiation Facilities Used for the Treatment of Foods'. Psychological barriers to a process associated with the word 'radiation' are still formidable; it appears, however, that acceptance by authorities, food industry and consumers continues to grow

  15. [International Thermonuclear Experimental Reactor support

    International Nuclear Information System (INIS)

    Dean, S.O.

    1990-01-01

    This report summarizes the activities under LLNL Purchase Order B089367, the purpose of which is to ''support the University/Lawrence Livermore National Laboratory Magnetic Fusion Program by evaluating the status of research relative to other national and international programs and assist in long-range plans and development strategies for magnetic fusion in general and for ITER in particular.'' Two specific subtasks are included: ''to review the LLNL Magnet Technology Development Program in the context of the International Thermonuclear Experimental Reactor Design Study'' and to ''assist LLNL to organize and prepare materials for an International Thermonuclear Experimental Reactor Design Study information meeting.''

  16. BR2 Reactor: Irradiation of fuels

    International Nuclear Information System (INIS)

    Verwimp, A.

    2005-01-01

    Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

  17. Food irradiation receives international acceptance

    International Nuclear Information System (INIS)

    Beddoes, J.M.

    1982-01-01

    Irradition has advantages as a method of preserving food, especially in the Third World. The author tabulates some examples of actual use of food irradiation with dates and tonnages, and tells the story of the gradual acceptance of food irradiation by the World Health Organization, other international bodies, and the U.S. Food and Drug Administration (USFDA). At present, the joint IAEA/FAO/WHO standard permits an energy level of up to 5 MeV for gamma rays, well above the 1.3 MeV energy level of 60 Co. The USFDA permits irradiation of any food up to 10 krad, and minor constituents of a diet may be irradiated up to 5 Mrad. The final hurdle to be cleared, that of economic acceptance, depends on convincing the food processing industry that the process is technically and economically efficient

  18. Internal friction in irradiated textolite

    International Nuclear Information System (INIS)

    Zajkin, Yu.A.; Kozhamkulov, B.A.; Koztaeva, U.P.

    1996-01-01

    Structural relaxation in irradiated textolite of ST and ST-EhTF trade marks presenting pressed material got by method of impregnation of fibreglass by phenole and epoxytriphenole binders relatively. Measuring of temperature dependences of internal friction (TDIF) is carried out in torsional pendulum at oscillation frequency 0.6-1.0 Hz before and after irradiation by stopped gamma-quanta with energy 3 MeV on electron accelerator EhLU-4. α and β peaks, related with segments motion in base and side chains of macromolecular have being observed on TDIF of all textolite. Growth of peaks height after irradiation evident about increase of segments mobility in base chain and about de-freezing of segments in side chains and it could be considered as qualitative measure of radiation destruction rate. Comparison of temperature dependences of internal friction indicates on higher radiation stability of textolite of ST-EhTF trade mark

  19. Irradiation facilitates at the advanced test reactor

    International Nuclear Information System (INIS)

    Grover, Blaine S.

    2006-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

  20. Internal irradiation for cystic craniopharyngioma

    International Nuclear Information System (INIS)

    Kobayashi, Tatsuya; Kageyama, Naoki

    1979-01-01

    Internal irradiation with P-32 chromic phosphate and Au-198 colloid was used to treat cystic craniopharyngioma. A newly developed dosimetric formula, by which the radiation dose can be calculated simultaneously at the cyst wall and at a point far from the radioactive source and the untoward effect of irradiation on surrounding brain tissue can be eliminated, especially in cases in which the wall is thin and can be penetrated by beta emission, was used. Radioactive phosphate or gold was injected into eight craniopharyngioma cysts throught the Ommaya reservoir and a tube inserted at the first craniotomy. All cysts were effectively treated for 3 to 33 months, to eliminate fluid retention or collapse. A collapsed cyst was removed at the second craniotomy and irradiation was histologically shown to be effective. Oculomotor palsy, a side effect of irradiation, occurred 10 days after the injection of 5 mc of P-32 chromic phosphate only in a case of small cysts (5.0 ml) in the supra- and intracellular regions. The thickness of the cyst wall was less than 0.5 mm and the oculomotor nerves were thought to adhere to the wall. Not only the amount of wall dose but also the thickness of the wall and localization of the cyst are important factors in internal irradiation. Sufficient and safer doses which kill tumor cells in the wall and have no side effects, are 9,000 to 30,000 rad. Internal irradiation can be used to treat large cysts of more than 10 ml which are supposedly difficult to remove radically and or multiple cysts. It is effective not only for cystic craniopharyngioma but also for intracrania cystic tumors other than craniopharyngioma, if dosimetry is accurate. (J.P.N.)

  1. Internal irradiation for cystic craniopharyngioma

    International Nuclear Information System (INIS)

    Kobayashi, T.; Kageyama, N.; Ohara, K.

    1981-01-01

    The authors report the results of internal irradiation with labeled chromic phosphate (32P) and gold-198 (198Au) colloid in eight cases of cystic craniopharyngiomas. They used a newly developed dosimetric formula, by which the radiation dose at the cyst wall and at any point far from the radioactive source can be calculated. Ten courses of irradiation in eight patients were carried out by injection of either 32P or 198Au colloid into the cyst through an Ommaya drainage system that had been placed at craniotomy. Follow-up studies ranging from 13 to 156 months revealed that all cysts were effectively treated, with elimination of fluid or collapse of the cyst. This was confirmed by Conray cystography and/or computerized tomography. Not only the dose delivered to the wall but also the thickness of the cyst wall and the location of the cyst are important factors in planning internal irradiation. A safe and adequate dose to the cyst wall could range between 9000 to 30,000 rads for craniopharyngioma. This treatment is suitable for large cysts that are thought to be difficult to remove radically, recurrent cysts resistant to previous treatment, or multiple cysts. Internal irradiation may also be applicable in other cystic intracranial tumors if dosimetry is calculated accurately

  2. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

  3. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985

  4. Maintenance technologies for reactor internals

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Kenji [Nuclear Energy Systems and Services Div., Toshiba Corp., Tokyo (Japan); Kobayashi, Masahiro [Toshiba Corp., Yokohama (Japan). Keihin Product Operations; Sano, Yuji; Kimura, Seiichiro [Power and Industrial Systems Research and Development Center, Toshiba Corp., Tokyo(Japan)

    2000-10-01

    Toshiba places the highest priority on maintenance technologies for the reactor pressure vessel (RPV) and its internals in operating nuclear power plants. This paper summarizes the status of applied laser maintenance technologies, both preventive and repair. For laser peeing and laser desensitization treatment (LDT) technologies in particular, field applications are also described in detail. In the future, the area of field applications for preventive maintenance, repair, and inspection technologies will be further expanded. (author)

  5. Irradiation devices for fusion reactor materials results obtained from irradiated lithium aluminate at the OSIRIS reactor

    International Nuclear Information System (INIS)

    Lefevre, F.; Thevenot, G.; Rasneur, B.; Botter, F.

    1986-06-01

    Studies about controlled fusion reactor of the Tokamak type require the examination of the radiation effects on the behaviour of various potential materials. Thus, in the first part of this paper, are presented the devices adapted to these materials studies and used in the OSIRIS reactor. In a second part, is described an experiment of irradiation ceramics used as candidates for breeding material and are given the first results

  6. Internal friction in irradiated silicon

    International Nuclear Information System (INIS)

    Kalanov, M.U.; Pajzullakhanov, M.S.; Khajdarov, T.; Ummatov, Kh.

    1999-01-01

    The submicroscopic heterogeneities in mono- and polycrystal silicon and the influence of X-ray radiation on them were investigated using the ultrasound resonance method. Disk-shaped samples of 27.5 mm in diameter and 4 mm in thickness, with the flat surface parallel to crystallographic plane (111), were irradiated by X-ray beam of 1 Wt/cm 2 (50 KeV, Mo K α ) during 10 hours. Relations of internal frictions (Q -1 ) of samples and their relative attitude (ψ) - Q -1 (ψ) show that there is a presence of double-humped configuration for monocrystal silicon with the peaks at ψ=900 and 270 degrees. The relations Q -1 (ψ) remain the same after the irradiation. However, the peak width becomes larger. This data show that the configuration and attitude of the heterogeneities remain the same after the irradiation. The double-humped configuration was not discovered for the relations Q -1 (ψ) of polycrystal silicon. It is explained by the fact that there is an isotropic distribution in the content of many blocks and granules

  7. International cooperation on breeder reactors

    International Nuclear Information System (INIS)

    Gray, J.E.; Kratzer, M.B.; Leslie, K.E.; Paige, H.W.; Shantzis, S.B.

    1978-01-01

    In March 1977, as the result of discussions which began in the fall of 1976, the Rockefeller Foundation requested International Energy Associates Limited (IEAL) to undertake a study of the role of international cooperation in the development and application of the breeder reactor. While there had been considerable international exchange in the development of breeder technology, the existence of at least seven major national breeder development programs raised a prima facie issue of the adequacy of international cooperation. The final product of the study was to be the identification of options for international cooperation which merited further consideration and which might become the subject of subsequent, more detailed analysis. During the course of the study, modifications in U.S. breeder policy led to an expansion of the analysis to embrace the pros and cons of the major breeder-related policy issues, as well as the respective views of national governments on those issues. The resulting examination of views and patterns of international collaboration emphasizes what was implicit from the outset: Options for international cooperation cannot be fashioned independently of national objectives, policies and programs. Moreover, while similarity of views can stimulate cooperation, this cannot of itself provide compelling justification for cooperative undertakings. Such undertakings are influenced by an array of other national factors, including technological development, industrial infrastructure, economic strength, existing international ties, and historic experience

  8. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  9. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Condie, K.G.; Daw, J.E.; Taylor, S.C.

    2009-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  10. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    International Nuclear Information System (INIS)

    Chichester, Heather Jean MacLean; Hayes, Steven Lowe; Dempsey, Douglas; Harp, Jason Michael

    2016-01-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  11. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  12. Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

    International Nuclear Information System (INIS)

    Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

    2011-01-01

    The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

  13. The international thermonuclear reactor (ITER)

    International Nuclear Information System (INIS)

    Fowler, T.K.; Henning, C.D.

    1987-01-01

    Four governmental groups, representing Europe, Japan, USSR and U.S. met in March 1987 to consider a new international design of a magnetic fusion device for the 1990's. An interim group was appointed. The author gives a brief synopsis of what might be thought of as a draft charter. The starting point is the objective of the ITER device, which is summarized as demonstrating both scientific and technical feasibility of fusion. The paper presents an update on the current thinking and technical aspects for the International Thermonuclear Experimental Reactor (ITER). This covers not only what is happening in the U.S. but also some reports of preliminary thinking of the last technical work that occurred in Vienna

  14. Irradiation routine in the IPR-R1 Triga reactor

    International Nuclear Information System (INIS)

    Maretti Junior, F.

    1980-01-01

    Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

  15. Influence of high dose irradiation on core structural and fuel materials in advanced reactors

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA International Working Group on Fast Reactors (IWGFR) periodically organizes meeting to discuss and review important aspects of fast reactor technology. The fifth meeting held in Obninsk, Russian Federation, 16-19 June 1997, was devoted to the influence of high dose irradiation on the mechanical properties of reactor core structural and fuel materials. The proceedings includes the papers submitted at this meeting each with a separate abstract

  16. Thermogravimetric analysis of reactor-neutrons-irradiated LEXAN polycarbonate film

    International Nuclear Information System (INIS)

    Kalsi, P.C.

    2000-01-01

    The effects of reactor-neutrons irradiation on the thermogravimetric (TG) analysis of LEXAN polycarbonate film in air were studied. Irradiation enhances the degradation rate and the effect increases further with increasing neutron fluence. The kinetics of the different steps of degradation were also evaluated from the TG curves. The activation energy values calculated for all the degradation stages decrease on irradiation. (author)

  17. The International Reactor Dosimetry File (IRDF-85)

    International Nuclear Information System (INIS)

    Cullen, D.E.; McLaughlin, P.K.

    1985-04-01

    This document describes the contents of the second version of the International Reactor Dosimetry File (IRDF-85), distributed by the Nuclear Data Section of the International Atomic Energy Agency. This library superseded IRDF-82. (author)

  18. Studies on the behavior of graphite structures irradiated in the Dragon Reactor. Dragon Project report

    Energy Technology Data Exchange (ETDEWEB)

    Everett, M. R.; Graham, L. W.; Ridealgh, F.

    1971-11-15

    Design data for the physical and mechanical property changes which occur in graphite structural and fuel body components irradiated in an HTR are largely obtained from small specimens tested in the laboratory and in materials test reactors. A brief data summary is given. This graphite physics data can be used to predict dimensional changes, internal stress generation and strength changes in the graphite materials of HTR fuel elements irradiated in the Dragon Reactor. In this paper, the results which have been obtained from post-irradiation examination of a number of fuel pins, are compared with prediction.

  19. Prospects of international trade in irradiated foods

    International Nuclear Information System (INIS)

    Loaharanu, P.

    1990-01-01

    Irradiation is gaining recognition as a physical process for reducing food losses, enhancing hygienic quality of food and facilitating food trade. At present, 36 countries have approved the use of irradiation for processing collectively over 40 food items either on an unconditional or restricted basis. Commercial use of irradiated foods and food ingredients is being carried out in 22 countries. Technology transfer on food irradiation is being intensified to local industry in different regions. Worldwide, a total of 40 commercial/demonstration irradiators available for treating foods have been or are being constructed. Acceptance and control of international trade in irradiated foods were discussed at the International Conference on the Acceptance, Control of and Trade in Irradiated Food, jointly convened by FAO, IAEA, WHO and ITC-UNCTAD/GATT in Geneva, Switzerland, 12-16 December 1988. An ''International Document on Food Irradiation'' was adopted by consensus at this Conference which will facilitate wider acceptance and control of international trade in irradiated foods. (author)

  20. The post-irradiated examination of CANDU type fuel irradiated in the Institute for Nuclear Research TRIGA reactor

    International Nuclear Information System (INIS)

    Tuturici, I.L.; Parvan, M.; Dobrin, R.; Popov, M.; Radulescu, R.; Toma, V.

    1995-01-01

    This post-irradiation examination work has been done under the Research Contract No. 7756/RB, concluded between the International Atomic Energy Agency and the Institute for Nuclear Research. The paper contains a general description of the INR post-irradiation facility and methods and the relevant post-irradiation examination results obtained from an irradiated experimental CANDU type fuel element designed, manufactured and tested by INR in a power ramp test in the 100 kW Pressurised Water Irradiation Loop of the TRIGA 14 MW(th) Reactor. The irradiation experiment consisted in testing an assembly of six fuel elements, designed to reach a bumup of ∼ 200 MWh/kgU, with typical CANDU linear power and ramp rate. (author)

  1. A new materials irradiation facility at the Kyoto university reactor

    International Nuclear Information System (INIS)

    Yoshiie, T.; Hayashi, Y.; Yanagita, S.; Xu, Q.; Satoh, Y.; Tsujimoto, H.; Kozuka, T.; Kamae, K.; Mishima, K.; Shiroya, S.; Kobayashi, K.; Utsuro, M.; Fujita, Y.

    2003-01-01

    A new materials irradiation facility with improved control capabilities has been installed at the Kyoto University Reactor (KUR). Several deficiencies of conventional fission neutron material irradiation systems have been corrected. The specimen temperature is controlled both by an electric heater and by the helium pressure in the irradiation tube without exposure to neutrons at temperatures different from the design test conditions. The neutron spectrum is varied by the irradiation position. Irradiation dose is changed by pulling the irradiation capsule up and down during irradiation. Several characteristics of the irradiation field were measured. The typical irradiation intensity is 9.4x10 12 n/cm 2 s (>0.1 MeV) and the irradiation temperature of specimens is controllable from 363 to 773 K with a precision of ±2 K

  2. Age-related degradation of boiling water reactor vessel internals

    International Nuclear Information System (INIS)

    Ware, A.G.; Shah, V.N.

    1992-01-01

    Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection. (orig.)

  3. Proceedings of the international symposium on materials testing reactors

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Kawamura, Hiroshi

    2009-01-01

    This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

  4. International standardization of instruments for neutron irradiation tests

    International Nuclear Information System (INIS)

    Tanimoto, Masataka; Shibata, Akira; Nakamura, Jinichi; Tsuchiya, Kunihiko; Cho, M.; Lee, C.; Park, S.; Choo, K.

    2012-01-01

    The JMTR in JAEA and HANARO in KAERI are the foremost testing/research reactors in the world and these are expected to contribute to many nuclear fields. As a part of instrument development in irradiation field, information exchange of instruments started from 2010 under the cooperation agreements between KAERI and JAEA. The instruments developed in JMTR and HANARO are introduced and cooperation experiments as future plan are discussed for international standardization. (author)

  5. Irradiation can for the activation of materials in nuclear reactors

    International Nuclear Information System (INIS)

    Schneider, B.; Findeisen, A.; Katzmann, H.

    1985-01-01

    The invention is concerning with an irradiation can for the activation of materials in nuclear reactors in particular for materials with a high heat generation due to irradiation. A good heat transfer between the irradiated material and the irradiation can environment has been guaranteed by a special can design. The outside of the can consists of a tube or a tube bandle which has been formed as a water guide tube. One or more tubes containing the irradiated materials have been positioned at the inner areas of the irradiated can

  6. Conceptual design for simulator of irradiation test reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Ohto, Tsutomu; Magome, Hirokatsu; Izumo, Hironobu; Hori, Naohiko

    2012-03-01

    A simulator of irradiation test reactors has been developed since JFY 2010 for understanding reactor behavior and for upskilling in order to utilize a nuclear human resource development (HRD) and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR, one of the irradiation test reactors, and it simulates operation, irradiation tests and various kinds of accidents caused by the reactor and irradiation facility. The development of the simulator is sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. The training using the simulator will be started for the nuclear HRD from JFY 2012. This report summarizes the result of the conceptual design of the simulator in JFY 2010. (author)

  7. Surveillance of irradiation embrittlement of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Najzer, M.

    1982-01-01

    Surveillance of irradiation embrittlement of nuclear reactor pressure vessels is briefly discussed. The experimental techniques and computer programs available for this work at the J. Stefan Institute are described. (author)

  8. Radioresistance and immunization effectiveness under internal irradiation

    International Nuclear Information System (INIS)

    Kal'nitskij, S.A.

    1978-01-01

    The effect of preliminary immunization on the radioresistance of mice to internal irradiation from incorporated 137 Cs or 90 Sr was studied, and it was found that a preliminary single immunization with bacterial vaccines had a favorable effect on the outcome of radiation injury. The present results suggested that vaccination had a very pronounced radioprotective effect and so may be used as a means of biologic protection from internal irradiation

  9. International Facility for Food Irradiation Technology

    International Nuclear Information System (INIS)

    Farkas, J.

    1982-01-01

    The International Facility for Food Irradiation Technology (IFFIT) was set up in November 1978 for a period of five years at the Pilot Plant for Food Irradiation, Wageningen, The Netherlands under an Agreement between the FAO, IAEA and the Ministry of Agriculture and Fisheries of the Government of the Netherlands. Under this Agreement, the irradiation facilities, office space and services of the Pilot Plant for Food Irradiation are put at IFFIT's disposal. Also the closely located Research Foundation, ITAL, provides certain facilities and laboratory services within the terms of the Agreement. The FAO and IAEA contribute US-Dollar 25,000. Annually for the duration of IFFIT. (orig.) [de

  10. Reliability test for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Uchiyama, Junichi

    1998-01-01

    41 transparencies were presented on the subject of 'Reliability test for reactor internals rejuvenation technology'. The items presented give an introduction on the management of plant life in Japan and introduce the Nuclear Power Engineering Corporation (NUPEC). The question of what reliability tests for rejuvenation of reactor internals are is discussed in some detail and an outline of each test is given. Altogether six methods to rejuvenate reactor internals are presented, two of which have already been applied to actual plants. The presentation was supported by many detailed drawings and images

  11. ITER at the international conference on fusion reactor materials

    International Nuclear Information System (INIS)

    Kalinin, G.; Barabash, V.; Matera, R.

    1998-01-01

    The reports summarizes the topics of the eighth International Conference on Fusion Reactor Materials (ICFRM-8) which was held in Sendai, Japan, on 26-31 October 1997. The ICFRM is focused on the whole spectrum of materials and technologies to be applied in fusion reactors and related facilities. The total number of conference participants was over 500, representing 24 countries and about 600 oral and poster papers were presented at the conference. Three sessions were devoted to ITER materials: (i) Design-Materials Interface and ITER (oral session); (ii) ITER, Irradiation Facility and Technology, (poster session); (iii) ITER and Beyond (discussion session)

  12. International acceptance of irradiated food. Legal aspects

    International Nuclear Information System (INIS)

    1979-01-01

    The three international organizations competent in the field of irradiation processing for the preservation of food (FAO, WHO, IAEA), convened, at the end of 1977, an Advisory Group to revise and update the recommendations of a similar group which met in early 1972. The Advisory Group considered how national regulations could be harmonized so as to facilitate the international movement of irradiated food. This publication contains the Report of the Advisory Group, which summarizes the considerations of the Group on regulatory control over the irradiation plant and irradiation of foods, and on assurances for comparability of control (international labelling and documentation). Annexes 1 to 6 are included in order to complete the relevant information on the legal aspects of this subject. They include a Draft General Standard for Irradiated Foods, a Draft Code of Practice for the Operation of Radiation Facilities Used for the Treatment of Foods, Recommendations of a Consultation Group on the Legal Aspects of Food Irradiation, a Listing of the Legislation on Food Irradiation Adopted in Member States (1971-1976), and Model Regulations for the Control of and Trade in Irradiated Food

  13. Defects investigation in neutron irradiated reactor steels by positron annihilation

    International Nuclear Information System (INIS)

    Slugen, V.

    2003-01-01

    Positron annihilation spectroscopy (PAS) based on positron lifetime measurements using the Pulsed Low Energy Positron System (PLEPS) was applied to the investigation of defects of irradiated and thermally treated reactor pressure vessel (RPV) steels. PLEPS results showed that the changes in microstructure of the RPV-steel properties caused by neutron irradiation and post-irradiation heat treatment can be well detected. From the lifetime measurements in the near-surface region (20-550 nm) the defect density in Russian types of RPV-steels was calculated using the diffusion trapping model. The post-irradiation heat treatment studies performed on non-irradiated specimens are also presented. (author)

  14. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  15. Reactor Structure Materials: Corrosion of Reactor Core Internals

    International Nuclear Information System (INIS)

    Van Dyck, S.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on the corrosion of reactor core internals are: (1) to gain mechanistic insight into the Irradition Assisted Stress Corrosion Cracking (IASCC) phenomenon by studying the influence of separate parameters in well controlled experiments; (2) to develop and validate a predictive capability on IASCC by model description and (3) to define and validate countermeasures and monitoring techniques for application in reactors. Progress and achievements in 1999 are described

  16. Irradiation Facilities at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2005-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

  17. Review of irradiation experiments for water reactor safety research

    International Nuclear Information System (INIS)

    Tobioka, Toshiaki

    1977-02-01

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

  18. Irradiation devices at the upgraded research reactor BER II

    International Nuclear Information System (INIS)

    Gawlik, D.; Robertson, T.

    1992-06-01

    An overview is given of those properties of the BER II research reactor which are important for carrying out irradiation experiments. The subsequent section describes the irradiation devices currently installed in the reactor, or which are under construction, and some of the experiments which can be conducted using them. The field of application of these experiments extends from the study of the metabolism of trace elements in man, employing a highly sensitive element analysis, via radiation damage of high-tech materials, to the identification of paintings of the old masters. The report concludes with a review of the technical details of the irradiation devices, giving information of interest for potential users. (orig.)

  19. Modeling irradiation embrittlement in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  20. New pallet irradiator for Gammaster International

    International Nuclear Information System (INIS)

    Moehlmann, J.H.F.

    1993-01-01

    State of the art technology offers many opportunities to improve production processes with ionizing radiation. Internal and external audits, in particular by regulatory bodies such as the FDA and DHSS/MCA, are necessary and commonplace requirements when operating in the healthcare and pharmaceutical sectors. Efficiency is a major requirement for industry and also for our radiation business. With all the above mentioned items in mind, Gammaster International B.V. has developed and built a new type Pallet Irradiator. For this irradiator Gammaster International designed and built a new process console, incorporating state of the art technology. (author)

  1. Effects of cryogenic reactor irradiation on organic insulators

    International Nuclear Information System (INIS)

    Kato, Teruo

    1995-01-01

    Insulators for the superconducting magnets of fusion reactor are classified as electrical and thermal insulators for which tough organic materials will be used. When the magnet is exposed by fast neutrons and gamma-rays from plasma in a fusion reactor, the fusion reactor systems will cause fatal damage by the degradation of insulators. Therefore, it is necessary to select materials resistant irradiation damage for use as insulators. Electrical and mechanical tests were carried out at 4.2 K without warmup after the reactor irradiation at 5 K. The effects of reactor irradiation at the dose of 10 7 Gy on epoxy resins (bisphenol-A), G-10 CR, VL-E 200 and G-11 CR caused large decreases in mechanical strength. Polyetheretherketone (PEEK), polyimide and phenol novolac resins, which were used to laminate reinforced plastics with glass-cloth against irradiation, showed good resistance. Effects of cryogenic reactor irradiation on several organic materials and epoxy laminate-reinforced plastics with glass-cloth and Kevlar-cloth were also discussed. (author)

  2. Standard irradiation facilities for use in TRIGA reactors

    International Nuclear Information System (INIS)

    Kolbasov, B.N.; Luse, R.A.

    1972-01-01

    The standard neutron irradiation facility (SNIP) was developed under IAEA and FAO co-ordinated research program for the standardization of neutron irradiation facilities for radiobiological research, resulting in the possibility to use fast neutrons from pool-type reactors for radiobiological studies. The studies include irradiation of seeds for crop improvement, of Drosophila for genetic studies, and of microorganisms for developing industrially useful mutants, as well as fundamental studies in radiation biology. The facilities, located in the six pool-type reactors (in Austria, Bulgaria, India, Philippines, Thailand and Taiwan), have been calibrated and utilized to compare the response to fast neutrons of barley seeds (variety Himalaya CI 000620) which were selected as a standard biological monitor by which to estimate neutron fluxes in different reactors. These comparative irradiation studies showed excellent agreement and reproducibility

  3. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10 6 R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  4. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan)] [and others

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10{sup 6} R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  5. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

    1993-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

  6. Calculations on neutron irradiation damage in reactor materials

    International Nuclear Information System (INIS)

    Sone, Kazuho; Shiraishi, Kensuke

    1976-01-01

    Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

  7. Neutron irradiation experiments for fusion reactor materials through JUPITER program

    International Nuclear Information System (INIS)

    Abe, K.; Namba, C.; Wiffen, F.W.; Jones, R.H.

    1998-01-01

    A Japan-USA program of irradiation experiments for fusion research, ''JUPITER'', has been established as a 6 year program from 1995 to 2000. The goal is to study ''the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment''. This is phase-three of the collaborative program, which follows RTNS-II program (phase-1: 1982-1986) and FFTF/MOTA program (phase-2: 1987-1994). This program is to provide a scientific basis for application of materials performance data, generated by fission reactor experiments, to anticipated fusion environments. Following the systematic study on cumulative irradiation effects, done through FFTF/MOTA program. JUPITER is emphasizing the importance of dynamic irradiation effects on materials performance in fusion systems. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. (orig.)

  8. Analytical electron microscopy of neutron-irradiated reactor alloys

    International Nuclear Information System (INIS)

    Thomas, L.E.

    1982-01-01

    Exposure to the high neutron fluxes and temperatures from 400 to 650 0 C in the core region of a fast breeder reactor profoundly alters the microstructure and properties of structural steels and superalloys. The development of irradiation-induced voids, dislocations and precipitates, as well as segregation of alloying elements on a microscopic scale has been related to macroscopic swelling, creep, hardening and embrittlement which occur during prolonged exposures in reactor. Microanalytical studies using TEM/STEM methods, primarily energy dispersive x-ray (EDX) microanalysis, have greatly aided understanding of alloy behavior under irradiation. The main uses of analytical electron microscopy in studying irradiated alloys have been the identification of irradiation-induced precipitates and determination of the changes in local composition due to irradiation-induced solute segregation

  9. Gamma irradiation plants using reactor fuel elements

    International Nuclear Information System (INIS)

    Suckow, W.

    1976-11-01

    Recent irradiation plants utilizing fuel elements are described. Criteria for optimizing such plants, evaluation of the plants realized so far, and applications for the facilities are discussed. (author)

  10. Fabrication of internally instrumented reactor fuel rods

    International Nuclear Information System (INIS)

    Schmutz, J.D.; Meservey, R.H.

    1975-01-01

    Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

  11. Reliability tests for reactor internals replacement technology

    International Nuclear Information System (INIS)

    Fujimaki, K.; Uchiyama, J.; Ohtsubo, T.

    2000-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  12. Reliability tests for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Fujimaki, Katsumi; Hitoki, Yoichi; Otsubo, Toru; Uchiyama, Junichi

    1998-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for rejuvenating reactor internals which has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995. The project follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the test plans and results which are directed at preventive maintenance before damage and repair after damage for reactor internals aging degradation. The test results for the replacement methods of ICM housing and BWR core shroud have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  13. DNA damage induced by radionuclide internal irradiation

    International Nuclear Information System (INIS)

    Cui Fengmei; Zhao Jingyong; Hong Chengjiao; Lao Qinhua; Wang Liuyi; Yang Shuqin

    2004-01-01

    Objective: To study the DNA damage of peripheral blood mononuclear cell (PBMC) in rats exposed to radionuclide internal irradiation. Methods: The radionuclides were injected into the rats and single cell get electrophoresis (SCGE) was performed to detect the length of DNA migration in the rat PBMC. Results: DNA migration in the rat PBMC increased with accumulative dose or dose-rate. It showed good relationship of dose vs. response and of dose-rate vs. response, both relationship could be described as linear models. Conclusion: Radionuclide internal irradiation could cause DNA damage in rat PBMC. (authors)

  14. Lifetime assessment on PWR reactor vessel internals in Korea

    International Nuclear Information System (INIS)

    Jung, Sung-Gyu; Jin, Tae-Eun; Jeong, Ill-Seok

    2002-01-01

    In order to extend the operating time of the Kori Unit 1 reactor internals, a comprehensive review of the potential ageing problems and a safety assessment have been performed. As the plant ages, reactor internal components which are subject to various ageing mechanism should be identified and evaluated based on the systematic technical procedure. In this respect, technical procedure for lifetime evaluation had been developed and applied to reactor internals. This paper describes a overall assessment and ageing management procedure and evaluation results for reactor internals. Also this paper suggests the optimal ageing management programs to maintain the integrity of reactor internals beyond design life based on the evaluation results. A review of all known potential ageing mechanisms was performed for each of the reactor internal subcomponents. From these results, 8 ageing mechanisms such as void swelling, irradiation and thermal embrittlement, fatigue, stress corrosion cracking, IASCC, stress relaxation, and wear for the reactor internal components were expected to be of major concerns during the current or extended plant life. In this study, 8 ageing mechanisms were identified for lifetime evaluation. For these ageing mechanisms, lifetime assessment was performed. As a result of this evaluation, it is expected that core barrel will exceed the IASCC threshold value during 40 operating years, and baffle/former and baffle former bolts will exceed the threshold value for void swelling, irradiation embrittlement, IASCC, stress relaxation during 40 operating years. However, for all other reactor internals subcomponents, thermal embrittlement, fatigue, SCC, and wear were identified as nonsignificant. As a result of lifetime evaluations, 4 ageing mechanisms were established to be plausible for 3 subcomponents. These results are shown. The existing ageing management programs (AMPs) for Kori Unit 1, such as ISI, water chemistry control, rod drop time testing etc., were

  15. Irradiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Bros, J.

    2000-01-01

    From the historical decision of closing the Yankee Rowe NPP because of the uncertainties on the level of reactor pressure vessel neutron embrittlement, this paper reviews the technical-scientist bases of the degradation phenomena, and refers to the evolution of reactor pressure vessel radiation surveillance programs. (Author)

  16. New irradiation devices at the FRN reactor

    International Nuclear Information System (INIS)

    Stark, W.

    1980-01-01

    In order to fulfill the experimental demands three additional devices were constructed and installed. The first is a vertical irradiation tube in air surrounded by a lead cylinder (in the irradiation position). The second device is a rabbit system ending within the graphite moderator of the thermal column. The third device is so called rotating disk assembly, built to replace the rotary specimen rack

  17. 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.)

  18. International standards and agreements in food irradiation

    International Nuclear Information System (INIS)

    Cetinkaya, N.

    2004-01-01

    Full text: The economies of both developed and developing countries have been effected by their exported food and agricultural products. Trading policies of food and agricultural products are governed by international agreement as well as national regulations. Trade in food and agricultural commodities may be affected by both principal Agreements within the overall World Trade Organization (WTO) Agreement, though neither specifically refers to irradiation or irradiated foods. The principal Agreements are the Technical Barriers to Trade (TBT) Agreement and the Sanitary and Phyto sanitary (SPS) Agreement. The SPS of the WTO requires governments to harmonize their sanitary and phyto sanitary measures on as wide basis as possible. Related standards, guidelines and recommendations of international standard setting bodies such as the Codex Alimentarius Commission (food safety); the International Plant Protection Convention (IPPC) (plant health and quarantine); and International Office of Epizootic (animal health and zoo noses) should be used in such a harmonization. International Standards for Phyto sanitary Measures (ISPM) no.18 was published under the IPPC by FAO (April 2003, Rome-Italy). ISPM standard provides technical guidance on the specific procedure for the application of ionizing radiation as a phyto sanitary treatment for regulated pests or articles. Moreover, Codex Alimentarius Commission, Codex General Standard for Irradiated Foods (Stand 106-1983) and Recommended International Code of Practice were first published in 1983 and revised in March 2003. Scope of this standard applies to foods processed by ionizing radiation that is used in conjunction with applicable hygienic codes, food standards and transportation codes. It does not apply to foods exposed to doses imparted by measuring instruments used for inspection purposes. Codex documents on Principles and Guidelines for the Import/Export Inspection and Certification of Foods have been prepared to guide

  19. International standards and agreements in food irradiation

    International Nuclear Information System (INIS)

    Cetinkaya, N.

    2004-01-01

    The economies of both developed and developing countries have been effected by their exported food and agricultural products. Trading policies of food and agricultural products are governed by international agreement as well as national regulations. Trade in food and agricultural commodities may be affected by both principal Agreements within the overall World Trade Organization (WTO) Agreement, though neither specifically refers to irradiation or irradiated foods. The principal Agreements are the Technical Barriers to Trade (TBT) Agreement and the Sanitary and Phyto sanitary (SPS) Agreement. The SPS of the WTO requires governments to harmonize their sanitary and phyto sanitary measures on as wide basis as possible. Related standards, guidelines and recommendations of international standard setting bodies such as the Codex Alimentarius Commission (food safety); the International Plant Protection Convention (IPPC) (plant health and quarantine); and International Office of Epizootic (animal health and zoo noses) should be used in such a harmonization. International Standards for Phyto sanitary Measures (ISPM) no.18 was published under the IPPC by FAO (April 2003, Rome-Italy). ISPM standard provides technical guidance on the specific procedure for the application of ionizing radiation as a phyto sanitary treatment for regulated pests or articles. Moreover, Codex Alimentarius Commission, Codex General Standard for Irradiated Foods (Stand 106-1983) and Recommended International Code of Practice were first published in 1983 and revised in March 2003. Scope of this standard applies to foods processed by ionizing radiation that is used in conjunction with applicable hygienic codes, food standards and transportation codes. It does not apply to foods exposed to doses imparted by measuring instruments used for inspection purposes. Codex documents on Principles and Guidelines for the Import/Export Inspection and Certification of Foods have been prepared to guide international

  20. New pallet irradiation for Gammaster International

    Science.gov (United States)

    Möhlmann, Mr. J. H. F.

    1993-07-01

    State of the art technology offers many opportunities to improve production process with ionizing radiation. Also the requirements to this process become more and more stringent; who was aware of the ISO-9000 Standards? Now it is a necessary concept? Internal and external audits, in particular by Regulatory Bodies such as the FDA and DHSS/MCA, are necessary and commonplace requirements when operating in the Healthcare and Pharmaceutical sectors. Efficiency is a major requirement for industry and also for our radiation business. With all the above mentioned items in mind, Gammaster International B.V. in close cooperation with Nordion International Inc. has developed and built a new type Pallet Irradiator. In this new concept, our 22 years experience in gamma processing and the expertise gained in 10 years of operation of pallet irradiators was incorporated. For this irradiator Gammaster International designed and built a new process control console, incorporating state of the art technology such as: - A well-balanced safety system - A control panel, with status indication for all essential operating and safety parameters - Advances warning for necessary maintenance work - Automatic timer-setting control in addition to an automatic timer setting modification - Clear indication of the position of the product inside the unit and its status - Fully automatic product information generated when the product leaves the irradiation unit - Visual process control with the help of a color monitor - Incremental dose-control - Fault identification

  1. Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

    International Nuclear Information System (INIS)

    Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

    2011-01-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

  2. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1982-06-01

    This publication is the third in a proposed series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. Subsequent issues in this series will update the information contained herein. This publication contains basically three kinds of information: (1) routes approved by the Commission for the shipment of irradiated reactor fuel, (2) information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and (3) cumulative amounts of material shipped

  3. Microstructural evolution in neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    English, C.A.; Phythian, W.J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhujie [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Bartels, David [Univ. of Notre Dame, IN (United States)

    2015-04-02

    This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that the effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.

  6. Improvement of neutron irradiation field of research reactors for BNCT

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    1992-01-01

    The modification of research reactors for an improvement of the irradiation field for BNCT has been investigated in comparison with the field characteristics of the 'old' configuration at the Musashi reactor. The new point of this study is that the evaluation has been done by using an arrangement including both the facility structure and a whole-body phantom, and also by considering the whole-body absorbed dose. (author)

  7. Low temperature irradiation facility at Kyoto University Reactor (KUR)

    International Nuclear Information System (INIS)

    Atobe, Kozo; Okada, Moritami; Yoshida, Hiroyuki; Kodaka, Hisao; Miyata, Kiyomi.

    1977-01-01

    A new refrigeration system has been substituted to the low temperature irradiation facility at KUR instead of the previous one, since April in 1975. The model 1204 CTi He liquifier was designed to be modified for the refrigerator with the capacity of 30 watts at 10 K. The refrigeration capacity of 38 watts at 10 K was defined using a special cryostat and transfer-tubes, and the lowest temperature of about 18 K was measured using the irradiation loop without reactor operation. The reconstructed facility enables us to hold the many specimens simultaneously in the sample chamber of the irradiation loop at about 25 K during reactor operation of 5 MW. The irradiation dose has been reached about 6.6 x 10 16 n sub(f)/cm 2 and 6.1 x 10 17 n sub(th)/cm 2 with the normal reactor operation cycle of up to 77 hours. The stable operation condition of the machine and the special safety system for the refrigeration system enable us to maintain easily the facility with a constant operation condition for such a long time irradiation. Many kinds of low temperature neutron irradiation experiments are carried out using the facility, which techniques are partially reported. (auth.)

  8. ITER [International Thermonuclear Experimental Reactor] reactor building design study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

  9. Actinides in irradiated graphite of RBMK-1500 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Plukienė, R., E-mail: rita@ar.fi.lt; Plukis, A.; Barkauskas, V.; Gudelis, A.; Gvozdaitė, R.; Duškesas, G.; Remeikis, V.

    2014-10-01

    Highlights: • Activation of actinides in the graphite of the RBMK-1500 reactor was analyzed. • Numerical modeling using SCALE 6.1 and MCNPX was used for actinide calculation. • Measurements of the irradiated graphite sample were used for model validation. • Results are important for further decommissioning process of the RBMK type reactors. - Abstract: The activation of graphite in the nuclear power plants is the problem of high importance related with later graphite reprocessing or disposal. The activation of actinide impurities in graphite due to their toxicity determines a particular long term risk to waste management. In this work the activation of actinides in the graphite constructions of the RBMK-1500 reactor is determined by nuclear spectrometry measurements of the irradiated graphite sample from the Ignalina NPP Unit I and by means of numerical modeling using two independent codes SCALE 6.1 (using TRITON-VI sequence) and MCNPX (v2.7 with CINDER). Both models take into account the 3D RBMK-1500 reactor core fragment with explicit graphite construction including a stack and a sleeve but with a different simplification level concerning surrounding graphite and construction of control roads. The verification of the model has been performed by comparing calculated and measured isotope ratios of actinides. Also good prediction capabilities of the actinide activation in the irradiated graphite have been found for both calculation approaches. The initial U impurity concentration in the graphite model has been adjusted taking into account the experimental results. The specific activities of actinides in the irradiated RBMK-1500 graphite constructions have been obtained and differences between numerical simulation results, different structural parts (sleeve and stack) as well as comparison with previous results (Ancius et al., 2005) have been discussed. The obtained results are important for further decommissioning process of the Ignalina NPP and other RBMK

  10. Combined effect of external and internal irradiation

    International Nuclear Information System (INIS)

    Kiradzhiev, G.

    1987-01-01

    Some of the general regularities of the combined effect of external irradiation and iodine 131 are discussed. Data are adduced showing that modification of the effects of these two radiation factors, when jointly applied, is also determined by the quantitative relations of the applied doses of external and internal irradiation, referred to a particular moment of the effects. It was shown that the effects of the radionuclides in these combined radiation injuries are basically realized by two mechanisms: 1. changes are found in the radionuclide kinetic parameters (nonspecific effects); 2. changes in their kinetic parameters are absent (specific effect). These two mechanisms underlie different approaches to therapy

  11. Behavior of high Tc-superconductors and irradiated defects under reactor irradiation

    International Nuclear Information System (INIS)

    Atobe, Kozo; Honda, Makoto; Fukuoka, Noboru; Yoshida, Hiroyuki.

    1991-01-01

    It has been well known that the lattice defects of various types are introduced in ceramics without exception, and exert large effect to the function of these materials. Among oxides, the electronic materials positively using oxygen defect control have been already put in practical use. Also in the oxide high temperature superconductors which are Perovskite type composite oxides, the superconductive characteristics are affected largely by the concentration of the oxygen composing them. This is regarded as an important factor for causing superconductivity, related with the oxygen cavities arising at this time and the carriers bearing superconductivity. In this study, the irradiation effect with relatively low dose, the measurement under irradiation, the effect of irradiation temperature, and the effect of radiation quality were evaluated by the irradiation of YBCO, EBCO and LBCO. The experimental method, and the irradiation effect at low temperature and normal temperature, the effect of Co-60 gamma ray irradiation instead of reactor irradiation are reported. (K.I.)

  12. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    International Nuclear Information System (INIS)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste

  13. The International Thermonuclear Experimental Reactor configuration evolution

    International Nuclear Information System (INIS)

    Lousteau, D.C.; Nelson, B.E.; Lee, V.D.; Thomson, S.L.; Miller, J.M.; Lindquist, W.B.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) conceptual design activities consist of two phases: a definition phase, completed in September 1988, and a design phase, now in progress. The definition phase was successful in identifying a consistent set of technical characteristics and the broad definition of the required reactor configuration and hardware. Scheduled for completion in November 1990, the design phase is producing a more detailed definition of the required components, a first cost estimate, and a description of site requirements. A major activity in the ITER design phase is the period of joint work conducted at the Max Planck Institute for Plasma Physics, Garching, Federal Republic of Germany, from June through October 1989. An official report of the findings and conclusions of this activity will be submitted to and published by the International Atomic Energy Agency (IAEA). This paper highlights the evolution of the reactor mechanical configuration since the conclusion of the definition phase. 8 figs., 2 tabs

  14. Cerenkov methodology for monitoring irradiated reactor fuel

    International Nuclear Information System (INIS)

    Nicholson, N.; Dowdy, E.J.

    1984-01-01

    Attribute measurement methods for confirming declared irradiated fuel inventories at nuclear installations under safeguards surveillance are of significant interest to inspectors. High-gain measurements of the intensity of the Cerenkov glow from exposed assemblies in water-filled storage ponds are promising for this purpose because the measured intensities depend on cooling times and burnup. We have developed a Cerenkov Measuring Device, a hand-held instrument that examines irradiated fuel assemblies in water-filled storage ponds and measures the intensity of the associated Cerenkov glow. In addition, we have developed a method for making such high-gain measurements in the presence of intense ambient light

  15. Positron annihilation studies on reactor irradiated and thermal annealed ferrocene

    International Nuclear Information System (INIS)

    Marques Netto, A.; Carvalho, R.S.; Magalhaes, W.F.; Sinisterra, R.D.

    1996-01-01

    Retention and thermal annealing following (n, γ) reaction in solid ferrocene, Fe(C 5 H 5 ) 2 , were studied by positron annihilation lifetime spectroscopy (PAL). Positronium (Ps) formation was observed in the non-irradiated compound with a probability or intensity (I 3 ) of 30%. Upon irradiation of the compound with thermal neutrons in a nuclear reactor, I 3 decreases with increasing irradiation time. Thermal treatment again increases I 3 values from 16% to 25%, revealing an important proportion of molecular reformation without variation of the ortho-positronium lifetime (τ 3 ). These results point out the major influence of the electronic structure as determining the Ps yields in the pure complex. In the irradiated and non irradiated complexes the results are satisfactorily explained on the basis of the spur model. (orig.)

  16. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, Stephen; Campbell, C.A.

    1994-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

  17. Food Irradiation. Proceedings of the International Symposium on Food Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-11-15

    For some years research has been done in several countries, with the object of contributing to the world's food supplies, on the application of nuclear methods to food preservation and processing. The importance of food preservation is of particular relevance in certain regions of the world where up to thirty per cent of harvested foodstuffs are being lost because of damage by animal pests and microorganisms. A series of international meetings have been held on this subject; the first, held in 1958 at Harwell, was followed by further meetings in 1960 in Paris and in 1961 in Brussels. The International Symposium on Food Irradiation organized by the International Atomic Energy Agency and the Food and Agriculture Organization of the United Nations through their Joint Division of Atomic Energy in Agriculture, and held at the Karlsruhe Nuclear Research Centre, Karlsruhe, from 6 to 10 June 1966, at the generous invitation of the Government of the Federal Republic of Germany, is the most recent of this series of meetings. It was held for the purpose of exchanging the most up-to-date results of research, of contributing towards co-operative efforts between Member States, and of stimulating trade in the international exchange of irradiated products between nations. Papers describing research over the past fourteen years were given by outstanding authorities; the results point to a breakthrough having been achieved in the use of ionizing radiation in food preservation, notwithstanding some problems still to be solved, such as overcoming changes in colour, flavour, odour or texture. The Symposium was attended by over 200 scientists from 25 countries and four international organizations. Sixty-nine papers were presented. It was shown that a wide variety of foodstuffs exist for which radiation could be used for three different purposes: to produce indefinitely stable products, to rid food of organisms that constitute health hazards, and to extend the normal shelf or market life

  18. Food Irradiation. Proceedings of the International Symposium on Food Irradiation

    International Nuclear Information System (INIS)

    1966-01-01

    For some years research has been done in several countries, with the object of contributing to the world's food supplies, on the application of nuclear methods to food preservation and processing. The importance of food preservation is of particular relevance in certain regions of the world where up to thirty per cent of harvested foodstuffs are being lost because of damage by animal pests and microorganisms. A series of international meetings have been held on this subject; the first, held in 1958 at Harwell, was followed by further meetings in 1960 in Paris and in 1961 in Brussels. The International Symposium on Food Irradiation organized by the International Atomic Energy Agency and the Food and Agriculture Organization of the United Nations through their Joint Division of Atomic Energy in Agriculture, and held at the Karlsruhe Nuclear Research Centre, Karlsruhe, from 6 to 10 June 1966, at the generous invitation of the Government of the Federal Republic of Germany, is the most recent of this series of meetings. It was held for the purpose of exchanging the most up-to-date results of research, of contributing towards co-operative efforts between Member States, and of stimulating trade in the international exchange of irradiated products between nations. Papers describing research over the past fourteen years were given by outstanding authorities; the results point to a breakthrough having been achieved in the use of ionizing radiation in food preservation, notwithstanding some problems still to be solved, such as overcoming changes in colour, flavour, odour or texture. The Symposium was attended by over 200 scientists from 25 countries and four international organizations. Sixty-nine papers were presented. It was shown that a wide variety of foodstuffs exist for which radiation could be used for three different purposes: to produce indefinitely stable products, to rid food of organisms that constitute health hazards, and to extend the normal shelf or market life

  19. Numerical analysis of reactor internals under hydrodynamic loads

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Da Hye; Chang, Yoon Suk [Kyung Hee Univ., Yongin (Korea, Republic of); Jhung, Myung Jo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In the present study, six kinds of major equipment of a typical reactor internals were identified by incorporating recent research trend. Based on this, detailed numerical models were developed and used for establishment of optimum analysis methodology subjected to hydrodynamic loads. As a result, stress values of the major equipment were calculated through the acoustic-structure analysis under periodic hydrodynamic load and the turbulence-structure analysis under random hydrodynamic load. The numerical analysis scheme can be used for development of preventive action plan and management procedures of the reactor internals. Reactor internals installed in a pressure vessel have been exposed to harsh environment such as high neutron irradiation and temperature with complex fluid flow. As the increase of operational years of NPPs(Nuclear Power Plants), possibility of functional loss of the reactor internals is increased due to degradation caused by radiation embrittlement, thermal aging, fatigue, corrosion and FIV(Flow-Induced Vibration) etc. In practice, defects were detected at core support structure as well as upper and lower parts of structural assembly in European and United States NPPs. Recently, in a GALL(Generic Aging Lessons Learned) report, US NRC(Nuclear Regulatory Commission) identified reactor internals as a high priority component and addressed relevant management programs. In Korea, similar activities have been conducted for long-term operation beyond design lifetime but most of them were limited to qualitative evaluation based on examination and maintenance programs. Therefore, not only to reduce repair and replacement efforts but also to secure the stability of NPPs, necessity for development of quantitative evaluation technique as well as establishment of preventive action plan and management procedures is on the rise. The FIV represents the structural vibration phenomenon induced by liquid flow and generally occurs at contact surfaces. In the present

  20. Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

    International Nuclear Information System (INIS)

    Conlin, J.A.

    1985-08-01

    A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 200 0 C. The design description and results of the prototype capsule performance are presented

  1. Education and training by utilizing irradiation test reactor simulator

    International Nuclear Information System (INIS)

    Eguchi, Shohei; Koike, Sumio; Takemoto, Noriyuki; Tanimoto, Masataka; Kusunoki, Tsuyoshi

    2016-01-01

    The Japan Atomic Energy Agency, at its Japan Materials Testing Reactor (JMTR), completed an irradiation test reactor simulator in May 2012. This simulator simulates the operation, irradiation test, abnormal transient change during operation, and accident progress events, etc., and is able to perform operation training on reactor and irradiation equipment corresponding to the above simulations. This simulator is composed of a reactor control panel, process control panel, irradiation equipment control panel, instructor control panel, large display panel, and compute server. The completed simulator has been utilized in the education and training of JMTR operators for the purpose of the safe and stable operation of JMTR and the achievement of high operation rate after resuming operation. For the education and training, an education and training curriculum has been prepared for use in not only operation procedures at the time of normal operation, but also learning of fast and accurate response in case of accident events. In addition, this simulator is also being used in operation training for the purpose of contributing to the cultivation of human resources for atomic power in and out of Japan. (A.O.)

  2. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1983-07-01

    This publication contains basically three kinds of information: routes approved by the Commission for the shipment of irradiated reactor fuel, information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and cumulative amounts of material shipped

  3. Irradiation proposition of ferritic steels in a russian reactor

    International Nuclear Information System (INIS)

    Seran, J.L.; Decours, J.; Levy, L.

    1987-04-01

    Using the low temperatures of russian reactors, a sample irradiation is proposed to study mechanical properties and swelling of martensitic steels (EM10, T91, 1.4914, HT9), ferrito-martensitic (EM12) and ferritic (F17), at temperatures lower than 400 0 C [fr

  4. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    Markgraf, J.F.W.

    1985-01-01

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  5. Irradiation effects on fuels for space reactors

    International Nuclear Information System (INIS)

    Ranken, W.A.; Cronenberg, A.W.

    1984-01-01

    A review of irradiation-induced swelling and gas release experience is presented here for the three principal fuels UO 2 , UC, and UN. The primary advantage of UC and UN over UO 2 is higher thermal conductivity and attendant lower fuel temperature at equivalent pellet diameter and power density, while UO 2 offers the distinct benefit of well-known irradiation performance. Irradiation test results indicate that at equivalent burnup, temperature, and porosity conditions, UC experiences higher swelling than UO 2 or UN. Fission gas swelling becomes important at fuel temperatures above 1320 K for UC, and at somewhat higher temperatures for UO 2 and UN. Evidence exists that at equivalent fuel temperatures and burnups, high density UO 2 and UN experience comparable swelling behavior; however, differences in thermal conductivity influence overall irradiation performance. The low conductivity of UO 2 results in higher thermal gradients which contribute to fuel microcracking and gas release. As a result UO 2 exhibits higher fractional gas release than UN, at least or burnups up to about 3%

  6. Design experiences for medical irradiation field at the musashi reactor

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    1994-01-01

    The design of the medical irradiation field at the Musashi reactor was carried out from 1974 to 1975, about 20 years ago. Various numerical analyses have been carried out recently, and it is astonishing to find out that the performance close to the optimum as a 100 kW reactor has been obtained. The reason for this is that the design was carried out by dividing into the stationary part and the moving part, and as for the moving part, the structure was determined by repeating trial and error and experiments. In this paper, the comparison of the analysis carried out later with the experimental data and the change of the absorbed dose at the time of medical irradiation accompanying the change of neutron energy spectra are reported. As the characteristics of the medical irradiation field at the Musashi reactor, the neutron energy spectra and the absorbed dose and mean medical irradiation time are shown. As the problems in boron neutron capture therapy, the neutron fluence required for the therapy, the way of thinking on background dose, and the problem of determining the irradiation time are discussed. The features of epithermal neutron beam are explained. (K.I.)

  7. BNFL's experience in the sea transport of irradiated research reactor fuel to the USA

    International Nuclear Information System (INIS)

    Hudson, I.A.; Porter, I.

    2000-01-01

    BNFL provides worldwide transport for a wide range of nuclear materials. BNFL Transport manages an unique fleet of vessels, designed, built, and operated to the highest safety standards, including the highest rating within the INF Code recommended by the International Maritime Organisation. The company has some 20 years of experience of transporting irradiated research reactor fuel in support of the United States' programme for returning US obligated fuel from around the world. Between 1977 and 1988 BNFL performed 11 shipments of irradiated research reactor fuel from the Japan Atomic Energy Research Institute to the US. Since 1997, a further 3 shipments have been performed as part of an ongoing programme for Japanese research reactor operators. Where possible, shipments of fuel from European countries such as Sweden and Spain have been combined with those from Japan for delivery to the US. (author)

  8. Calibration of the enigma code for Finnish reactor fuel with support from experimental irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Kelppe, S; Ranta-Puska, K [VTT Energy, Jyvaeskylae (Finland)

    1997-08-01

    Assessment by VTT of the ENIGMA fuel performance code, the original version by Nuclear Electric plc of the UK amended by a set of WWER specific materials correlations, is described. The given examples of results include analyses for BWR 9 x 9 fuel, BWR fuel irradiated in the reinstrumented test of an international Riso project, pre-characterized commercial WWER fuel irradiated in Loviisa reactor in Finland, and instrumented WWER test fuel irradiations in the MR reactor in Russia. The effects of power uncertainty and some model parameters are discussed. Considering the fact that the described cases all mean prototypic application of the code, the results are well encouraging. The importance of the accuracy in temperature calculations is emphasized. (author). 2 refs, 12 figs, 1 tab.

  9. Remote Inspection Techniques for Reactor Internals of Liquid Metal Reactor by using Ultrasonic Waveguide Sensor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Kim, Seok Hun; Lee, Jae Han

    2006-02-01

    The primary components such as a reactor core, heat exchangers, pumps and internal structures of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection and continuous monitoring as major in-service inspection (ISI) methods of reactor internal structures. Reactor core and internal structures of LMR can not be visually examined due to an opaque liquid sodium. The under-sodium viewing and remote inspection techniques by using an ultrasonic wave should be applied for the in-service inspection of reactor internals. The remote inspection techniques using ultrasonic wave have been developed and applied for the visualization and ISI of reactor internals. The under sodium viewing technique has a limitation for the application of LMR due to the high temperature and irradiation environment. In this study, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium viewing and remote inspection. The Lamb wave propagation of a waveguide sensor has been analyzed and the zero-order antisymmetric A 0 plate wave was selected as the application mode of the sensor. The A 0 plate wave can be propagated in the dispersive low frequency range by using a liquid wedge clamped to the waveguide. A new technique is presented which is capable of steering the radiation beam angle of a waveguide sensor without a mechanical movement of the sensor assembly. The steering function of the ultrasonic radiation beam can be achieved by a frequency tuning method of the excitation pulse in the dispersive range of the A 0 mode. The technique provides an opportunity to overcome the scanning limitation of a waveguide sensor. The beam steering function has been evaluated by an experimental verification. The ultrasonic C-scanning experiments are performed in water and the feasibility of the ultrasonic waveguide sensor has been verified. The various remote inspection

  10. Investigation of neutron irradiated reactor vessel steels using post-irradiation annealing techniques

    Energy Technology Data Exchange (ETDEWEB)

    Nakata, Hayato; Fukuya, Koji [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    The matrix damage is known to be a major factor that contributes to embrittlement and hardening of irradiated reactor vessel steels, and is assumed to be composed of the point defect clusters. However field emission gun scanning transmission electron microscopy (FEGSTEM) and atom probe (AP) could not detect any evidence of the matrix damage. In this study, post irradiation annealing experiments combining positron annihilation lineshape analysis (PALA) and hardness experiments were applied to an actual surveillance test specimen and a sample of reactor vessel steel irradiated in a material test reactor (MTR), in order to investigate the matrix damage recovery behavior and its contribution to hardening. It was confirmed that higher fluence increased the hardness and the volume fraction of open volume defects and that higher flux decreased the thermal stability of matrix damage and the effect on hardening. The contribution of matrix damage to hardening could be estimated to be below 30%. (author)

  11. Irradiation creep in reactor graphites for HTR applications. [Neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Veringa, H J; Blackstone, R [Stichting Reactor Centrum Nederland, Petten

    1976-01-01

    A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400/sup 0/C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400/sup 0/C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density.

  12. The Advanced Test Reactor Irradiation Facilities and Capabilities

    International Nuclear Information System (INIS)

    S. Blaine Grover; Raymond V. Furstenau

    2007-01-01

    The Advanced Test Reactor (ATR) is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR's unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments

  13. Meso-scale modeling of irradiated concrete in test reactor

    International Nuclear Information System (INIS)

    Giorla, A.; Vaitová, M.; Le Pape, Y.; Štemberk, P.

    2015-01-01

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  14. Meso-scale modeling of irradiated concrete in test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giorla, A. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Vaitová, M. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic); Le Pape, Y., E-mail: lepapeym@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Štemberk, P. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic)

    2015-12-15

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  15. The irradiation test program for transmutation in the French Phenix fast reactor

    International Nuclear Information System (INIS)

    Guidez, J.; Chaucheprat, P.; Fontaine, B.; Brunon, E.

    2004-01-01

    Put on commercial operation in July 1974, the French fast reactor Phenix reached a 100 000 hours operation time in september 2003. When the French law relative to long lived radioactive waste management was promulgated on December 1991, priority was given to Phenix to be run as a research reactor and to carry on a wide irradiation program dedicated to study transmutation of minor actinides and long-lived fission products. After a major renovation program required to extend the reactor lifetime, Phenix power buildup took place in 2003. Experimental irradiations have been loaded in the core, involving components for heterogeneous and homogeneous transmutation modes, americium targets, technetium 99 metal pins and isolated isotopes for integral cross-sections measurements. Associated post- irradiated examination programs are already underway or planned. With new experiments to be loaded in the core in 2006 the Phenix reactor remains to be a powerful tool providing an important experimental data on fast reactors and on transmutation of minor actinides and long-lived fission products, as well as it will contribute to gain further experience in the framework of the GENERATION IV International Forum. (authors)

  16. Final report of the HFIR [High Flux Isotope Reactor] irradiation facilities improvement project

    International Nuclear Information System (INIS)

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987

  17. Neutron irradiation facilities for fission and fusion reactor materials studies

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

    The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

  18. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

    Luk, K.H.

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR

  19. Accelerated irradiation test of gundremmingen reactor vessel trepan material

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279 degrees C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed

  20. Accelerated irradiation test of Gundremmingen reactor vessel trepan material

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279{degrees}C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed.

  1. Characteristics of irradiation creep in the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available

  2. Thermal analysis of LEU modified Cintichem target irradiated in TRIGA reactor

    International Nuclear Information System (INIS)

    Catana, A; Toma, C.

    2009-01-01

    Actions conceived during last years at international level for conversion of Molybdenum fabrication process from HEU to LEU targets utilization created opportunities for INR to get access to information and participating to international discussions under IAEA auspices. Concrete steps for developing fission Molybdenum technology were facilitated. Institute of Nuclear Research bringing together a number of conditions like suitable irradiation possibilities, direct communication between reactor and hot cell facility, handling capacity of high radioactive sources, and simultaneously the existence of an expanding internal market, decided to undertake the necessary steps in order to produce fission molybdenum. Over the course of last years of efforts in this direction we developed the steps for fission Molybdenum technology development based on modified Cintichem process in accordance with the Argonne National Laboratory proved methodology. Progress made by INR to heat transfer computations of annular target using is presented. An advanced thermal-hydraulic analysis was performed to estimate the heat removal capability for an enriched uranium (LEU) foil annular target irradiated in TRIGA reactor core. As a result, the present analysis provides an upper limit estimate of the LEU-foil and external target surface temperatures during irradiation in TRIGA 14 MW reactor. (authors)

  3. Project accent: graphite irradiated creep in a materials test reactor

    International Nuclear Information System (INIS)

    Brooking, M.

    2014-01-01

    Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)

  4. Present status of irradiation tests on tritium breeding blanket for fusion reactor

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Sagawa, Hisashi; Shimakawa, Satoshi; Tsuchiya, Kunihiko; Kuroda, Toshimasa; Kawamura, Hiroshi.

    1994-01-01

    To develop a tritium breeding blanket for a fusion reactor, irradiation tests in fission reactors are indispensable for obtaining data on irradiation effects on materials, and neutronics/thermal characteristics and tritium production/recovery performance of the blanket. Various irradiation tests have been conducted in the world, especially to investigate tritium release characteristics from tritium breeding and neutron multiplier materials, and materials integrity under irradiation. In Japan, VOM experiments at JRR-2 for ceramic breeders and experiments at JMTR for ceramic breeders and beryllium as a neutron multiplier have been performed. Several universities have also investigated ceramic breeders. In the EC, the EXOTIC experiments at HFR in the Netherlands and the SIBELIUS, the LILA, the LISA and the MOZART experiments for ceramic breeders have carried out. In Canada, NRU has been used for the CRITIC experiments. The TRIO experiments at ORR(ORNL), experiments at RTNS-II, FUBR and ATR have been conducted in the USA. The last two are experiments with high neutron fluence aiming at investigating materials integrity under irradiation. The BEATRIX-I and -II experiments have proceeded under international collaboration of Japan, Canada, the EC and the USA. This report shows the present status of these irradiation tests following a review of the blanket design in the ITER CDA(Conceptual Design Activity). (author)

  5. ''Sleeping reactor'' irradiations: Shutdown reactor determination of short-lived activation products

    International Nuclear Information System (INIS)

    Jerde, E.A.; Glasgow, D.C.

    1998-01-01

    At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux (φ) of ∼ 4 x 10 14 n/cm 2 · s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of ∼ 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about ± 0.5 s) make irradiations of 9 Be(γ,n) 8 Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to ∼ 1 x 10 10 n/cm 2 · s within 1 h. By the time the fuel elements are removed, the flux has dropped to ∼ 6 x 10 8 . Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant

  6. Light water reactor mixed-oxide fuel irradiation experiment

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-01-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

  7. Accelerated Irradiations for High Dose Microstructures in Fast Reactor Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhijie [Univ. of Michigan, Ann Arbor, MI (United States)

    2017-03-31

    The objective of this project is to determine the extent to which high dose rate, self-ion irradiation can be used as an accelerated irradiation tool to understand microstructure evolution at high doses and temperatures relevant to advanced fast reactors. We will accomplish the goal by evaluating phase stability and swelling of F-M alloys relevant to SFR systems at very high dose by combining experiment and modeling in an effort to obtain a quantitative description of the processes at high and low damage rates.

  8. Experience with mechanical segmentation of reactor internals

    International Nuclear Information System (INIS)

    Carlson, R.; Hedin, G.

    2003-01-01

    Operating experience from BWE:s world-wide has shown that many plants experience initial cracking of the reactor internals after approximately 20 to 25 years of service life. This ''mid-life crisis'', considering a plant design life of 40 years, is now being addressed by many utilities. Successful resolution of these issues should give many more years of trouble-free operation. Replacement of reactor internals could be, in many cases, the most favourable option to achieve this. The proactive strategy of many utilities to replace internals in a planned way is a market-driven effort to minimize the overall costs for power generation, including time spent for handling contingencies and unplanned outages. Based on technical analyses, knowledge about component market prices and in-house costs, a cost-effective, optimized strategy for inspection, mitigation and replacements can be implemented. Also decommissioning of nuclear plants has become a reality for many utilities as numerous plants worldwide are closed due to age and/or other reasons. These facts address a need for safe, fast and cost-effective methods for segmentation of internals. Westinghouse has over the last years developed methods for segmentation of internals and has also carried out successful segmentation projects. Our experience from the segmentation business for Nordic BWR:s is that the most important parameters to consider when choosing a method and equipment for a segmentation project are: - Safety, - Cost-effectiveness, - Cleanliness, - Reliability. (orig.)

  9. Experimental fuel channel for samples irradiation at the RB reactor

    International Nuclear Information System (INIS)

    Pesic, M.; Markovic, H.; Sokcic-Kostic, M.; Miric, I.; Prokic, M.; Strugar, P.

    1984-12-01

    An 80% enriched UO 2 fuel channel at the RB nuclear reactor in the 'Boris Kidric' Institute of Nuclear Sciences is modified for samples irradiation by fast neutrons. Maximum sample diameter is 25 mm and length up to 1000 mm. Characteristics of neutron and gamma radiation fields of this new experimental channel are investigated. In the centre of the channel, the main contribution to the total neutron absorbed dose, i.e. 0.29 Gy/Wh of reactor operation, is due to the fast neutron spectrum component. Only 0.05 Gy and 0.07 Gy in the total neutron absorbed dose are due to intermediate and thermal neutrons, respectively. At the same time the gamma absorbed dose is 0.35 Gy. The developed experimental fuel channel, EFC, has wide possibilities for utilization, from fast neutron spectrum studies, electronic component irradiations, dosemeters testing, up to cross-section measurements. (author)

  10. Microstructural evolution in reactor pressure vessel steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

  11. Guidelines for prediction of irradiation embrittlement of operating WWER-440 reactor pressure vessels

    International Nuclear Information System (INIS)

    2005-06-01

    This TECDOC has been developed under an International Atomic Energy Agency Coordinated Research Project (CRP) entitled Evaluation of Radiation Damage of WWER Reactor Pressure Vessels (RPV) using Database on RPV Materials to develop the guidelines for prediction of radiation damage to WWER-440 PRVs. The WWER-440 RPV was designed by OKB Gidropress, Russian Federation, the general designer. Prediction of irradiation embrittlement of RPV materials is usually done in accordance with relevant codes and standards that are based on the large amounts of information from surveillance and research programmes. The existing Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86) for the WWER RPV irradiation embrittlement assessment was approved more than twenty years ago and based mostly on the experimental data obtained in research reactors with accelerated irradiation. Nevertheless, it is still in use and generally consistent with new data. The present publication presents the analyses using all available data required for more precise prediction of radiation embrittlement of WWER-440 RPV materials. Based on the fact that it contains a large amount of data from surveillance programmes as well as research programmes, the IAEA International Database on RPV Materials (IDRPVM) is used for the detailed analysis of irradiation embrittlement of WWER RPV materials. Using IDRPVM, the guideline is developed for assessment of irradiation embrittlement of RPV ferritic materials as a result of degradation during operation. Two approaches, i.e. transition temperatures based on Charpy impact notch toughness, as well as based on static fracture toughness tests, are used in RPV integrity evaluation. The objectives of the TECDOC are the analysis of irradiation embrittlement data for WWER- 440 RPV materials using IDRPVM database, evaluation of predictive formulae depending on chemical composition of the material, neutron fluence, flux, and

  12. The proposed cold neutron irradiation facility at the Breazeale reactor

    International Nuclear Information System (INIS)

    Dimeo, R. M.; Sokol, P. E.; Carpenter, J. M.

    1997-01-01

    We discuss the design considerations of a Cold Neutron Irradiation Facility (CNIF) originally to have been installed at the Penn State Breazeale Reactor (PSBR). The goal of this project was to study the effects of radiation-induced damage to cryogenic moderators and, in particular, solid methane. This work evolved through the design stage undergoing a full safety analysis and received tentative approval from the PSBR Safeguards Committee but was discontinued due to budgetary constraints. (auth)

  13. Miniature tensile test specimens for fusion reactor irradiation studies

    International Nuclear Information System (INIS)

    Klueh, R.L.

    1985-01-01

    Three miniature sheet-type tensile specimens and a miniature rod-type specimen are being used to determine irradiated tensile properties for alloy development for fusion reactors. The tensile properties of type 316 stainless steel were determined with these different specimens, and the results were compared. Reasonably good agreement was observed. However, there were differences that led to recommendations on which specimens are preferred. 4 references, 9 figures, 6 tables

  14. Primary Damage Characteristics in Metals Under Irradiation in the Cores of Thermal and Fast Reactors

    International Nuclear Information System (INIS)

    Pechenkin, V.A.

    2012-01-01

    For an analysis and forecasting of radiation-induced phenomena in structural materials of WWERs, PWRs and BN reactors the fast neutron fluence is usually used (for structural materials of the reactor cores and internals the fluence of neutrons with energy > 0.1 MeV, for WWER and PWRs vessel steels the fluence of neutrons with energy > 0.5 MeV in Russia and East Europe, and with energy > 1.0 MeV in USA and France). Displacements per atom (dpa) seem to be a more appropriate correlation parameter, because it allows comparing the results of materials irradiation in different neutron energy spectra or with different types of particles (neutrons, ions, fast electrons). Energy spectra of primary knocked atoms (PKA) and 'effective' dpa, which are introduced to take into account the point defect recombination during the relaxation stage of a displacement cascade, can be still better representation of the effect of irradiation on material properties. In this work the results of calculating dose rates (dpa/s, NRT-model), PKA energy spectra and PKA mean energies in metals under irradiation in the cores of Russian reactors WWER-440, WWER-1000 (both power thermal reactors) and BN-600 (power fast reactor) and BR-10 (test fast reactor) are presented. In all the reactors Fe and Zr are considered, with addition of Ti and W in BN-600. 'Effective' dose rates in these metals are calculated. Limitations and uncertainties in the standard dpa formulation (the NRT-dpa) are discussed. IPPE activities in the fields related to the TM subject are considered

  15. Estimation of damage by inmates of a PWR Reactor neutron irradiation. Project ZIRP; Estimacion del Dano por Irradiacion Neutronica en los Internos de un Reactor PWR. Proyecto ZIRP

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.

    2013-07-01

    The study presented here focuses on the analysis of neutron and gamma irradiation damage suffered by the inmates of the JC NPP reactor metallic materials throughout its operational life. Such analysis of radiation are part of a project of great international impact, led by EPRI (Electric Power Research Institute) from the MRP (Materials Reliability Program), which aims to relate the degradation of the properties of metallic materials of the inmates of the reactor, with the conditions of operation and irradiation to which have been subjected during the operational life of the plant.

  16. Synthetic study of reactor irradiation for medical use

    International Nuclear Information System (INIS)

    An, Shigehiro; Furuhashi, Akira; Kanda, Keiji; Sumita, Kenji; Kakihana, Hidetake.

    1978-01-01

    This report is described on the results of the study on the reactor irradiation for medical use shared by the Nuclear Engineering Research Laboratory, Faculty of Engineering, University of Tokyo, and other seventeen facilities. Boron neutron-capturing therapy developed in Japan is extremely significant treating method for tumors by destroying tumor cells of encephaloma, etc. selectively. This is the synthetic study for promoting the above therapeutic method. Two existing reactors were reconstructed into the thermal neutron reactors for boron neutron-capturing therapy. The various preparatory and physical researches were made with the reconstruction, and the therapy was tried on eleven cases. Further experiments were made on the following points: (1) To promote treatment on encephaloma by boron neutron-capturing therapy. (2) To develop its application to malignant tumors other than encephaloma. (3) Animal irradiation experiments. (4) The basic experiments on the cellular level. (5) The study of remote controlled anesthesia. (6) To control irradiated dose. (7) To improve boron compounds. (8) To condense radioisotopes. (Kobatake, H.)

  17. Biological samples positioning device for irradiations on a radial channel at the nuclear research reactor

    International Nuclear Information System (INIS)

    Rodriguez Gual, Maritza; Mas Milian, Felix; Deppman, Airton; Pinto Coelho, Paulo Rogerio

    2010-01-01

    For the demand of an experimental device for biological samples positioning system for irradiations on a radial channel at the nuclear research reactor in operation was constructed and started up a device for the place and remove of the biological samples from the irradiation channels without interrupting the operation of the reactor. The economical valuations are effected comparing with another type of device with the same functions. This work formed part of an international project between Cuba and Brazil that undertook the study of the induced damages by various types of ionizing radiation in DNA molecules. Was experimentally tested the proposed solution, which demonstrates the practical validity of the device. As a result of the work, the experimental device for biological samples irradiations are installed and operating in the radial beam hole No3(BH3) for more than five years at the IEA-R1 Brazilian research reactor according to the solicited requirements the device. The designed device increases considerably the type of studies can be conducted in this reactor. Its practical application in research taking place in that facility, in the field of radiobiology and dosimetry, and so on is immediate

  18. Helium production in mixed spectrum reactor-irradiated pure elements

    International Nuclear Information System (INIS)

    Kneff, D.W.; Oliver, B.M.; Skowronski, R.P.

    1986-01-01

    The objectives of this work are to apply helium accumulation neutron dosimetry to the measurement of neutron fluences and energy spectra in mixed-spectrum fission reactors utilized for fusion materials testing, and to measure helium generation rates of materials in these irradiation environments. Helium generation measurements have been made for several Fe, Cu Ti, Nb, Cr, and Pt samples irradiated in the mixed-spectrum High Flux Isotope Reactor (HFIR) and Oak Ridge Research Reactor (ORR) at the Oak Ridge National Laboratory. The results have been used to integrally test the ENDF/B-V Gas Production File, by comparing the measurements with helium generation predictions made by Argonne National Laboratory using ENDF/B-V cross sections and adjusted reactor spectra. The comparisons indicate consistency between the helium measurements and ENDF/B-V for iron, but cross section discrepancies exist for helium production by fast neutrons in Cu, Ti, Nb, and Cr (the latter for ORR). The Fe, Cu, and Ti work updates and extends previous measurements

  19. XAFS study on silica glasses irradiated in a nuclear reactor

    International Nuclear Information System (INIS)

    Yoshida, Tomoko; Yoshida, Hisao; Hara, Takanobu; Ii, Tatsuya; Okada, Tomohisa; Tanabe, Tetsuo

    2000-01-01

    X-ray absorption technique (XANES and EXAFS) was applied to study the local structures of silica glasses before and after the irradiation in a nuclear reactor. Although our separate photoluminescence (PL) measurements clearly showed the different aspects about oxygen vacancies in these samples, i.e., at least the B 2β type oxygen-deficient center exists as an intrinsic defect in the fused silica glass while another type B 2α center is formed in the synthesized silica glass, such differences did not directly reflect on the X-ray absorption spectra (XANES and EXAFS). However, the curve-fitting analysis of EXAFS showed that the number of oxygen atoms coordinated to Si relatively increased after the irradiation. This result may indicate the occurrence of the structural relaxation in the irradiated samples, that is, a slightly distorted SiO 4 tetrahedra in silica glasses relaxed to the regular SiO 4 tetrahedra due to the break of some connections between SiO 4 units in the silica glasses. Thus, the X-ray absorption technique gave the important information of the in-reactor irradiated silica glasses which complements the results obtained from PL measurements

  20. Description of the PIE facility for research reactors irradiated fuels in CNEA

    International Nuclear Information System (INIS)

    Bisca, A.; Coronel, R.; Homberger, V.; Quinteros, A.; Ratner, M.

    2002-01-01

    The PIE Facility (LAPEP), located at the Ezeiza Atomic Center (CAE), was designed to carry out destructive and non-destructive post-irradiation examinations (PIE) on research and power reactor spent fuels, reactor internals and other irradiated materials, and to perform studies related with: Station lifetime extension; Fuel performance; Development of new fuels; and Failures and determination of their causes. LAPEP is a relevant facility where research and development can be carried out. It is worth mentioning that in this facility the PIE corresponding to the Surveillance Program for the Atucha I Nuclear Power Plant (CNA-1) were successfully performed. Materials testing during the CNA-1 repair and the study of failures in fuel element plugs of the Embalse Nuclear Power Plant (CNE) were also performed. (author)

  1. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1983-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

  2. International Fusion Materials Irradiation Facility conceptual design activity. Present status and perspective

    International Nuclear Information System (INIS)

    Kondo, Tatsuo; Noda, Kenji; Oyama, Yukio

    1998-01-01

    For developing the materials for nuclear fusion reactors, it is indispensable to study on the neutron irradiation behavior under fusion reactor conditions, but there is not any high energy neutron irradiation facility that can simulate fusion reactor conditions at present. Therefore, the investigation of the IFMIF was begun jointly by Japan, USA, Europe and Russia following the initiative of IEA. The conceptual design activities were completed in 1997. As to the background and the course, the present status of the research on heavy irradiation and the testing means for fusion materials, the requirement and the technical basis of high energy neutron irradiation, and the international joint design activities are reported. The materials for fusion reactors are exposed to the neutron irradiation with the energy spectra up to 14 MeV. The requirements from the users that the IFMIF should satisfy, the demand of the tests for the materials of prototype and demonstration fusion reactors and the evaluation of the neutron field characteristics of the IFMIF are discussed. As to the conceptual design of the IFMIF, the whole constitution, the operational mode, accelerator system and target system are described. (K.I.)

  3. Reactor Core Internals Replacement of Ikata Units 1 and 2

    International Nuclear Information System (INIS)

    Ikeda, K.; Ishikawa, T.; Miyoshi, T.; Takagi, T.

    2012-01-01

    This paper presents an overview of the reactor core internals replacement project carried out at the Ikata Nuclear Power Station in Japan, which was the first of its kind among PWRs in the world. Failure of baffle former bolts was first reported in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in Belgium and in the U.S., but not in Japan. However, the possibility of these bolts failing in Japanese plants cannot be denied in the future as operating hours increase. Ageing degradation mechanisms for the reactor core internals include irradiation-assisted stress corrosion cracking of baffle former bolts and mechanical wear of control rod guide cards. Two different approaches can be taken to address these ageing issues: to inspect and repair whenever a problem is found; and to replace the entire core internals with those of a new design having advanced features to prevent ageing degradation problems. The choice of our company was the latter. This paper explains the reasons for the choice and summarizes the replacement project activities at Ikata Units 1 and 2 as well as the improvements incorporated in the new design. (author)

  4. Internal corium catcher of a nuclear reactor

    International Nuclear Information System (INIS)

    Anatolii S Vlasov; Vladimir N Mineev; Aleksandr S Sidorov; Yuri A Zeigarnik

    2005-01-01

    Full text of publication follows: A corium catcher is one of the main devices of a nuclear reactor that provides corium melt and fission products retention within a containment during severe accidents. Several studies and design developments have shown that corium retention within a reactor vessel can be attained with a moderate capacity of the latter (up to 600 - 650 MW el.). With a higher reactor capacity external corium catchers are applied both at Russian (VVER-1000) and European (EPR) reactors. In the external catcher of a VVER-1000 reactor, most technological problems are solved due to using sacrificial material. They are as follows: (a) endo-thermal interaction of corium and sacrificial material reduces a level of the temperatures in the final melt pool; (b) solution in the melt of a great amount of the sacrificial material reduces the specific heat release density and the heat flux density at the boundaries of a melt; (c) due to changing of the oxide-component density an inverse stratification of the metallic and oxide components of the corium takes place, thus excluding heat-flux focusing in the zone of the metallic layer and making it possible to supply water on the free surface of the corium without a danger of incipience of the vapor explosion; (d) final oxidation of zirconium occurs without hydrogen generation. The above principles have been realized in the external catcher of the VVER- 1000 reactor at Tyanvan NPS that is presently under construction in China. Successfully solving of the problems concerning to the external catcher makes it possible to return on the new conceptual and technological basis to the idea of retention of the corium melt inside the vessel of a nuclear reactor of large capacity, that is, to provide the reactor vessel to play a role of an internal catcher. For this purpose, a reactor vessel is elongated by approximately two meters. In the lower part of the vessel, on elliptical bottom, pieces of sacrificial material are arranged

  5. 78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

    Science.gov (United States)

    2013-08-19

    ... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Orders; rescission. SUMMARY... the NRC published a final rule, ``Physical Protection of Irradiated Fuel in Transit,'' on May 20, 2013... of Irradiated Reactor Fuel in Transit'' (RIN 3150-AI64; NRC-2009-0163). The final rule incorporates...

  6. Segmentation and packaging reactor vessels internals

    International Nuclear Information System (INIS)

    Boucau, Joseph

    2014-01-01

    Document available in abstract form only, full text follows: With more than 25 years of experience in the development of reactor vessel internals and reactor vessel segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since disposal cost is a key factor, it is important to plan and optimize waste segmentation and packaging. The choice of the optimum cutting technology is also important for a successful project implementation and depends on some specific constraints. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. The usual method is to start at the end of the process, by evaluating handling of the containers, the waste disposal requirements, what type and size of containers are available for the different disposal options, and working backwards to select a cutting method and finally the cut geometry required. The 3-D models can include intelligent data such as weight, center of gravity, curie content, etc, for each segmented piece, which is very useful when comparing various cutting, handling and packaging options. The detailed 3-D analyses and thorough characterization assessment can draw the attention to material potentially subject to clearance, either directly or after certain period of decay, to allow recycling and further disposal cost reduction. Westinghouse has developed a variety of special cutting and handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals

  7. Calculation of DPA in the Reactor Internal Structural Materials of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kim, Yong Deong; Lee, Hwan Soo

    2014-01-01

    The embrittlement is mainly caused by atomic displacement damage due to irradiations with neutrons, especially fast neutrons. The integrity of the reactor internal structural materials has to be ensured over the reactor life time, threatened by the irradiation induced displacement damage. Accurate modeling and prediction of the displacement damage is a first step to evaluate the integrity of the reactor internal structural materials. Traditional approaches for analyzing the displacement damage of the materials have relied on tradition model, developed initially for simple metals, Kinchin and Pease (K-P), and the standard formulation of it by Norgett et al. , often referred to as the 'NRT' model. An alternative and complementary strategy for calculating the displacement damage is to use MCNP code. MCNP uses detailed physics and continuous-energy cross-section data in its simulations. In this paper, we have performed the evaluation of the displacement damage of the reactor internal structural materials in Kori NPP unit 1 using detailed Monte Carlo modeling and compared with predictions results of displacement damage using the classical NRT model. The evaluation of the displacement damage of the reactor internal structural materials in Kori NPP unit 1 using detailed Monte Carlo modeling has been performed. The maximum value of the DPA rate was occurred at the baffle side of the reactor internal where the node has the maximum neutron flux

  8. International standards, Agreements and Policy of food Irradiation

    International Nuclear Information System (INIS)

    Roberts, P.B.

    1997-01-01

    There are few internationally recognised standards and agreements related to irradiated foods. Codex Alimentarius has its General standard for Irradiated foods. This sets standards for the production of irradiated foods that are safe and nutritionally adequate. Guidelines for the proper processing of foods by irradiation are covered in the Codex Recommended International Code of Practice for the Operation of Radiation Facilities Used for the Treatment of Food. For irradiation as a quarantine treatment for fruit, vegetables and other plants, the relevant international organization is the International Plant Protection Convention (IPPC), IPPC has no standards or guidelines for irradiation treatments. However, regional organizations within IPPC are moving towards recognition of irradiation as a technically viable and effective method of insect disinfestation. Especially notable are actions within the North American Plant Protection Organisation (NAPPO). NAPPO has endorsed a standard on the use of irradiation as a quarantine treatment. Other speakers have provided considerable detail on the Codex standard and on the situation with regard to quarantine issues. In this talk I will concentrate on irradiated foods as commodities that will be traded internationally in increasing amounts as we approach the next century. International trade is governed by bilateral arrangements. However, these arrangements should be consistent with the overarching multilateral agreements of the World trade Organization (WTO). The WTO Agreements do not refer directly to irradiation or irradiated foods. However, in this talk I will try to interpret the implications of the Agreements for trade in irradiated food. (Author)

  9. International standards, Agreements and Policy of food Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, P.B. [Industrial and Biological Section. Institute of Geological and Nuclear Science. P.O. Box 31. Lower Hutt (New Zealand)

    1997-12-31

    There are few internationally recognised standards and agreements related to irradiated foods. Codex Alimentarius has its General standard for Irradiated foods. This sets standards for the production of irradiated foods that are safe and nutritionally adequate. Guidelines for the proper processing of foods by irradiation are covered in the Codex Recommended International Code of Practice for the Operation of Radiation Facilities Used for the Treatment of Food. For irradiation as a quarantine treatment for fruit, vegetables and other plants, the relevant international organization is the International Plant Protection Convention (IPPC), IPPC has no standards or guidelines for irradiation treatments. However, regional organizations within IPPC are moving towards recognition of irradiation as a technically viable and effective method of insect disinfestation. Especially notable are actions within the North American Plant Protection Organisation (NAPPO). NAPPO has endorsed a standard on the use of irradiation as a quarantine treatment. Other speakers have provided considerable detail on the Codex standard and on the situation with regard to quarantine issues. In this talk I will concentrate on irradiated foods as commodities that will be traded internationally in increasing amounts as we approach the next century. International trade is governed by bilateral arrangements. However, these arrangements should be consistent with the overarching multilateral agreements of the World trade Organization (WTO). The WTO Agreements do not refer directly to irradiation or irradiated foods. However, in this talk I will try to interpret the implications of the Agreements for trade in irradiated food. (Author)

  10. Design, Manufacturing and Irradiation Behaviour of Fast Reactor Fuel. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2013-04-01

    Fast reactors are vital for ensuring the sustainability of nuclear energy in the long term. They offer vastly more efficient use of uranium resources and the ability to burn actinides, which are otherwise the long-lived component of high level nuclear waste. These reactors require development, qualification, testing and deployment of improved and innovative nuclear fuel and structural materials having very high radiation resistance, corrosion/erosion and other key operational properties. Several IAEA Member States have made efforts to advance the design and manufacture of technologies of fast reactor fuels, as well as to investigate their irradiation behaviour. Due to the acute shortage of fast neutron testing and post-irradiation examination facilities and the insufficient understanding of high dose radiation effects, there is a need for international exchange of knowledge and experience, generation of currently missing basic data, identification of relevant mechanisms of materials degradation and development of appropriate models. Considering the important role of nuclear fuels in fast reactor operation, the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) proposed a Technical Meeting (TM) on 'Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels', which was hosted by the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russian Federation, from 30 May to 3 June 2011. The TM included a technical visit to the fuel production plant MSZ in Elektrostal. The purpose of the meeting was to provide a forum to share knowledge, practical experience and information on the improvement and innovation of fuels for fast reactors through scientific presentations and brainstorming discussions. The meeting brought together 34 specialists from national nuclear agencies, R and D and design institutes, fuel vendors and utilities from 10 countries. The presentations were structured into four sections: R and D Programmes on FR Fuel

  11. Ageing Management of the reactor internals in Belgian nuclear units in view of Long Term Operation

    International Nuclear Information System (INIS)

    Gerard, R.; Bertolis, D.; Vissers, S.

    2012-01-01

    The reactor internals support the reactor core, distribute the coolant flow through the core, and guide and protect the rod control cluster assemblies and in-core instrumentation. Their integrity must be guaranteed in all operating and accident conditions. They are exposed to specific degradation mechanisms linked to the intense neutron irradiation, like Irradiation Assisted Stress Corrosion Cracking (IASCC) or potentially void swelling, in addition to more classical mechanisms like fatigue, wear and stress corrosion cracking. A rigorous follow-up of in-service degradation and an effective ageing management is therefore of crucial importance and contributes to the safe and economical operation of nuclear PWR units. (author)

  12. Irradiation creep experiments on fusion reactor candidate structural materials

    International Nuclear Information System (INIS)

    Hausen, H.; Cundy, M.R.; Schuele, W.

    1991-01-01

    Irradiation creep rates were determined for annealed and cold-worked AMCR- and 316-type steel alloys in the high flux reactor at Petten, for various irradiation temperatures, stresses and for neutron doses up to 4 dpa. Primary creep elongations were found in all annealed materials. A negative creep elongation was found in cold-worked materials for stresses equal to or below about 100 MPa. An increase of the negative creep elongation is found for decreasing irradiation temperatures and decreasing applied stresses. The stress exponent of the irradiation creep rate in annealed and cold-worked AMCR alloys is n = 1.85 and n = 1.1, respectively. The creep rates of cold-worked AMCR alloys are almost temperature independent over the range investigated (573-693 K). The results obtained in the HFR at Petten are compared with those obtained in ORR and EBR II. The smallest creep rates are found for cold-worked materials of AMCR- and US-PCA-type at Petten which are about a factor two smaller than the creep rates obtained of US-316 at Petten or for US-PCA at ORR or for 316L at EBR II. The scatter band factor for US-PCA, 316L, US-316 irradiated in ORR and EBR II is about 1.5 after a temperature and damage rate normalization

  13. Regulatory Assessment Technologies for Aging of Reactor Vessel Internals

    Energy Technology Data Exchange (ETDEWEB)

    Jhung, Myung Jo; Park, Jeong Soon; Ko, Hanok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In order to develop the audit calculation system, it is required to develop crack evaluation, seismic analysis and thermal-hydraulic analysis techniques for RVIs so that integrity of RVIs under the aging environment can be evaluated and be assured. In addition, regulatory requirements including safety review and inspection guides should be developed in order to assure the quality and uniformity of safety reviews and inspections regarding aging assessment and management of RVIs. Since Reactor Vessel Internals (RVIs) are installed within the reactor pressure vessel and surround the fuel assemblies, some of them are exposed to the environment such as high neutron irradiation, high temperature and reactor coolant flow. Those environmental factors can cause damage to RVIs including cracks, loss of material, fatigue, loss of fracture toughness and change of dimension as the operation time of nuclear power plants (NPPs) increases. For long-term operation more than 40 years, aging management of RVIs is important. The final objectives of this study are to establish the audit calculation system for RVIs and to develop regulatory requirements for aging assessment and management of RVIs considering their operating conditions, materials, and possible aging mechanisms.

  14. An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

    NARCIS (Netherlands)

    Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

    2009-01-01

    In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

  15. CR-39 as induced track detector in reactor: irradiation effect

    International Nuclear Information System (INIS)

    Zylberberg, H.

    1989-07-01

    A systematic study about reactor's neutrons radiation effect and gamma radiation effect on the properties of CR-39 that are significant for its use as induced fission track detector is showed. The following studies deserved attention: kinetics of the fission track chemical development; efficiency to register and to develop fission track; losses of developable tracks; variation in the number of developable tracks and variation in the visible and ultraviolet radiation spectrum. The dissertation is organized in seven specific chapters: solid state nuclear tracks (SSNT); CR-39 as SSNT; objectives and problems presentation; preparation and characterization of CR-39 as SSNT; gamma irradiation effect on the properties of CR-39 as SSNT; reactor neutron irradiation effect on the properties of CR-39 as SSNT and, results discussions and conclusions. The main work contributions are the use of CR-39 in the determination of fissionable nuclide as thorium and uranium in solid and liquid samples; gamma radiation damage on CR-39 as well as the reactor's neutron damage on CR-39. (B.C.A.) 62 refs, 53 figs, 21 tabs

  16. Irradiation experience with HTGR fuels in the Peach Bottom Reactor

    International Nuclear Information System (INIS)

    Scheffel, W.J.; Scott, C.B.

    1974-01-01

    Fuel performance in the Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR) is reviewed, including (1) the driver elements in the second core and (2) the test elements designed to test fuel for larger HTGR plants. Core 2 of this reactor, which is operated by the Philadelphia Electric Company, performed reliably with an average nuclear steam supply availability of 85 percent since its startup in July 1970. Core 2 had accumulated a total of 897.5 equivalent full power days (EFPD), almost exactly its design life-time of 900 EFPD, when the plant was shut down permanently on October 31, 1974. Gaseous fission product release and the activity of the main circulating loop remained significantly below the limits allowed by the technical specifications and the levels observed during operation of Core 1. The low circulating activity and postirradiation examination of driver fuel elements have demonstrated the improved irradiation stability of the coated fuel particles in Core 2. Irradiation data obtained from these tests substantiate the performance predictions based on accelerated tests and complement the fuel design effort by providing irradiation data in the low neutron fluence region

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-01

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

  18. A review of formulas for predicting irradiation embrittlement of reactors vessel materials

    International Nuclear Information System (INIS)

    Petrequin, P.

    1995-01-01

    Formulas developed in different countries for predicting irradiation embrittlement of reactors vessel materials are presented. Results of predictions were compared with different data sets, from surveillance programmes or studies in test reactors, with different residual elements contents. Figs

  19. Modified fuel channel for sample irradiation at the RB reactor

    International Nuclear Information System (INIS)

    Pesic, M.; Markovic, H.; Sokcic, M.; Miric, I.; Prokic, M.; Strugar, P.

    1983-01-01

    Fuel channel of 80% enriched UO 2 at RB reactor in Boris Kidric Institute of nuclear sciences is modified for sample irradiation in the fast neutron field. Maximum sample diameter is 25 mm and length up to 100 mm. Characteristics of neutron as well as gamma radiation fields of this new experimental channel are investigated. In the center of channel, the main contribution to the total neutron absorbed dose i.e. 0.29 Gy per 1 Wh of reactor operation, is due to the fast neutron spectrum component. Only 0.05 Gy and 0.07 Gy in the total neutron absorbed dose are due to epithermal and thermal neutrons respectively. At the same time gamma absorption dose is 0.35 Gy. The development of experimental fuel channel GRK has wide possibility for utilization, from electronic components fast neutron studies, dosimeters testing, to cross section measurements for fast neutron energies. (author)

  20. Examination in hot laboratories of irradiated fuels from fast reactors

    International Nuclear Information System (INIS)

    Clottes, G.; Peray, R.; Ratier, J.L.

    1980-05-01

    Low irradiation rate examinations were carried out soon after the Rapsodie, Rapsodie Fortissimo and Phenix reactors were started up for the first time in order to check the level of maximum temperatures reached and the radial migration of oxygen and plutonium and to assess the movements of fuels inside the cladding. The other examinations were effected at a high specific burnup in order to defines the limit specific burnup securing the integrity of the fuel pin claddings (distortion, ruptures and possible consequences). The examinations carried out so far on fuel elements coming from Phenix or Rapsodie have allowed good fuel surveillance to be undertaken and the acquisition of a large number of data, thanks to which the fuel characteristics of future reactors of the system have been developed [fr

  1. Irradiation effects on the ductility of fusion reactor structural materials

    International Nuclear Information System (INIS)

    Boudamous, F.

    1986-10-01

    Austenitic and ferritic-martensitic stainless steels have been proposed as first wall structural materials for the next generation of fusion devices. In order to study the effect of high temperature irradiation on their tensile properties, specimens of the steel AISI 316 L (CEC reference), of the martensitic steel W. Nr 1.4914 and of the duplex ferritic-martensitic steel EM12 have been irradiated in the BR2 reactor in Mol. The austenitic steel was irradiated at 470 0 C to about 1.1 10 22 n/cm 2 ( E>0.1 MeV) while the ferritic-martensitic steels were irradiated at 590 0 C to about 7.7 10 22 n/cm 2 (E>0.1 MeV). The tensile tests of the 316 L steel have been performed between 250 and 750 0 C. Below around 550 0 C, the yield stress after irradiation increased from about 160 to 270 MPa and the total elongation decreased from 42 to about 26%. At 750 0 C, the yield stress increase was small but the total elongation decreased from 60 to only 10%. At this temperature, the rupture of the irradiated specimen was intergranular while all the other specimens presented a transgranular rupture. At 650 0 C the variations were intermediate. The change of the ultimate tensile strength was small at all test temperatures. The EM12 and W. Nr 1.4914 steels tested only at 550 0 C, showed a decrease of the yield and tensile strength as well as an increase of the total elongation. The same tests performed on specimens which have been heat treated in parallel showed that the observed changes were due, in a large part, if not completely, to the maintenance of steels at high temperature

  2. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1993-03-01

    This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the Nuclear Regulatory Commission (NRC), and to meet the requirements of Public Law 96--295. The report provides a brief description of NRC authority for certain aspects of transporting spent fuel. It provides descriptive statistics on spent fuel shipments regulated by the NRC from 1979 to 1992. It also lists detailed highway and railway segments used within each state from October 1, 1987 through December 31, 1992

  3. Alternative methodology for irradiation reactor experimental shielding calculation

    International Nuclear Information System (INIS)

    Vellozo, Sergio de Oliveira; Vital, Helio de Carvalho

    1996-01-01

    Due to a change in the project of the Experimental Irradiation Reactor, its shielding design had to be recalculated according to an alternative simplified analytical approach, since the standard transport calculations were temporarily unavailable. In the calculation of the new width for the shielding made up of steel and high-density concrete layers, the following radiation components were considered: fast neutrons and primary gammas (produced by fission and beta decay), from the core; and secondary gammas, produced by thermal neutron capture in the shielding. (author)

  4. Degradation of austenitic stainless steel (SS) light water ractor (LWR) core internals due to neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Appajosula S., E-mail: Appajosula.Rao@nrc.gov

    2014-04-01

    Austenitic stainless steels (SSs) are extensively being used in the fabrication of light water reactor (LWR) core internal components. It is because these steels have relatively high ductility, fracture toughness and moderate strength. However, the LWR internal components exposure to neutron irradiation over an extended period of plant operation degrades the materials mechanical properties such as the fracture toughness. This paper summarizes some of the results of the existing open literature data on irradiation assisted stress corrosion cracking (IASCC) of 316 CW steels that have been published by the United States Nuclear Regulatory Commission (USNRC), industry, academia, and other research agencies.

  5. Design of unique pins for irradiation of higher actinides in a fast reactor

    International Nuclear Information System (INIS)

    Basmajian, J.A.; Birney, K.R.; Weber, E.T.; Adair, H.L.; Quinby, T.C.; Raman, S.; Butler, J.K.; Bateman, B.C.; Swanson, K.M.

    1982-03-01

    The actinides produced by transmutation reactions in nuclear reactor fuels are a significant factor in nuclear fuel burnup, transportation and reprocessing. Irradiation testing is a primary source of data of this type. A segmented pin design was developed which provides for incorporation of multiple specimens of actinide oxides for irradiation in the UK's Prototype Fast Reactor (PFR) at Dounreay Scotland. Results from irradiation of these pins will extend the basic neutronic and material irradiation behavior data for key actinide isotopes

  6. International Reactor Innovative and Secure (IRIS) summary

    International Nuclear Information System (INIS)

    Carelli, Mario D.

    2001-01-01

    The IRIS (International Reactor Innovative and Secure) reactor is described in the first part of the presentation. IRIS is a light water cooled reactor with an integral configuration, where steam generators, pumps and pressurizer are inside the reactor vessel. Partially funded by the DOE NERI program, IRIS is being developed by an international consortium of 16 organizations from seven countries. A key IRIS characteristic is its 'safety by design' approach which strives to eliminate, by design, as many accidents as possible rather than coping with their consequences. Initial returns are very positive; out of the eight Class IV accidents considered in the AP600 only one remains as a Class IV in IRIS, and at much reduced probability. Small-to-medium LOCAs have minimal consequences as the core remains safely under water for days, without the need for safety injection or water makeup. In spite of its novelty IRIS is firmly grounded on proven LWR technology and therefore a prototype is not needed to assure design certification. Rather, very extensive scaled tests will be performed to investigate the performance of in-vessel components such as steam generators and pumps, both individually and as interactive systems. Accident sequences will also be simulated and tested to prove IRIS safety by design claims. The first core fuel is less than 5% enriched and the fuel assembly is very similar to existing PWR assemblies, so there is no licensing challenge regarding the fuel. Because of the safety by design approach, yielding simplifications In design and accident management (e.g., IRIS does not have an emergency core cooling system), some accident scenarios are eliminated and others have lesser consequences. Thus, simplification and streamlining of the regulatory process might be possible. Risk informed regulation will be coupled with safety by design to show lower accident and damage probabilities. This could lead to a relaxation of siting regulatory requirements. It is

  7. Method for temporary shielding of reactor vessel internals

    International Nuclear Information System (INIS)

    Grimm, N.P.; Sejvar, J.

    1991-01-01

    This patent describes a method for shielding stored internals for reactor vessel annealing. It comprises removing nuclear fuel from the reactor vessel containment building; removing and storing upper and lower core internals under water in a refueling canal storage area; assembling a support structure in the refueling canal between the reactor vessel and the stored internals; introducing vertical shielding tanks individually through a hatch in the containment building and positioning each into the support structure; introducing horizontal shielding tanks individually through a hatch in the containment building and positioning each above the stored internals and vertical tanks; draining water from the refueling canal to the level of a flange of the reactor vessel; placing an annealing apparatus in the reactor vessel; pumping the remaining water from the reactor vessel; and annealing the reactor vessel

  8. Thermal-hydraulic safety aspects related to irradiation capabilities in MTR reactors

    International Nuclear Information System (INIS)

    Khedr, A.

    2009-01-01

    MTR research reactor such as ETRR-2 is an open pool type reactor that has a capability for irradiation into a number of irradiation boxes (IBs) installed at different positions on a separate grid called irradiation grid (I G). The I B has a lower removable plug to open or close its lower nozzle according to the I B is used or not.Increasing the used No. of I Bs in irradiation means that a valuable change in the flow distribution on the I G will occur. This paper is focused on the optimum number of I Bs that could be used without deterioration the cooling of I G components and avoiding the formation of hot spots. RELAP5 system code is used for thermal hydraulic analysis of the I G cooling system. Mathematical models and fortran program is developed to calculate the heat distribution in the I G components and the equivalent nozzle diameter that compensate the I B pressure drop due to the irradiated material (I M). This equivalent diameter simulates the used I B nozzle in the RELAP5 input deck. The results show that, the internal flow into the I Bs has significant effect on the coolability of the I G components. The number of I Bs that can be used is inversely proportional with the reactor power, the IM's void fraction and directly proportional with the PCS flow rate. Different cases of operating power and void fraction at two values for PCS flow are studied. In all of the cases considered limited number of the I Bs is permissible to use in order to avoid the excessive heating of the I G components

  9. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Lee, Jae Han

    2007-02-01

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

  10. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  11. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    International Nuclear Information System (INIS)

    Edwards, G.W.R.; Priest, N.D.; Richardson, R.B.

    2013-01-01

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  12. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, G.W.R.; Priest, N.D.; Richardson, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  13. Results of assembly test of HTTR reactor internals

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

  14. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

    2010-02-16

    Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

  15. Legal aspects and international implications of food irradiation

    International Nuclear Information System (INIS)

    Gerard, Alain.

    1977-11-01

    This paper reports on the status of work on food irradiation at international level, namely the IAEA/FAO/WHO Vienna recommendations, the proposed EEC directive, and the Codex alimentarius draft standards. It then deals with the legal aspects of the subject, in particular the problems concerning definitions, controls and instructions, and finally reviews the regulations for international trade in irradiated foodstuffs. (NEA) [fr

  16. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1981-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and USSR. The Zero-Phase of the INTOR Workshop, which was conducted during 1979, assessed the technical data base that would support the construction of the next major device in the tokamak program to operate in the early 1990s and defined the objectives and characteristics of this device. The INTOR workshop was extended into phase-1, the Definition Phase, in early 1980. The objective of the Phase-1 Workshop was to develop a conceptual design of the INTOR experiment. The purpose of this paper is to give an overview of the work of the Phase-1 INTOR Workshop (January 1980-June 1981, with emphasis upon the conceptual design

  17. Fabrication of Fast Reactor Fuel Pins for Test Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Karsten, G. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Dippel, T. [Institute for Radiochemistry, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Laue, H. J. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

    1967-09-15

    An extended irradiation programme is being carried out for the fuel element development of the Karlsruhe fast breeder project. A very important task within the programme is the testing of plutonium-containing fuel pins in a fast-reactor environment. This paper deals with fabrication of such pins by our laboratories at Karlsruhe. For the fast reactor test positions at present envisaged a fuel with 15% plutonium and the uranium fully enriched is appropriate. Hie mixed oxide is both pelletized and vibro-compacted with smeared densities between 80 and 88% theoretical. The pin design is, for example, such that there are two gas plena at the top and bottom, and one blanket above the fuel with the fuel zone fitting to the test reactor core length. The specifications both for fuel and cladding have been adapted to the special purpose of a fast-breeder reactor - the outer dimensions, the choice of cladding and fuel types, the data used and the kind of tests outline the targets of the development. The fuel fabrication is described in detail, and also the powder line used for vibro-compaction. The source materials for the fuel are oxalate PuO{sub 2} and UO{sub 2} from the UF{sub 6} process. The special problems of mechanical mixing and of plutonium homogeneity have been studied. The development of the sintering technique and grain characteristics for vibratory compactive fuel had to overcome serious problems in order to reach 82-83% theoretical. The performance of the pin fabrication needed a major effort in welding, manufacturing of fits and decontamination of the pin surfaces. This was a stimulation for the development of some very subtle control techniques, for example taking clear X-ray photographs and the tube testing. In general the selection of tests was a special task of the production routine. In conclusion the fabrication of the pins resulted in valuable experiences for the further development of fast reactor fuel elements. (author)

  18. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  19. Irradiation Experiments on Plutonium Fuels for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Frost, B. R.T.; Wait, E. [Atomic Energy Research Establishment Harwell, Berks. (United Kingdom)

    1967-09-15

    An assessment carried out some years ago indicated that cermet fuels might provide the high burn-up and integrity required for fast reactors. An irradiation programme was started at Harwell on (U, Pu)O{sub 2} -SS cermet plates and rods, mainly In thermal neutron fluxes, to gain experience of dimensional stability at temperatures typical of modern sodium-cooled fast reactor designs (600-650 Degree-Sign C). A subsequent assessment showed that cermets carried a large penalty as far as breeding was concerned and (U, Pu)C was chosen by Harwell for long-term study as an alternative, economic, fast reactor fuel. However, the results from the cermet experiments were of sufficient promise to proceed with parallel irradiation programmes on cermets and carbide. The studies of cermets showed that dimensional instability (swelling and cladding rupture) were caused by the pressures exerted on the steel matrix by the fuel particles, and that the initial density of the fuel particles was important in determining the burn-up at which failure occurred. Further, it was shown that cermets provided a useful vehicle for studying the changes occurring in oxide fuel particles with increasing burn-up. The disappearance of initial porosity and its replacement by fission gas bubbles and segregated solid fission products was studied in some detaiL No significant differences were observed between UO{sub 2} and(U,Pu)O{sub 2} particles. The initial studies of (U, Pu)C were concerned with the effect of varying composition and structure on swelling and fission gas release. A tantalum-lined nickel alloy cladding material was used to contain both pellet and powder specimens In an irradiation experiment in the core of the Dounreay fast reactor. This showed that the presence of a metal phase in the fuel led to a high swelling rate, that fission gas release was low up to {approx} 3% bum-up, and that a low density powder accommodated the swelling without excessive straining of the can. A subsequent

  20. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  1. Neutronics analysis of International Fusion Material Irradiation Facility (IFMIF). Japanese contributions

    International Nuclear Information System (INIS)

    Oyama, Yukio; Noda, Kenji; Kosako, Kazuaki.

    1997-10-01

    In fusion reactor development for demonstration reactor, i.e., DEMO, materials tolerable for D-T neutron irradiation are absolutely required for both mechanical and safety point of views. For this requirement, several kinds of low activation materials were proposed. However, experimental data by actual D-T fusion neutron irradiation have not existed so far because of lack of fusion neutron irradiation facility, except fundamental radiation damage studies at very low neutron fluence. Therefore such a facility has been strongly requested. According to agreement of need for such a facility among the international parties, a conceptual design activity (CDA) of International Fusion Material Irradiation Facility (IFMIF) has been carried out under the frame work of the IEA-Implementing Agreement. In the activity, a neutronics analysis on irradiation field optimization in the IFMIF test cell was performed in three parties, Japan, US and EU. As the Japanese contribution, the present paper describes a neutron source term as well as incident deuteron beam angle optimization of two beam geometry, beam shape (foot print) optimization, and dpa, gas production and heating estimation inside various material loading Module, including a sensitivity analysis of source term uncertainty to the estimated irradiation parameters. (author)

  2. An analysis of cobalt irradiation in CANDU 6 reactor core

    International Nuclear Information System (INIS)

    Gugiu, E.D.; Dumitrache, I.

    2003-01-01

    In CANDU reactors, one has the ability to replace the stainless steel adjuster rods with neutronically equivalent Co assemblies with a minimum impact on the power plant safety and efficiency. The 60 Co produced by 59 Co irradiation is used extensively in medicine and industry. The paper mainly describes some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronically equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and the heating of the irradiated cobalt rods are performed using the Monte Carlo codes MCNP5 and MONTEBURNS2.1. The 60 Co activity and heating evaluations are closely related to the neutronics computations and to the density evolution of cobalt isotopes during assumed in-core irradiation period. Unfortunately, the activities of these isotopes could not be evaluated directly using the burn-up capabilities of the MONTEBURNS code because of the lack of their neutron cross-section from the MCNP5 code library. Additional MCNP5 runs for all the cobalt assemblies have been done in order to compute the flux-spectrum, the 59 Co and the 60 Co radiative capture reaction rates in the adjusters. The 60m Co cross-section was estimated using the flux-spectrum and the ORIGEN2.1 code capabilities THERM and RES. These computational steps allowed the evaluation of the one-group cross-section for the radiative capture reactions of cobalt isotopes. The values obtained replaced the corresponding ones from the ORIGEN library, which have been estimated using the flux-spectrum specific to the fuel. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters. (authors)

  3. Clarification of dissolved irradiated light-water-reactor fuel

    International Nuclear Information System (INIS)

    Rodrigues, G.C.

    1983-02-01

    Bench-scale studies with actual dissolved irradiated light water reactor (LWR) fuels showed that continuous centrifugation is a practical clarification method for reprocessing. Dissolved irradiated LWR fuel was satisfactorily clarified in a bench-scale, continuous-flow bowl centrifuge. The solids separated were successfully reslurried in water. When the reslurried solids were mixed with clarified centrate, the resulting suspension behaved similar to the original dissolver solution during centrifugation. Settling rates for solids in actual irradiated fuel solutions were measured in a bottle centrifuge. The results indicate that dissolver solutions may be clarified under conditions achievable by available plant-scale centrifuge technology. The effective particle diameter of residual solids was calculated to be 0.064 microns for Oconee-1 fuel and 0.138 microns for Dresden-1 fuel. Filtration was shown unsuitable for clarification of LWR fuel solutions. Conventional filtration with filter aid would unacceptably complicate remote canyon operation and maintenance, might introduce dissolved silica from filter aids, and might irreversibly plug the filter with dissolver solids. Inertial filtration exhibited irreversible pluggage with nonradioactive stand-in suspensions under all conditions tested

  4. Wholesomeness studies in the International Food Irradiation Project

    Energy Technology Data Exchange (ETDEWEB)

    Elias, P S [International Food Irradiation Project, Federal Research Centre for Nutrition

    1980-01-01

    Despite more than 25 years of history as an effective food preservation method, food irradiation is still subject to strict legislative control in many countries and scientific investigations are required to provide reassurance as to the safety of irradiated food. The International Food Irradiation Project was set up on October 14, 1970 to facilitate the objective evaluation of the wholesomeness of irradiated foodstuffs. Its major activities are; (1) wholesomeness testing of irradiated foods, (2) research on and investigations into the methodology of wholesomeness testing, (3) dissemination of information, and (4) assisting national and international authorities in their consideration of acceptance of irradiated food. In particular, the project over the past nine years had been devoted to the provision of data to national health authorities and international bodies. Up to now, 23 studies were and are being carried out for the project under contract. Subjects for the studies include wheat, wheat flour, potatoes, fish, rice, mango, spices, dried dates, onions and cocoa beans.

  5. CEA fuel pencil qualification under irradiation: from component conception to fuel assembly irradiation in a power reactor

    International Nuclear Information System (INIS)

    Marin, J.-F.; Pillet, Claude; Francois, Bernard; Morize, Pierre; Petitgrand, Sylvie; Atabek, R.-M.; Houdaille, Brigitte.

    1981-06-01

    Fabrication of fuel pins made of uranium oxide pellets and of a zircaloy 4 cladding is described. Irradiation experiment results are given. Thermomechanical behavior of the fuel pin in a power reactor is examined [fr

  6. Aging Management Strategy and Requirements of Pressurized Water Reactor Internal Components

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jun Seog; Oh, Sung Jin; Won, Se Yol; Jeong, Sun Mi [KHNP, Daejeon (Korea, Republic of)

    2016-05-15

    The demonstration that the effects of degradation in the components of PWR internals are adequately managed is essential for maintaining a healthy fleet and ensuring the continued functionality of the reactor internals. It is also very important to determine when and where irradiation susceptibility may occur for the continued operation. This paper introduces the aging management strategies and requirements for PWR internals components and discusses effects of irradiation aging results from the functionality assessments based on the categorization of internal components. This paper introduces aging management strategies and requirements for PWR internals components. The aging management requirements for PWR internals are specified in four final component groups, which are Primary, Expansion, Existing Program and No Additional Measures. Among these groups, Primary groups include any restriction on general applicability, degradation mechanism, forward link to any Expansion components, examination method, initial examination and frequency, and examination coverage and accessibility. Expansion groups are backward link to the Primary component.

  7. Irradiation-accelerated corrosion of reactor core materials

    International Nuclear Information System (INIS)

    Bartels, David; Was, Gary; Jiao, Zhijie

    2012-09-01

    The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, but also applies to most all other GenIV concepts. Of these four drivers, the combination of radiation and corrosion presents a unique and extremely challenging environment for materials, for which an understanding of the fundamental science is essentially absent. Irradiation can affect corrosion or oxidation in at least three different ways. Radiation interaction with water results in the decomposition of water into radicals and oxidizing species that will increase the electrochemical corrosion potential and lead to greater corrosion rates. Irradiation of the solid surface can produce excited states that can alter corrosion, such as in the case of photo-induced corrosion. Lastly, displacement damage in the solid will result in a high flux of defects to the solid-solution interface that can alter and perhaps, accelerate interface reactions. While there exists reasonable understanding of how corrosion is affected by irradiation of the aqueous environment, there is little understanding of how irradiation affects corrosion through its impact on the solid, whether metal or oxide. The reason is largely due to the difficulty of conducting experiments that can measure this effect separately. We have undertaken a project specifically to separate the several effects of irradiation on the mechanisms of corrosion. We seek to answer the question: How does radiation damage to the solution-oxide couple affect the oxidation process differently from radiation damage to either component alone? The approach taken in this work is to closely compare corrosion accelerated by (1) proton irradiation, (2) electron irradiation, and (3) chemical corrosion potential effects alone, under typical PWR operating conditions at 300 deg. C. Both 316 stainless steel and zirconium are to be studied. The proton

  8. Nuclear reactor internals construction and failed fuel rod detection system

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

  9. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    International Nuclear Information System (INIS)

    Iracane, Daniel; Bignan, Gilles; Lindbaeck, Jan-Erik; Blomgren, Jan

    2010-01-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  10. International feedback experience on the cutting of reactor internal components

    International Nuclear Information System (INIS)

    Boucau, J.

    2014-01-01

    Westinghouse capitalizes more than 30 years of experience in the cutting of internal components of reactor and their packaging in view of their storage. Westinghouse has developed and validated different methods for cutting: plasma torch cutting, high pressure abrasive water jet cutting, electric discharge cutting and mechanical cutting. A long feedback experience has enabled Westinghouse to list the pros and cons of each cutting technology. The plasma torch cutting is fast but rises dosimetry concerns linked to the control of the cuttings and the clarity of water. Abrasive water jet cutting requires the installation of costly safety devices and of an equipment for filtering water but this technology allows accurate cuttings in hard-to-reach zones. Mechanical cutting is the most favourable technology in terms of wastes generation and of the clarity of water but the cutting speed is low. (A.C.)

  11. Overview of the FUTURIX-FTA Irradiation Experiment in the Phénix Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Heather J.M. Chichester; Steve L. Hayes; Kenneth J. McClellan; Jean-Luc Paul; Marc Masson; Stewart L. Voit; Fabienne Delage

    2015-09-01

    The Advanced Fuels Campaign utilizes the Advanced Test Reactor (ATR) for most of its irradiation testing. Cadmium-shrouded baskets are used in ATR to modify the neutron spectrum to simulate a fast reactor environment for the fuel. FUTURIX-FTA is an irradiation experiment conducted in the Phenix fast reactor in France. Results from FUTURIX-FTA and irradiation tests in ATR using identical fuel compositions will be compared to identify and evaluate any differences in fuel behavior due to differences in the irradiation source.

  12. Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

    International Nuclear Information System (INIS)

    Elbel, H.; Kummerer, K.; Bojarsky, K.; Lopez Jimenez, J.; Otero de la Gandara, J.L.

    1979-01-01

    Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

  13. Experimental Irradiations of Materials and Fuels in the BR2 Reactor: An Overview of Current Programmes

    International Nuclear Information System (INIS)

    Van Dyck, S.; Koonen, E.; Verwerft, M.; Wéber, M.

    2013-01-01

    The BR2 material test reactor offers a variety of experimental irradiation possibilities for testing of materials, fuels and instruments. The current paper gives an overview of the recent and ongoing programmes in order to illustrate the experimental potential of the reactor. Three domains of applications are reviewed: Irradiation of materials and fuels for pressurised water reactors (PWR); irradiation of materials for accelerator driven systems (ADS), cooled by liquid lead alloys; and irradiation of fuel for Material Test Reactors (MTR). For PWR relevant tests, a dedicated loop is available, providing a full simulation of the thermo hydraulic conditions of a PWR. ADS related tests require particular control of the irradiation environment and the necessary safety precautions in order to avoid 210 Po contamination. In-core mechanical testing of materials is done in comparison and complimentarily to post-irradiation examinations in order to assess flux related effects on the deformation behaviour of materials. (author)

  14. Post-Irradiation Examination and In-Pile Measurement Techniques for Water Reactor Fuels

    International Nuclear Information System (INIS)

    2009-12-01

    in the 1960s when the construction of NPPs was being started. Evidently it can be assumed that infrastructure with basic unique equipments is old enough, both morally and physically, and needs to be up-graded or replaced. Thus, a sharp increase of the hydrocarbon fuel cost, green-house effect, necessity to construct more safe and efficient NPPs, justification of the lifetime prolongation of the existing NPPs, moral and physical ageing of the hot labs and research reactors equipment lead to the strong necessity to develop more perfect and more precise methods and equipment to examine irradiated components of nuclear reactors, first of all the most expensive one - nuclear fuel. Now the national hot laboratories and material testing reactors usually act as individual independent research establishments without any common and coordinated technical and business strategy towards the future needs and challenges. Even if there are not many joint programs for the development of nuclear power engineering in different countries, the method base and accumulated experience of the in- and post-reactor experiments should be widely shared so as to decrease the cost of this base in each country and to enforce its development. Thus, both problems and results of the application of new techniques to examine nuclear reactor components, as well as the conditions of separate labs should be discussed at the international level. The IAEA technical meetings are one of the most convenient means of arranging such discussion on the problems of the hot labs and research reactors development and application of new original techniques for examination of reactor materials properties. This publication represents a summary and proceedings of the two technical meetings (TMs) organized by IAEA on the subjects of Hot Cell Post-Irradiation Examination (PIE) Techniques and Pool Side Inspection of Water Reactor Fuel Assemblies and Fuel Rod Instrumentation and In-Pile Measurement Techniques. The first TM was

  15. Multi-physic simulations of irradiation experiments in a technological irradiation reactor

    International Nuclear Information System (INIS)

    Bonaccorsi, Th.

    2007-09-01

    A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

  16. Irradiation Performance of HTGR Fuel in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Ueta, Shohei; Sakaba, Nariaki; Shaimerdenov, Asset; Gizatulin, Shamil; Chekushina, Lyudmila; Chakrov, Petr; Honda, Masaki; Takahashi, Masashi; Kitagawa, Kenichi

    2014-01-01

    A capsule irradiation test with the high temperature gas-cooled reactor (HTGR) fuel is being carried out using WWR-K research reactor in the Institute of Nuclear Physics of the Republic of Kazakhstan (INP) to attain 100 GWd/t-U of burnup under normal operating condition of a practical small-sized HTGR. This is the first HTGR fuel irradiation test for INP in Kazakhstan collaborated with Japan Atomic Energy Agency (JAEA) in frame of International Science and Technology Center (ISTC) project. In the test, TRISO coated fuel particle with low-enriched UO_2 (less than 10 % of "2"3"5U) is used, which was newly designed by JAEA to extend burnup up to 100 GWd/t-U comparing with that of the HTTR (33 GWd/t-U). Both TRISO and fuel compact as the irradiation test specimen were fabricated in basis of the HTTR fuel technology by Nuclear Fuel Industries, Ltd. in Japan. A helium-gas-swept capsule and a swept-gas sampling device installed in WWR-K were designed and constructed by INP. The irradiation test has been started in October 2012 and will be completed up to the end of February 2015. The irradiation test is in the progress up to 69 GWd/t of burnup, and integrity of new TRISO fuel has been confirmed. In addition, as predicted by the fuel design, fission gas release was observed due to additional failure of as-fabricated SiC-defective fuel. (author)

  17. Pathfinder irradiation of advanced fuel (Th/U mixed oxide) in a power reactor

    International Nuclear Information System (INIS)

    Brant Pinheiro, R.

    1993-01-01

    Within the joint Brazilian-German cooperative R and D Program on Thorium Utilization in Pressurized Water Reactors carried out from 1979 to 1988 by Nuclebras/CDTN, KFA-Juelich, Siemens/KWU and NUKEM, a pathfinder irradiation of Th/U mixed oxide fuel in the Angra 1 nuclear power reactor was planned. The objectives of this irradiation testing, the irradiation strategy, the work performed and the status achieved at the end of the joint Program are presented. (author)

  18. Preliminary Design Concept for a Reactor-internal CRDM

    International Nuclear Information System (INIS)

    Lee, Jae Seon; Kim, Jong Wook; Kim, Tae Wan; Choi, Suhn; Kim, Keung Koo

    2013-01-01

    A rod ejection accident may cause severer result in SMRs because SMRs have relatively high control rod reactivity worth compared with commercial nuclear reactors. Because this accident would be perfectly excluded by adopting a reactor-internal CRDM (Control Rod Drive Mechanism), many SMRs accept this concept. The first concept was provided by JAERI with the MRX reactor which uses an electric motor with a ball screw driveline. Babcock and Wilcox introduced the concept in an mPower reactor that adopts an electric motor with a roller screw driveline and hydraulic system, and Westinghouse Electric Co. proposes an internal Control Rod Drive in its SMR with an electric motor with a latch mechanism. In addition, several other applications have been reported thus far. The reactor-internal CRDM concept is now widely adopted in many SMR designs, and this concept may also be applied in an evolutionary reactor development. So the preliminary study is conducted based on the SMART CRDM design. A preliminary design concept for a reactor-internal CRDM was proposed and evaluated through an electromagnetic analysis. It was found that there is an optimum design for the motor housing, and the results may contribute to the realization a reactor-internal CRDM for an evolutionary reactor development. More detailed analysis results will be reported later

  19. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1991-01-01

    This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1989 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials. 11 figs., 3 tabs

  20. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1992-06-01

    The circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the US Nuclear Regulatory Commission (NRC). It provides a brief description of spent fuel shipment safety and safeguards requirements of general interest, a summary of data for 1979--1991 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials

  1. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  2. Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Brauer, G.; Matz, W.; Liszkay, L.; Molnar, B.

    1990-11-01

    Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

  3. Irradiation of an uranium silicide prototype in RA-3 reactor

    International Nuclear Information System (INIS)

    Calabrese, R.; Estrik, G.; Notari, C.

    1996-01-01

    The factibility of irradiation of an uranium silicide (U 3 Si 2 ) prototype in the RA-3 reactor was studied. The standard RA-3 fuel element uses U 3 O 8 as fissible material. The enrichment of both standard and prototype is the same: 20% U 235 and also the frame geometry and number of plates is identical. The differences are in the plate dimensions and the fissile content which is higher in the prototype. The cooling conditions of the core allow the insertion of the prototype in any core position, even near the water trap, if the overall power is kept below 5Mw. Nevertheless, the recommendation was to begin irradiation near the periphery and later on move the prototype towards more central positions in order to increase the burnup rate. The prototype was effectively introduced in a peripheral position and the thermal fluxes were measured between plates with the foil activation technique. These were also evaluated with the fuel management codes and a reasonable agreement was found. (author). 5 refs., 3 figs., 3 tabs

  4. International Experience with Fast Reactor Operation & Testing

    International Nuclear Information System (INIS)

    Sackett, John I.; Grandy, C.

    2013-01-01

    Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

  5. Technical Meeting on Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels. Presentations

    International Nuclear Information System (INIS)

    2011-01-01

    The purpose of this meeting was to enable a rationalization and advancement of the design and manufacturing processes, a better selection of promising fuels, and a reduction of the time and costs currently required for R and D and testing, as well as to contribute to the improvement of the safety features of fuels under all operational states and accidental conditions. An overview of the status and perspective of the design, manufacturing and irradiation behaviour of fast reactors fuels were provided during this meeting. The main objectives are the following: Ensure sharing and dissemination of knowledge and expertise; Discuss specific features and issues of existing fuels; Improve knowledge and data for the design and engineering of fast reactor fuel and core structural materials; Discuss perspectives on advanced fuels; Consider modern technological, design and testing tools enabling reliable performance of fuels in current and planned operational environments; Establish international consensus in the developmental efforts on advanced fast reactor technologies, including collaborative programs and experiments. Contribute to the preparation and outline of the planned IAEA Coordinated Research Project on 'Examination of advanced fast reactor fuel and core structural materials. Each of the 24 presentations made at the meeting have been indexed separately

  6. Effect of irradiation in design of LMFBR internals

    International Nuclear Information System (INIS)

    Tavassoli, A.A.; Cowan, A.; Vries, M. de; Heesen, E.

    1990-01-01

    Internal structures of nuclear power reactors are essentially made with the austenitic stainless steels Types 304L and 316L. In service these structures receive low to moderate neutron doses. In this paper, the work undertaken by the European Fast Breeder Working Group is reviewed. Conclusions drawn to this date are presented and tentative reduction factors to be used in design are discussed in terms of the number of displacements per atom (dpa) and the quantity of helium generated in the steel (appm He). For the lower core structure which operates at about 400 0 C existing design rules can be used for parts which are subjected to less than 2 dpa despite a reduction in ductility and toughness which occurs above about 0.8 dpa. For the above core structure which operates at about 550 0 C interim and rather conservative stress reduction factors are proposed which can become effective at helium levels as low as 10 -4 appm. Particular attention is paid to: - irradiation temperature, - neutron flux and fluence, - steel type and grade, - heat treatment and boron distribution, - weld metal composition and procedure (TIG,MMA,...), to ensure that service conditions are represented as closely as possible

  7. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Maeda, Koji [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

    2011-09-30

    Highlights: > We evaluated diametral strain of fast reactor MOX fuel pins irradiated to 130 GWd/t. > The strain was due to cladding void swelling and irradiation creep. > The irradiation creep was caused by internal gas pressure and PCMI. > The PCMI was associated with pellet swelling by rim structure or by cesium uranate. > The latter effect tended to increase the cumulative damage fraction of the cladding. - Abstract: The C3M irradiation test, which was conducted in the experimental fast reactor, 'Joyo', demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, 'Monju'. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and {sup 137}Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  8. Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

    Energy Technology Data Exchange (ETDEWEB)

    Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J B [eds.

    1995-07-01

    Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics.

  9. Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

    International Nuclear Information System (INIS)

    Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J.B.

    1995-01-01

    Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics

  10. Development of neutron irradiation embrittlement correlation of reactor pressure vessel materials of light water reactors

    International Nuclear Information System (INIS)

    Soneda, Naoki; Dohi, Kenji; Nomoto, Akiyoshi; Nishida, Kenji; Ishino, Shiori

    2007-01-01

    A large amount of surveillance data of the RPV embrittlement of the Japanese light water reactors have been compiled since the current Japanese embrittlement correlation has been issued in 1991. Understanding on the mechanisms of the embrittlement has also been greatly improved based on both experimental and theoretical studies. CRIEPI and the Japanese electric power utilities have started research project to develop a new embrittlement correlation method, where extensive study of the microstructural analyses of the surveillance specimens irradiated in the Japanese commercial reactors has been conducted. The new findings obtained from the experimental study are that the formation of solute-atom clusters with little or no copper is responsible for the embrittlement in low-copper materials, and that the flux effect exists especially in high-copper materials and this is supported by the difference in the microstructure of the high-copper materials irradiated at different fluxes. Based on these new findings, a new embrittlement correlation method is formulated using rate equations. The new methods has higher prediction capability than the current Japanese embrittlement correlation in terms of smaller standard deviation as well as smaller mean value of the prediction error. (author)

  11. Study on prediction model of irradiation embrittlement for reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Wang Rongshan; Xu Chaoliang; Huang Ping; Liu Xiangbing; Ren Ai; Chen Jun; Li Chengliang

    2014-01-01

    The study on prediction model of irradiation embrittlement for reactor pres- sure vessel (RPV) steel is an important method for long term operation. According to the deep analysis of the previous prediction models developed worldwide, the drawbacks of these models were given and a new irradiation embrittlement prediction model PMIE-2012 was developed. A corresponding reliability assessment was carried out by irradiation surveillance data. The assessment results show that the PMIE-2012 have a high reliability and accuracy on irradiation embrittlement prediction. (authors)

  12. Irradiation tests of THTR fuel elements in the DRAGON reactor (irradiation experiment DR-K3)

    International Nuclear Information System (INIS)

    Burck, W.; Duwe, R.; Groos, E.; Mueller, H.

    1977-03-01

    Within the scope of the program 'Development of Spherical Fuel Elements for HTR', similar fuel elements (f.e.) have been irradiated in the DRAGON reactor. The f.e. were fabricated by NUKEM and were to be tested under HTR conditions to scrutinize their employability in the THTR. The fuel was in the form of coated particles moulded into A3 matrix. The kernels of the particles were made of mixed oxide of uranium and thorium with an U 235 enrichment of 90%. One aim of the post irradiation examination was the investigation of irradiation induced changes of mechanical properties (dimensional stability and elastic behaviour) and of the corrosion behaviour which were compared with the properties determined with unirradiated f.e. The measurement of the fission gas release in annealing tests and ceramografic examinations exhibited no damage of the coated particles. The measured concentration distribution of fission metals led to conclusions about their release. All results showed, that neither the coated particles nor the integral fuel spheres experienced any significant changes that could impair their utilization in the THTR. (orig./UA) [de

  13. Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

    International Nuclear Information System (INIS)

    1998-10-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

  14. In-cell refabrication of experimental pencils from pencils pre-irradiated in a power reactor

    International Nuclear Information System (INIS)

    Vignesoult, N.; Atabek, R.; Ducas, S.

    1980-05-01

    For the fuel-cladding study, small irradiated pencils were fabricated in a hot cell from long elements taken from power reactors. This reconstitution in a hot cell makes it possible to: - avoid long and costly fabrications of pencils and pre-irradiations in experimental reactors, - perform re-irradiations on very long fuel elements from power reactors, - fabricate several small pencils from one pre-irradiation pencil having homogeneous characteristics. This paper describes (a) the various in-cell fabrication stages of small pre-irradiated pencils, stressing the precautions taken to avoid any pollution and modifications in the characteristics of the pencil, in order to carry out a perfectly representative re-irradiation, (b) the equipment used and the quality control made, and (c) the results achieved and the qualification programme of this operation [fr

  15. Accumulation of the radionuclides in a target irradiated in the reactor of tajoura nuclear research center

    International Nuclear Information System (INIS)

    Abdunnobi, A.R.; Arebi, B.

    1998-01-01

    One of the main stages of radionuclides production in reactor is the distinguishing of a regime on target irradiation in order to acquire the sufficient activity and the purity of radioisotope required. The authors have derived formula for calculating radionuclidic accumulation on a target irradiated in the reactor operating 10 hours per day, 4 days a week during 4 weeks. The results of I-131 and other radionuclide accumulation are illustrated by a tellurium target irradiation in the reactor operating continuously or with interruptions

  16. Evaluation of aluminum capsules according to ISO 9978 to irradiation of gaseous samples in nuclear reactor

    International Nuclear Information System (INIS)

    Costa, Osvaldo L. da.; Tiezzi, Rodrigo; Souza, Daiane C.B.; Feher, Anselmo; Moura, Joao A.; Souza, Carla D.; Moura, Eduardo S.; Oliveira, Henrique B.; Zeituni, Carlos A.; Rostelato, Maria Elisa C.M.

    2015-01-01

    Gas irradiation in research nuclear reactors is an important way to produce radionuclides. Although some nuclear reactors centers offer this type of service, there are few publications about capsules to irradiation of gaseous samples. This paper describes a method to fabricate and evaluate aluminum capsules to irradiate gaseous samples in nuclear reactor. A semi-circular slotted die from a hydraulic press head was modified to seal aluminum tubes. The aluminum capsules were subjected to leak detection tests, which demonstrated the accordance with standard ISO 9978. (author)

  17. Irradiation positions for fission-track dating in the University of Pavia TRIGA Mark II nuclear reactor

    International Nuclear Information System (INIS)

    Oddone, Massimo; Meloni, Sandro; Balestrieri, Maria Laura; Bigazzi, Giulio

    2002-01-01

    An irradiation position arranged is described in the present paper for fission-track dating in the Triga Mark II reactor of the University of Pavia. Fluence values determined using the NIST glass standard SRM 962a for fission-track dating and the traditional metal foils are compared. Relatively good neutron thermalization (φ th /φ f = 0.956) and lack of significant fluence spatial gradients are good factors for fission-track dating. Finally, international age standards (or putative age standards) irradiated in this new position yielded results consistent with independent reference ages. (author)

  18. Design and construction of γ-rays irradiation facility for remote-handling parts and components of fusion reactor

    International Nuclear Information System (INIS)

    Yagi, Toshiaki; Morita, Yousuke; Seguchi, Tadao

    1995-03-01

    For the evaluation of radiation resistance of remote-handling system for International Thermonuclear Experimental Reactor(ITER), 'high dose-rate and high temperature (upper 350degC) γ-rays irradiation facility' was designed and constructed. In this facility, the parts and components of remote-handling system such as sensing devices, motors, optical glasses, wires and cables, etc., are tested by irradiation with 2x10 6 Roentgen/h Co-60 γ-rays at a temperature up to 350degC under various atmospheres (dry nitrogen gas, argon gas, dry air and vacuum). (author)

  19. Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

    2014-01-01

    Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

  20. Development of a Low Temperature Irradiation Capsule for Research Reactor Materials

    International Nuclear Information System (INIS)

    Choo, Kee Nam; Cho, Man Soon; Lee, Cheol Yong; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Kang, Suk Hoon; Kang, Young Hwan; Park, Sang Jun

    2013-01-01

    A new capsule design was prepared and tested at HANARO for a neutron irradiation of core materials of research reactors as a part of the research reactor development project. Irradiation testing of the materials including graphite, beryllium, and zircaloy-4 that are supposed to be used as core materials in research reactors was required for irradiation at up to 8 reactor operation cycles at low temperature (<100 .deg. C). Therefore, three instrumented capsules were designed and fabricated for an evaluation of the neutron irradiation properties of the core materials (Graphite, Be, Zircaloy-4) of research reactors. The capsules were first designed and fabricated to irradiate materials at low temperature (<100 .deg. C) for a long cycle of 8 irradiation cycles at HANARO. Therefore, the safety of the new designed capsule should be fully checked before irradiation testing. Out-pile performance and endurance testing before HANARO irradiation testing was performed using a capsule under a 110% condition of a reactor coolant flow amount. The structural integrity of the capsule was analyzed in terms of a vibration-induced fatigue cracking of a rod tip of the capsule that is suspected to be the most vulnerable part of a capsule. Another two capsules were irradiated at HANARO for 4 cycles, and one capsule was transferred to a hot cell to examine the integrity of the rod tip of the capsule. After confirming the soundness of the 4 cycle-irradiated capsule, the remaining capsule was irradiated at up to 8 cycles at HANARO. Based on the structural integrity analysis of the capsule, an improved capsule design will be suggested for a longer irradiation test at HANARO

  1. Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

    International Nuclear Information System (INIS)

    Porollo, S.I.; Konobeev, Y.V.; Ivanov, A.A.; Shulepin, S.V.; Garner, F.

    2007-01-01

    Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

  2. The post irradiation examination of three fuel rods from the IFA 429 experiment irradiated in the Halden Reactor

    International Nuclear Information System (INIS)

    Williams, J.

    1979-11-01

    A series of fuel rod irradiation experiments were performed in the Halden Heavy Boiling Water Reactor in Norway. These were designed to provide a range of fuel property data as a function of burn-up. One of these experiments was the IFA-429. This was designed to study the absorption of helium filling gas by the UO 2 fuel pellets, steady state and transient fission gas release and fuel thermal behaviour to high burn-up. This data was to be obtained as a function of fuel density, fuel grain size, initial fuel/cladding gap, average linear heat rating, burn-up and overpower transients. All the fuel is in the form of pressed and sintered UO 2 pellets enriched to 13 weight percent 235 U. All the rods were clad in Zircaloy 4 tube. The details of the experiment are given. The post irradiation examination included: visual examination, neutron radiography, dimensional measurements, gamma scanning, measurement of gases in fuel rods and internal free volume, burn-up analysis, metallographic examination, measurement of retained gas in UO 2 pellets, measurement of bulk density of UO 2 . The results are given and discussed. (U.K.)

  3. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    International Nuclear Information System (INIS)

    Clayton, J.C.

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated

  4. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  5. International Conference on Physics and Technology of Reactors and Applications

    International Nuclear Information System (INIS)

    2011-01-01

    Full text : The international conference on physics and technology of reactors is organized by the Moroccan Association for Nuclear enggineering and Reactor Technology (GMTR) with the collaboration of the Centre for Energy and Nuclear Sciences and Techniques (CNESTEN) and under the auspices of the ministry of Energy, Mining, Water and Environment. The programme of the PHYTRA2 conference covers a wide variety of topics. The conference is organised in one plenary session, eight oral technical sessions and one poster session. The oral and poster technical sessions covers the usual topics of nuclear engineering including one session on research reactors utilisation and computational methods for research reactors

  6. Irradiation experiments on materials for core internals, pressure vessel and fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Materials degradation due to the aging phenomena is one of the key issues for the life assessment and extension of the light water reactors (LWRs). This presentation introduces JAERI`s activities in the field of LWR material researches which utilize the research and testing reactors for irradiation experiments. The activities are including the material studies for the core internals, pressure vessel and fuel cladding. These materials are exposed to the neutron/gamma radiation and high temperature water environments so that it is worth reviewing their degradation phenomena as the continuum. Three topics are presented; For the core internal materials, the irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels is the present major concern. At JAERI the effects of alloying elements on IASCC have been investigated through the post-irradiation stress corrosion cracking tests in high-temperature water. The radiation embrittlement of pressure vessel steels is still a significant issue for LWR safety, and at JAERI some factors affecting the embrittlement behavior such as a dose rate have been investigated. Waterside corrosion of Zircaloy fuel cladding is one of the limiting factors in fuel rod performance and an in-situ measurement of the corrosion rate in high-temperature water was performed in JMTR. To improve the reliability of experiments and to extent the applicability of experimental techniques, a mutual utilization of the technical achievements in those irradiation experiments is desired. (author)

  7. Effects of irradiation on ferritic alloys and implications for fusion reactor applications

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1986-07-01

    This paper reviews the ADIP irradiation effects data base on ferritic (martensitic) alloys to provide reactor teams with an understanding of how such alloys will behave for fusion reactor first wall applications. Irradiation affects dimensional stability, strength and toughness. Dimensional stability is altered by precipitation and void swelling. Swelling as high as 25% may occur in some ferritic alloys at 500 dpa. Irradiation alters strength both during and following irradiation. Irradiation at low temperatures leads to hardening whereas at higher temperatures and high exposures, precipitate coarsening can result in softening. Toughness can also be adversely affected by irradiation. Failure can occur in ferritic in a brittle manner and irradiation induced hardening causes brittle failure at higher temperatures. Even at high test temperatures, toughness is reduced due to reduced failure initiation stresses. 39 refs

  8. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M; Boehmert, J; Gilles, R [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  10. Comparison of advanced reactors program of different international vendors

    International Nuclear Information System (INIS)

    Agnihotri, N.K.

    2001-01-01

    The full text follows. Proposal for presenting a paper on Advanced Reactor Program Given below is the abstract for Track 6 session on Advanced Reactor at the ninth International Conference on Nuclear Engineering being held in Nice, France from April 8. through 12. 2001. This paper will provide an update on Advanced Reactor Program of different vendors in the United States, Japan, and Europe. Specifically the paper will look at the history of different Advanced Reactor Programs, international experience, aspect of economy due to standardization, and the highlights of technical specifications. The paper will also review aspects of Economy due to standardization, public acceptance, required construction time, and the experience of different vendors. The objective of the presentation is to underscore the highlights of the Reactor Program of different vendors in order to keep the attendees of the conference up-to-date. The presentation will be an impartial overview from an outsider's (not part of the Nuclear Steam Supply System's staff). (author)

  11. Public information circular for shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    1988-04-01

    This circular has been prepared in response to numerous requests for information regarding routes for the shipment of irradiated reactor (spent) fuel subject to regulation by the Nuclear Regulatory Commission (NRC). The NRC staff approves such routes prior to their use, in accordance with the regulatory provisions of 10 CFR Part 73.37. The objective of the safeguards regulations contained in 10 CFR Part 73.37 is to provide protection against radioactive dispersal caused by malevolent acts by persons. The design and construction of the casks used to ship the spent fuel provide adequate radiological protection of the public health and safety against accidents. Therfore, transporting appropriately packaged spent fuel over existing rail systems and via any highway system is radiologically safe without specific NRC approval of the route. However, to assure adequate planning for protection against actual or attempted acts of radiological sabotage, the NRC requires advance route approval. This approval is given on a shipment-by-shipment or series basis, it is not general approval of the route for subsequent spent fuel shipments. Spent fuel shipment routes, primarily for road transportation, but also including three rail routes, are indicated on reproductions of road maps. Also included are the amounts of material shipped during the approximate 8-year period that safeguards regulations have been effective. This information is current as of September 30, 1987

  12. Delivery of completed irradiation vehicles and the quality assurance document to the High Flux Isotope Reactor for irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, Christian M. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); McDuffee, Joel Lee [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    This report details the initial fabrication and delivery of two Fuel Cycle Research and Development (FCRD) irradiation capsules (ATFSC01 and ATFSC02), with associated quality assurance documentation, to the High Flux Isotope Reactor (HFIR). The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0202268. These irradiation experiments are testing silicon carbide composite tubes in order to obtain experimental validation of thermo-mechanical models of stress states in SiC cladding irradiated under a prototypic high heat flux. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulics and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division (RNSD). A complete fabrication package record is maintained by the THIEG and is available upon request.

  13. Behavior of Type 316 stainless steel under simulated fusion reactor irradiation

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Maziasz, P.J.; Bloom, E.E.; Stiegler, J.O.; Grossbeck, M.L.

    1978-05-01

    Fusion reactor irradiation response in alloys containing nickel can be simulated in thermal-spectrum fission reactors, where displacement damage is produced by the high-energy neutrons and helium is produced by the capture of two thermal neutrons in the reactions: 58 Ni + n → 59 Ni + γ; 59 Ni + n → 56 Fe + α. Examination of type 316 stainless steel specimens irradiated in HFIR has shown that swelling due to cavity formation and degradation of mechanical properties are more severe than can be predicted from fast reactor irradiations, where the helium contents produced are far too low to simulate fusion reactor service. Swelling values are greater and the temperature dependence of swelling is different than in the fast reactor case

  14. Installation of remote-handling typed EBSD-OIM analyzer for heavy irradiated reactor materials

    International Nuclear Information System (INIS)

    Kato, Yoshiaki; Takada, Fumiki; Ohmi, Masao; Nakagawa, Tetsuya; Miwa, Yukio

    2008-06-01

    The remote-handling typed EBSD-OIM analyzer for heavy irradiated reactor materials was installed in the JMTR hot laboratory at the first time in the world. The analyzer is used to study on IASCC (irradiation assisted stress corrosion cracking) or IGSCC (inter granular stress corrosion cracking) in reactor materials. This report describes the measurement procedure, the measured results and the operating experiences on the analyzer in the JMTR hot laboratory. (author)

  15. An automatic device for sample insertion and extraction to/from reactor irradiation facilities

    International Nuclear Information System (INIS)

    Alloni, L.; Venturelli, A.; Meloni, S.

    1990-01-01

    At the previous European Triga Users Conference in Vienna,a paper was given describing a new handling tool for irradiated samples at the L.E.N.A plant. This tool was the first part of an automatic device for the management of samples to be irradiated in the TRIGA MARK ii reactor and successively extracted and stored. So far sample insertion and extraction to/from irradiation facilities available on reactor top (central thimble,rotatory specimen rack and channel f),has been carried out manually by reactor and health-physics operators using the ''traditional'' fishing pole provided by General Atomic, thus exposing reactor personnel to ''unjustified'' radiation doses. The present paper describes the design and the operation of a new device, a ''robot''type machine,which, remotely operated, takes care of sample insertion into the different irradiation facilities,sample extraction after irradiation and connection to the storage pits already described. The extraction of irradiated sample does not require the presence of reactor personnel on the reactor top and,therefore,radiation doses are strongly reduced. All work from design to construction has been carried out by the personnel of the electronic group of the L.E.N.A plant. (orig.)

  16. Design and fabrication of irradiation testing capsule for research reactor materials

    International Nuclear Information System (INIS)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu

    2012-01-01

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed

  17. Design and fabrication of irradiation testing capsule for research reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed.

  18. Investigation of structural materials of reactors using high-energy heavy-ion irradiations

    International Nuclear Information System (INIS)

    Wang Zhiguang

    2007-01-01

    Radiation damage in structural materials of fission/fusion reactors is mainly attributed to the evolution of intensive atom displacement damage induced by energetic particles (n, α and/or fission fragments) and high-rate helium doping by direct α particle bombardments and/or (n, α) reactions. It can cause severe degradation of reactor structural materials such as surface blistering, bulk void swelling, deformation, fatigue, embrittlement, stress erosion corrosion and so on that will significantly affect the operation safety of reactors. However, up to now, behavior of structural materials at the end of their service can hardly be fully tested in a real reactor. In the present work, damage process in reactor structural materials is briefly introduced, then the advantages of energetic ion implantation/irradiation especially high-energy heavy ion irradiation are discussed, and several typical examples on simulation of radiation effects in reactor candidate structural materials using high-energy heavy ion irradiations are pronounced. Experimental results and theoretical analysis suggested that irradiation with energetic particles especially high-energy heavy ions is very useful technique for simulating the evolution of microstructures and macro-properties of reactor structural materials. Furthermore, an on-going plan of material irradiation experiments using high energy H- and He-ions based on the Heavy Ion Research Facilities in Lanzhou (HIRFL) is also briefly interpreted. (authors)

  19. Irradiation creep of various ferritic alloys irradiated at {approximately}400{degrees}C in the PFR and FFTF reactors

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Company, Richland, WA (United States)

    1997-04-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400{degrees}C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 x 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

  20. Radioresistance of immunized animals in internal irradiation

    International Nuclear Information System (INIS)

    Kal'nitskij, S.A.; Ponomareva, T.V.; Shubik, V.M.

    1981-01-01

    The influence of an immunization with bacterial vaccines and antimeasles-gamma-globulin on the radioresistance of raceless white mice was studied. In the vaccinated animals a higher survival rate and duration of life, a better general condition and a better curve of weight and stronger physical stamina were stated compared to the merely irradiated mice. The higher radioresistance is ascribed to the stimulation of cellular and humoral factors of the unspecific protection against infection, to the repair of the lymphoid tissue of the immunized animals and to the decrease in autosensibilization. (author)

  1. Results from the CDE phase activity on neutron dosimetry for the international fusion materials irradiation facility test cell

    CERN Document Server

    Esposito, B; Maruccia, G; Petrizzi, L; Bignon, G; Blandin, C; Chauffriat, S; Lebrun, A; Recroix, H; Trapp, J P; Kaschuck, Y

    2000-01-01

    The international fusion materials irradiation facility (IFMIF) project deals with the study of an accelerator-based, deuterium-lithium source, producing high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials for fusion energy reactors. IFMIF would also provide calibration and validation of data from fission reactor and other accelerator based irradiation tests. This paper describes the activity on neutron/gamma dosimetry (necessary for the characterization of the specimens' irradiation) performed in the frame of the IFMIF conceptual design evaluation (CDE) neutronics tasks. During the previous phase (conceptual design activity (CDA)) the multifoil activation method was proposed for the measurement of the neutron fluence and spectrum and a set of suitable foils was defined. The cross section variances and covariances of this set of foils have now been used for tests on the sensitivity of the IFMIF neutron spectrum determination to cross section uncertainties...

  2. Elaboration of mini plates with U-Mo for irradiation in a high flux reactor

    International Nuclear Information System (INIS)

    Pasqualini, Enrique E.

    2005-01-01

    Full text: International new efforts for the reconversion of HEU in research, testing and radioisotopes production reactors, have greatly incremented U-Mo fuels qualification activities. These qualifications require the resolution of undesired interaction at high fluxes between UMo particles and the aluminum matrix in the case of dispersed fuels and the development of U-Mo monolithic fuels. These efforts are being manifested in the planning and execution of additional series of irradiation tests of mini plates and full size plates. Recently, CNEA has elaborated mini plates with different proposals for the irradiation at the ATR reactor (250 MWTH, maximum thermal neutron flux 10 15 n.cm -2 .seg -1 ) at Idaho National Laboratory, USA. Uranium 7% (w/w) molybdenum (U-7Mo) particles were coated with silicon. Chemical vapour deposition (CVD) of silane and high temperature diffusion of silicon were used. Hydrided, milled and dehydrated (HMD) particles heat treated at 1000 C degrees during four hours and centrifugal atomized powder were coated and the results compared. Mini plates were elaborated with both kinds of particles. Mini plates were also elaborated with U-7Mo and silicon particles dispersed in the aluminium matrix. Monolithic mini plates were also developed by co lamination of U-7Mo with a Zircaloy-4 cladding. The different steps of this process are detailed and the method is shown to be versatile, can be easily scaled up and is performed with small modifications of usual equipment in fuel plants. The irradiation experiment is called RERTR-7A, includes a total of 32 mini plates and it is planed to finalize by mid 2006. (author) [es

  3. Wholesomeness studies in the International Food Irradiation Project

    International Nuclear Information System (INIS)

    Elias, P.S.

    1980-01-01

    Despite more than 25 years history as an effective food preservation method, food irradiation is still subject to strict legislative control in many countries and it is required to carry out scientific investigations to reassure the safety of irradiated food. The International Food Irradiation Project was set up on October 14, 1970 to facilitate the objective evaluation of the wholesomeness of irradiated foodstuffs. Its major activities are; (1) wholesomeness testing of irradiated foods, (2) research on and investigations into the methodology of wholesomeness testing, (3) dissemination of information, and (4) assisting national and international authorities in their consideration of acceptance of irradiated food. In particular, the project over the past nine years had been devoted to the provision of data to national health authorities and international bodies. Up to now, 23 studies were and are being carried out for the project under contract. Subject to the studies include wheat, wheat flour, potatoes, fish, rice, mango, spices, dried dates, onions and cocoa beans. (Kitajima, A.)

  4. The Jules Horowitz reactor, a new high performance European material testing reactor open to international users: present status and objectives

    International Nuclear Information System (INIS)

    Iracane, D.; Bignan, G.

    2010-01-01

    The development of nuclear power as a sustainable and competitive energy source will continue to require research and development of fuel and material behaviour under irradiation. This necessitates a high performance material testing reactor (MTR). Facing the obsolescence of most of the existing MTR in Europe, France decided a few years ago the construction of the RJH (Jules Horowitz reactor). RJH is designed, built and will be operated as an international user facility. A first set of experimental hosting devices is being designed. For instance, there are the in-core CALIPSO Nak integrated loop for material studies and other loops for fuel studies under nominal or off-normal or accidental conditions. The RJH international program will focus on the following subjects: -) fuel reliability, assessed through power ramps tests and post-irradiation examination; -) Loss of coolant tests done out-of-pile in a first phase and in-pile in a possible second phase; and -) source term tests addressing fission products release. The paper reports also the point of view of VATTENFALL (a Swedish power utility), as a potential European RJH user. (A.C.)

  5. Neutron Field Characterization of Irradiation Locations Applied to the Slovenian TRIGA Reactor

    International Nuclear Information System (INIS)

    Barbot, Loic; Domergue, Christophe; Breaud, Stephane; Destouches, Christophe; Villard, Jean-Francois; Snoj, Luka; Stancar, Ziga; Radulovic, Vladimir; Trkov, Andrej

    2013-06-01

    This work deals with several neutron flux measurement instruments and particle transport calculations combined in a method to assess the neutron field in experimental locations in nuclear reactor core or reflector. First test of this method in the TRIGA Mark II of Slovenia led to the assessment of three energy groups neutron fluxes in central irradiation locations within reactor core. (authors)

  6. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion

    International Nuclear Information System (INIS)

    Marques, J.G.; Sousa, M.; Santos, J.P.; Fernandes, A.C.

    2011-01-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1 MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

  7. Internal structure of reactor building for Madras Atomic Power Project

    International Nuclear Information System (INIS)

    Pandit, D.P.

    1975-01-01

    The structural configuration and analysis of structural elements of the internal structure of reactor building for the Madras Atomic Power Project has been presented. Two methods of analysis of the internal structure, viz. Equivalent Plane Frame and Finite Element Method, are explained and compared with the use of bending moments obtained. (author)

  8. 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

  9. International Harmonization of Reactor Licensing Regulations

    International Nuclear Information System (INIS)

    Kuhnt, Dietmar.

    1977-01-01

    The purpose of a harmonization policy for reactor licensing regulations on the basis of already considerable experience is to attain greater rationalisation in this field, in the interest of economic policy and healthy competition, and most important, radiation protection and safety of installations. This paper considers the legal instruments for such harmonization and the conditions for their implementation, in particular within the Communities framework. (NEA) [fr

  10. Halden Reactor Project activities, achievements and international collaboration

    International Nuclear Information System (INIS)

    Wiesenack, W.

    2003-01-01

    This paper concentrates on the Halden Project research programme related to fuel testing. An overview of ongoing tests on WWER fuel performance is also included. The ongoing and planned experiments containing WWER-related fuels and materials - Irradiation of Standard and Modified WWER Fuel (IFA-503) and Corrosion Testing of Different Cladding Alloys (IFA-638) - are presented. The future experiments involving WWER fuel and cladding types foreseen in of the Halden Reactor Project programme are given

  11. Hydrodynamic study of an internal airlift reactor for microalgae culture.

    Science.gov (United States)

    Rengel, Ana; Zoughaib, Assaad; Dron, Dominique; Clodic, Denis

    2012-01-01

    Internal airlift reactors are closed systems considered today for microalgae cultivation. Several works have studied their hydrodynamics but based on important solid concentrations, not with biomass concentrations usually found in microalgae cultures. In this study, an internal airlift reactor has been built and tested in order to clarify the hydrodynamics of this system, based on microalgae typical concentrations. A model is proposed taking into account the variation of air bubble velocity according to volumetric air flow rate injected into the system. A relationship between riser and downcomer gas holdups is established, which varied slightly with solids concentrations. The repartition of solids along the reactor resulted to be homogenous for the range of concentrations and volumetric air flow rate studied here. Liquid velocities increase with volumetric air flow rate, and they vary slightly when solids are added to the system. Finally, liquid circulation time found in each section of the reactor is in concordance with those employed in microalgae culture.

  12. Effects of irradiation on properties of refractory alloys with emphasis on space power reactor applications

    International Nuclear Information System (INIS)

    Wiffen, F.W.

    1984-01-01

    The probable effects of irradiation on niobium and tungsten alloys in use as components of thermionic convertors in a space reactor were reviewed by the author in 1971. While considerably more data on refractory metals have been generated since that time, the data have not been reviewed with respect to space reactor applications. This paper attempts such a review. The approach used is to work from the most recently available review of irradiation effects for each alloy system (where such a review is available) and to discuss that review and more recent data judged to be the most useful in establishing likely behavior in high-temperature reactor service. 28 figures, 6 tables

  13. Large Object Irradiation Facility In The Tangential Channel Of The JSI TRIGA Reactor

    CERN Document Server

    Radulovic, Vladimir; Kaiba, Tanja; Kavsek, Darko; Cindro, Vladimir; Mikuz, Marko; Snoj, Luka

    2017-01-01

    This paper presents the design and installation of a new irradiation device in the Tangential Channel of the JSI TRIGA reactor in Ljubljana, Slovenia. The purpose of the device is to enable on-line irradiation testing of electronic components considerably larger in size (of lateral dimensions of at least 12 cm) than currently possible in the irradiation channels located in the reactor core, in a relatively high neutron flux (exceeding 10^12 n cm^-2 s^-1) and to provide adequate neutron and gamma radiation shielding.

  14. Baffle-former arrangement for nuclear reactor vessel internals

    International Nuclear Information System (INIS)

    Machado, O.J.; Berringer, R.T.

    1978-01-01

    A baffle-former arrangement for the reactor vessel internals of a nuclear reactor is described. The arrangement includes positioning of formers at the same elevations as the fuel assembly grids, and positioning flow holes in the baffle plates directly beneath selected former grid elevations. The arrangement reduces detrimental cross flows, maintains proper core barrel and baffle temperatures, and alleviates the potential of overpressurization within the baffle-former assembly under assumed major accident conditions

  15. Sloshing response of a reactor tank with internals

    International Nuclear Information System (INIS)

    Ma, D.C.; Gvildys, J.; Chang, Y.W.

    1984-01-01

    The sloshing response of a large reactor tank with in-tank components is presented. The study indicates that the presence of the internal components can significantly change the dynamic characteristics of the sloshing motion. The sloshing frequency of a tank with internals is considerably higher than that of a tank without internal. The higher sloshing frequency reduces the sloshing wave height on the free-surface but increases the dynamic pressure in the fluid

  16. Transient heat transfer in a directly-irradiated solar chemical reactor for the thermal dissociation of ZnO

    International Nuclear Information System (INIS)

    Mueller, R.; Lipinski, W.; Steinfeld, A.

    2008-01-01

    A numerical and experimental investigation is carried out in a solar thermochemical reactor for the thermal dissociation of ZnO at 2000 K using concentrated solar energy. The reactor consists of a cavity-receiver lined with ZnO particles and directly exposed to high-flux irradiation. A transient heat transfer model is formulated to link the rate of radiation, convection, and conduction heat transfer to the reaction kinetics. The radiosity and Monte Carlo methods are applied to obtain the distribution of net radiative fluxes at the internal surfaces of the reactor cavity and at the surface of the ZnO bed. Validation is accomplished in terms of the calculated and measured transient temperature profiles and chemical reaction rates

  17. 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.)

  18. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    International Nuclear Information System (INIS)

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO 3 and H 2 O 2 solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area)

  19. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    Energy Technology Data Exchange (ETDEWEB)

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO{sub 3} and H{sub 2}O{sub 2} solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area).

  20. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  1. The OPERA loop in the OSIRIS reactor core. Pressurized-water irradiation device to study Advanced Reactor Fuels

    International Nuclear Information System (INIS)

    Lucot, M.; Roche, M.

    1986-09-01

    This loop is designed to allow fuel qualification test, i.d. to allow irradiation of representative parts fuel assemblies operating in thermohydraulic and chemical conditions representative of these of present pressurized water reactors or in development. This paper presents the aims of the installation, the general design and the main specifications with a brief detailed description [fr

  2. Conceptual design for irradiation device used to irradiate experimental LFR fuel element in TRIGA reactor, ACPR zone

    International Nuclear Information System (INIS)

    Ioan, M.

    2013-01-01

    The paper presents the main steps followed to conceive a small, versatile and rather cheep irradiation device used for irradiation of an experimental fuel element, specific for Lead cooled Fast Reactor (LFR), adapted to TRIGA reactor, ACPR zone. This device must be instrumented with at least 4 thermocouples and a pressure transducer. The fuel element (150 mm fuel pellets column) will be immersed in maximum 0.350 kg pure hot lead (400 deg C). The system has three protection barriers, as follows: first is the fuel tube, second is the lead container (maximum 20 mm inner diameter) and third is the external container (maximum 180 mm outside diameter). Before the reactor pulse, the temperature of the lead is set at the prescribed value using an electrical heater (300 W), coil on the second barrier. Outside the second barrier a very good thermal insulation is provided. (authors)

  3. Method of starting internal pumps of a nuclear reactor

    International Nuclear Information System (INIS)

    Kumagami, Shoji.

    1985-01-01

    Purpose: To reduce the noise effects by decreasing the invading current into the main line upon starting an internal pump type nuclear reactor adapted to forcively recycle the reactor water by a plurality of internal pumps. Method: A plurality of internal pumps are divided into several groups and, upon starting pumps belonging to the individual unit group, the starting instances for the respective pumps are deviated to reduce the surges applied to the main line and suppress the invading current lower to reduce the earth noises. As a result, effects caused to other devices or equipments can be moderated to improve the reliability. Furthermore, by actuating the respective pumps on every group units in a starting pattern along the orthogonal line, flow rate distribution in the reactor can be balanced. Then, the instability region during low rotation of pumps, that is, instability of the flow rate near the resonance frequency can be decreased. (Kawakami, Y.)

  4. An Overview of the International Reactor Physics Experiment Evaluation Project

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Gulliford, Jim

    2014-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  5. Use of Reactor Pressure Vessel Surveillance Materials for Extended Life Evaluations Using Power and Test Reactor Irradiations

    International Nuclear Information System (INIS)

    Server, W.L.; Nanstad, R.K.; Odette, G.R.

    2012-01-01

    The most important component in assuring safety of the nuclear power plant is the reactor pressure (RPV). Surveillance programs have been designed to cover the licensed life of operating nuclear RPVs. The original surveillance programs were designed when the licensed life was 40 years. More than one-half of the operating nuclear plants in the USA have an extended license out to 60 years, and there are plans to continue to operate many plants out to 80 years. Therefore, the surveillance programs have had to be adjusted or enhanced to generate key data for 60 years, and now consideration must be given for 80 or more years. To generate the necessary data to assure safe operation out to these extended license lives, test reactor irradiations have been initiated with key RPV and model alloy steels, which include several steels irradiated in the current power reactor surveillance programs out to relatively high fluence levels. These data are crucial in understanding the radiation embrittlement mechanisms and to enable extrapolation of the irradiation effects on mechanical properties for these extended time periods. This paper describes the potential radiation embrittlement mechanisms and effects when assessing much longer operating times and higher neutron fluence levels. Potential methods for adjusting higher neutron flux test reactor data for use in predicting power reactor vessel conditions are discussed. (author)

  6. Aging management strategy for reactor internals of Korean nuclear power plants

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  7. Multi-physic simulations of irradiation experiments in a technological irradiation reactor; Modelisation pluridisciplinaire d'experiences d'irradiation dans un reacteur d'irradiation technologique

    Energy Technology Data Exchange (ETDEWEB)

    Bonaccorsi, Th

    2007-09-15

    A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

  8. Proceedings of the 4th international symposium on material testing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishihara, Masahiro; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

  9. Proceedings of the 4th international symposium on material testing reactors

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Suzuki, Masahide

    2012-03-01

    This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

  10. IFMIF [International Fusion Materials Irradiation Facility], an accelerator-based neutron source for fusion components irradiation testing: Materials testing capabilities

    International Nuclear Information System (INIS)

    Mann, F.M.

    1988-08-01

    The International Fusion Materials Irradiation Facility (IFMIF) is proposed as an advanced accelerator-based neutron source for high-flux irradiation testing of large-sized fusion reactor components. The facility would require only small extensions to existing accelerator and target technology originally developed for the Fusion Materials Irradiation Test (FMIT) facility. At the extended facility, neutrons would be produced by a 0.1-A beam of 35-MeV deuterons incident upon a liquid lithium target. The volume available for high-flux (>10/sup 15/ n/cm/sup 2/-s) testing in IFMITF would be over a liter, a factor of about three larger than in the FMIT facility. This is because the effective beam current of 35-MeV deuterons on target can be increased by a factor of ten to 1A or more. Such an increase can be accomplished by funneling beams of deuterium ions from the radio-frequency quadruple into a linear accelerator and by taking advantage of recent developments in accelerator technology. Multiple beams and large total current allow great variety in available testing. For example, multiple simultaneous experiments, and great flexibility in tailoring spatial distributions of flux and spectra can be achieved. 5 refs., 2 figs., 1 tab

  11. A Study on Conjugate Heat Transfer Analysis of Reactor Vessel including Irradiated Structural Heat Source

    Energy Technology Data Exchange (ETDEWEB)

    Yi, Kunwoo; Cho, Hyuksu; Im, Inyoung; Kim, Eunkee [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    Though Material reliability programs (MRPs) have a purpose to provide the evaluation or management methodologies for the operating RVI, the similar evaluation methodologies can be applied to the APR1400 fleet in the design stage for the evaluation of neutron irradiation effects. The purposes of this study are: to predict the thermal behavior whether or not irradiated structure heat source; to evaluate effective thermal conductivity (ETC) in relation to isotropic and anisotropic conductivity of porous media for APR1400 Reactor Vessel. The CFD simulations are performed so as to evaluate thermal behavior whether or not irradiated structure heat source and effective thermal conductivity for APR1400 Reactor Vessel. In respective of using irradiated structure heat source, the maximum temperature of fluid and core shroud for isotropic ETC are 325.8 .deg. C, 341.5 .deg. C. The total amount of irradiated structure heat source is about 5.41 MWth and not effect to fluid temperature.

  12. Fast-neutron dosimetry in the seed-irradiation facility, ASTRA reactor, Seibersdorf

    International Nuclear Information System (INIS)

    Ahnstroem, G.; Burtscher, A.; Casta, J.

    1967-01-01

    An important part of the co-ordinated programme on the neutron irradiation of seeds has been the construction of a fast-neutron irradiation facility for swimming-pool reactors. This facility was installed around 70 cm from the core in the ASTRA reactor swimming-pool at the end of December, 1966. Also, for this programme a pair of constant potential ionization chambers have been constructed at the Institute of Biochemistry, Stockholm University. These chambes are of the type described in the technical annex and are the same size as the seed-irradiation vials to be used in the seed-irradiation container (diam. =15 mm, length = 60 mm). Some preliminary dosimetry experiments were undertaken to test the irradiation facility and the ionization chambers, and to investigate the usefulness of the dosimetry instructions in the Technical Annex. The results of these experiments are discussed in this paper. 3 refs, 6 figs, 7 tabs

  13. A study on the proton irradiation effect of reactor materials using cyclotron

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Park, Jong Man; Park, Deuk Keun; Lee, Bong Sang; Oh, Jong Myung

    1993-02-01

    Understanding on radiation damage of important structural materials is important for safe operation and radiation damage evaluation of new reactor structural materials. This study was performed to simulate and evaluate 14 MeV neutron irradiation effects on mechanical properties of candidate structural materials (HT-9/SS316) of next generation reactors (FBR, Fusion) irradiated by Cyclotron(MC-50) using SP test technique. After qualification of SP test techniques from J IC and ε qf correlation, SP tests were performed to evaluate 16MeV proton irradiation effects on mechanical properties of irradiated and unirradiated HT-9/SS316 steels. Test results were evaluated for ε qf , energy and displacement up to failure and J IC change. In addition, damaged zone and dpa upon depth after irradiation were calculated using TRIM code and Doppler broadening line shapes were measured to evaluate defects for 15% cold worked HT-9 steel using PAS. (Author)

  14. Irradiation of fuels and materials in the Jules Horowitz reactor: The 6th European Union JHR co-ordination action (JHR-CA)

    International Nuclear Information System (INIS)

    Iracane, Daniel; Parrat, Daniel

    2005-01-01

    The Fermine thematic network in the 5th FP pointed out the need for a new MTR facility in Europe to answer the continuous need of irradiation capabilities for fission power reactors and fusion facilities and to face the ageing of present MTRs. The Jules Horowitz Reactor (JHR) Project in Cadarache copes with this context, as an international service-oriented user-facility. In the field of nuclear fuels and materials irradiation experiments, a 6th FP co-ordination action, called JHR-CA, has started at the beginning of 2004 for 2 years. The main objective is to network existing expertise on development of a new generation of experimental devices, through definition of conceptual designs, instrumentation and related in-reactor services. This paper presents the outline of the JHR project, the organization of the JHR-CA programme, and a choice of irradiation device conceptual design results. (author)

  15. The international fusion materials irradiation facility

    International Nuclear Information System (INIS)

    Shannon, T.E.; Cozzani, F.; Crandall, D.H.; Wiffen, F.W.; Katsuta, H.; Kondo, T.; Teplyakov, V.; Zavialsky, L.

    1994-01-01

    It is widely agreed that the development of materials for fusion systems requires a high flux, 14 MeV neutron source. The European Union, Japan, Russia and the US have initiated the conceptual design of such a facility. This activity, under the International Energy Agency (IEA) Fusion Materials Agreement, will develop the design for an accelerator-based D-Li system. The first organizational meeting was held in June 1994. This paper describes the system to be studied and the approach to be followed to complete the conceptual design by early 1997

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

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

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

  17. International collaboration on inherently safe nuclear reactors

    International Nuclear Information System (INIS)

    Barkenbus, J.N.

    1989-01-01

    Science and technology transcend economic and political ideologies, providing a means of communications and approach common to both the United States and the Soviet Union. This paper suggests that the field of nuclear fission is a logical and productive area for superpower and broader collaboration, but that the kind of collaboration characteristic of past and present activity is less than it optimally could be. The case for cost sharing is compelling with budget constraints and mounting concerns over global warming. The case for collaboration is based on economic, psychological, and political grounds. A collaborative effort in nuclear fission is presented as a near term effort by building and testing of a prototype reactor in the 1990s

  18. Irradiation testing of LEU fuels in the SILOE Reactor - Progress report

    International Nuclear Information System (INIS)

    Merchie, Francis; Baas, Claude; Martel, Patrick

    1985-01-01

    Irradiation of uranium-silicide fuels has continued in the SILOE reactor during the past year. Thickness vs. fission density data from four U 3 Si plates containing 5.5 and 6.0 g U/cm 3 have been analyzed, and the results are presented. The irradiation of a full 60 g U/cm 3 U 3 Si element has begun. In addition, four U 3 Si 2 plates containing 20 to 54 g U/cm 3 are now being irradiated. These irradiations and future plans are discussed in the paper. (author)

  19. Improved SAFARI-1 research reactor irradiation position modeling in OSCAR-3 code system

    International Nuclear Information System (INIS)

    Moloko, L. E.; Belal, M. G. A. H.

    2009-01-01

    The demand on the availability of irradiation positions in the SAFARI-1 reactor is continuously increasing due to the commercial pressure to produce isotopes more efficiently. This calls for calculational techniques and modeling methods to be improved regularly to optimize irradiation services. The irradiation position models are improved using the OSCAR-3 code system, and results are compared to experimental measurements. It is concluded that the irradiation position models are essential if realistic core follow and reload studies are to be performed and most importantly, for the realization of improved agreement between experimental data and calculated results. (authors)

  20. Colloid bands in silver chloride induced by reactor irradiation at low temperature

    International Nuclear Information System (INIS)

    Atobe, K.; Okada, M.; Nakagawa, M.

    1978-01-01

    It is well known that no trapped electron center exists stably in irradiated silver chlorides even at low temperatures. On the other hand, irradiation by ultra-violet light at room temperature produces a broad absorption (colloid bands) on the long wavelength side of the fundamental absorption. In this report, it is shown that one of the colloid bands appears in undoped AgCl crystals by reactor irradiation at low temperature (20 K) and the other colloid band by thermal annealing after the irradiation. The relation between the bands, which correspond to two types of colloidal silver, is represented. (author)

  1. How to improve the irradiation conditions for the International Fusion Materials Irradiation Facility

    CERN Document Server

    Daum, E

    2000-01-01

    The accelerator-based intense D-Li neutron source International Fusion Materials Irradiation Facility (IFMIF) provides very suitable irradiation conditions for fusion materials development with the attractive option of accelerated irradiations. Investigations show that a neutron moderator made of tungsten and placed in the IFMIF test cell can further improve the irradiation conditions. The moderator softens the IFMIF neutron spectrum by enhancing the fraction of low energy neutrons. For displacement damage, the ratio of point defects to cascades is more DEMO relevant and for tritium production in Li-based breeding ceramic materials it leads to a preferred production via the sup 6 Li(n,t) sup 4 He channel as it occurs in a DEMO breeding blanket.

  2. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    Science.gov (United States)

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  3. Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

    Science.gov (United States)

    Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

    2018-04-01

    1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

  4. Development of inertia-increased reactor internal pump

    International Nuclear Information System (INIS)

    Tanaka, Masaaki; Matsumura, Seiichi; Kikushima, Jun; Kawamura, Shinichi; Yamashita, Norimichi; Kurosaki, Toshikazu; Kondo, Takahisa

    2000-01-01

    The Reactor Internal Pump (RIP) was adopted for the Reactor Recirculation System (RRS) of Advanced Boiling Water Reactor (ABWR) plants, and ten RIPs are located at the bottom of the reactor pressure vessel. In order to simplify the power supply system for the RIPs, a new inertia-increased RIP was developed, which allows to eliminate the Motor-Generator (M-G) sets. The rotating inertia was increased approximately 2.5 times of current RIP inertia by addition of flywheel on its main shaft. A full scale proving test of the inertia-increased RIP under actual plant operating conditions using full scale test loop was performed to evaluate vibration characteristics and coast down characteristics. From the results of this proving test, the validity of the new inertia-increased RIP and its power supply system (without M-G sets) was confirmed. (author)

  5. Lower internals for pressurized water reactor

    International Nuclear Information System (INIS)

    Chevereau, G.; Babin, M.

    1989-01-01

    The lower internals for PWR has a separating plate mounted beneath its lower core plate and defining a distribution chamber with it, peripheral mechanical connectors joining the plates separated by coolant passage and apertures in the separation plate connected to a coolant pipe [fr

  6. Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

    Science.gov (United States)

    Shimomura, Y.

    1994-05-01

    The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

  7. Determination of Cl-36 in Irradiated Reactor Graphite

    International Nuclear Information System (INIS)

    Beer, H.-F.; Schumann, D.; Stowasser, T.; Hartmann, E.; Kramer, A.

    2016-01-01

    Two of the three research reactors at the Paul Scherrer Institute (PSI), the reactors DIORIT and PROTEUS, contained reactor graphite. Whereas the former research reactor DIORIT has been dismantled completely the PROTEUS is subject to a future decommissioning. In case of the DIORIT the reactor graphite was conditioned applying a procedure developed at PSI. In this case the 36 Cl content had to be determined after the conditioning. The result is reported in this paper. The radionuclide inventory including 36 Cl of the graphite used in PROTEUS was measured and the results are reported in here. It has been proven that the graphite from PROTEUS has a radionuclide inventory near the detection limits. All determined radionuclide activities are far below the Swiss exemptions limits. The graphite from PROTEUS therefore poses no radioactive waste. In contrast, the 36 Cl content of graphite from DIORIT is well above the exemption limits. (author)

  8. Effects of irradiation on ceramics for fusion-reactor applications

    International Nuclear Information System (INIS)

    Porter, D.L.

    1982-12-01

    The purpose of this study, coordinated with efforts of LANL and Grumman Aircraft, was to lay some basic groundwork to study the irradiation effects on the engineering properties of some useful classes of ceramic materials; ANL's efforts were pointed towards multiphase materials (glass ceramics and partially-stabilized zirconias). The materials were irradiated at 400 and 550 0 C to fast (E > 0.1 MeV) neutron fluences of approx. 2 x 10 22 n/cm 2 . Fluorophlogapite mica based glass ceramics (Macor, etc.) were found susceptible to weakening due to void formtion between mica plates. Composition variations within this class of glass ceramics seemed to cause sharp variations in the magnitude of the effect. Lithium silicate glass ceramic (ReX) showed sharp contrasts between the effects of ionization irradiation and displacement damage, neutron irradiation having little effect on the ReX structure while electron irradiation creating lithium silicate vitrification and rapid structural annealing

  9. Nuclear reactor internals with control elements guides

    International Nuclear Information System (INIS)

    Baujat, J.; Chevereau, G.

    1991-01-01

    The internals have a lower plate, a superior plate, support columns and guide tubes for the control rods displacements. The lower section of the control rod guide tube have a base that fits into a bevelled seat in the lower plate. The guide tube is held into the seat by a spring, compressed between the base of the upper section of the tube and the lower plate

  10. The post irradiation examination of a sphere-pac (UPu)C fuel pin irradiated in the BR-2 reactor (MFBS 7 experiment)

    International Nuclear Information System (INIS)

    Smith, L.; Aerne, E.T.; Buergisser, B.; Flueckiger, U.; Hofer, R.; Petrik, F.

    1979-09-01

    A pin fuelled with Swiss made (UPu)C microspheres has been successfully irradiated to a peak burn-up of 6% fima in the Belgian BR2 Reactor. The pin, rated up to 95 kW/m, was intact after irradiation and exhibited a peak strain of just over 0.5%. The results of the post irradiation examination are reported. (Auth.)

  11. The application of an internal state variable model to the viscoplastic behavior of irradiated ASTM 304L stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    McAnulty, Michael J., E-mail: mcanulmj@id.doe.gov [Department of Energy, 1955 Fremont Avenue, Idaho Falls, ID 83402 (United States); Potirniche, Gabriel P. [Mechanical Engineering Department, University of Idaho, Moscow, ID 83844 (United States); Tokuhiro, Akira [Mechanical Engineering Department, University of Idaho, Idaho Falls, ID 83402 (United States)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer An internal state variable approach is used to predict the plastic behavior of irradiated metals. Black-Right-Pointing-Pointer The model predicts uniaxial tensile test data for irradiated 304L stainless steel. Black-Right-Pointing-Pointer The model is implemented as a user-defined material subroutine in the finite element code ABAQUS. Black-Right-Pointing-Pointer Results are compared for the unirradiated and irradiated specimens loaded in uniaxial tension. - Abstract: Neutron irradiation of metals results in decreased fracture toughness, decreased ductility, increased yield strength and increased ductile-to-brittle transition temperature. Designers use the most limiting material properties throughout the reactor vessel lifetime to determine acceptable safety margins. To reduce analysis conservatism, a new model is proposed based on an internal state variable approach for the plastic behavior of unirradiated ductile materials to support its use for analyzing irradiated materials. The proposed modeling addresses low temperature irradiation of 304L stainless steel, and predicts uniaxial tensile test data of irradiated experimental specimens. The model was implemented as a user-defined material subroutine (UMAT) in the finite element software ABAQUS. Results are compared between the unirradiated and irradiated specimens subjected to tension tests.

  12. Quality assurance for the IAEA International Database on Irradiated Nuclear Graphite Properties

    International Nuclear Information System (INIS)

    Wickham, A.J.; Humbert, D.

    2006-06-01

    Consideration has been given to the process of Quality Assurance applied to data entered into current versions of the IAEA International Database on Irradiated Nuclear Graphite Properties. Originally conceived simply as a means of collecting and preserving data on irradiation experiments and reactor operation, the data are increasingly being utilised for the preparation of safety arguments and in the design of new graphites for forthcoming generations of graphite-moderated plant. Under these circumstances, regulatory agencies require assurances that the data are of appropriate accuracy and correctly transcribed, that obvious errors in the original documentation are either highlighted or corrected, etc., before they are prepared to accept analyses built upon these data. The processes employed in the data transcription are described in this document, and proposals are made for the categorisation of data and for error reporting by Database users. (author)

  13. General views about specimen irradiations in reactors; Considerations generales sur'les irradiations d'echantillons dans les reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Seguin, M [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1965-07-01

    Specimen irradiation of fissile or non-fissile materials, carried out under circumstances becoming more and more severe and in reactor of increasing flux bas led to an evolution of irradiation rigs. A survey of the problems arising from irradiating under these various circumstances leads to conclude that it is possible to devise one capsule type suitable to every particular case, and that in a wide temperature range. Consequently, once the various irradiation-parameters known, a general method of calculation can be followed so as to determine the various sizes of the parts constituting the capsule. These theoretical calculations might sometimes be corrected through benefits gained from previous irradiations. Similarly, practical experimentation might allow to foresee more handy assembling of the capsule, specimen loading-and unloading being easier at the same time. (author) [French] L'irradiation d'echantillons, fissiles ou non fissiles, dans des conditions imposees de plus en plus strictes et dans des reacteurs a flux de plus en plus eleve, a eu pour consequence une evolution dans la conception des dispositifs d'irradiation. Lorsqu'on examine les problemes souleves par ces differentes irradiations, on en conclut qu'il est possible de concevoir un type de capsule capable de donner satisfaction dans chaque cas particulier, et ce, dans une tres large gamme de temperature. Par consequent, les differents parametres de l'irradiation etant connus, une methode generale de calcul peut etre suivie pour determiner les differentes cotes des pieces constitutives de la capsule. Ces calculs theoriques devront quelquefois etre corriges grace aux enseignements tires d'irradiations precedentes. De meme, l'experience acquise permettra d'envisager un montage plus aise de la capsule, tout en facilitant l'enfournement et le defournement des echantillons.

  14. Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor

    Energy Technology Data Exchange (ETDEWEB)

    Magielsen, A.J., E-mail: magielsen@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Jong, M.; Bakker, T.; Luzginova, N.V.; Mutnuru, R.K.; Ketema, D.J.; Fedorov, A.V. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands)

    2011-08-31

    In the framework of the materials domain DEMETRA in the European Transmutation research and development project EUROTRANS, irradiation experiment IBIS has been performed in the High Flux Reactor in Petten. The objective was to investigate the synergystic effects of irradiation and lead bismuth eutectic exposure on the mechanical properties of structural materials and welds. In this experiment ferritic martensitic 9 Cr steel, austenitic 316L stainless steel and their welds have been irradiated for 250 Full Power Days up to a dose level of 2 dpa. Irradiation temperatures have been kept constant at 300 deg. C and 500 deg. C. During the post-irradiation test phase, tensile tests performed on the specimens irradiated at 300 deg. C have shown that the irradiation hardening of ferritic martensitic 9 Cr steel at 1.3 dpa is 254 MPa, which is in line with the irradiation hardening obtained for ferritic martensitic Eurofer97 steel investigated in the fusion program. This result indicates that no LBE interaction at this irradiation temperature is present. A visual inspection is performed on the specimens irradiated in contact with LBE at 500 deg. C and have shown blackening on the surface of the specimens and remains of LBE that makes a special cleaning procedure necessary before post-irradiation mechanical testing.

  15. Assessment of models predicting irradiation effects on tensile properties of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Pineau, L.; Landron, C.

    2015-01-01

    In this paper, an analysis of tensile data acquired as part of the French Reactor Vessel Surveillance Program (RVSP) is produced. This program contains amongst other mechanical tests, tensile tests at 20 and 300 C degrees on non irradiated base metals and at 300 C degrees only on irradiated materials. It shows that irradiation leads to an increase in the yield strength and a decrease in the strain hardening. The exploitation of tensile results has permitted to express a relationship between yield strength increase measured and fluence value, as well as between strain hardening decrease and yield strength evolution. The use of these relations in the aim at predicting evolution of tensile properties with irradiation has then permitted to propose a methodology to model entire stress-strain curves of irradiated base metal only based on the non irradiated stress-strain curve. These predictions were successfully compared with an experimental standard case. (authors)

  16. Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takamizawa, Hisashi, E-mail: takamizawa.hisashi@jaea.go.jp; Itoh, Hiroto, E-mail: ito.hiroto@jaea.go.jp; Nishiyama, Yutaka, E-mail: nishiyama.yutaka93@jaea.go.jp

    2016-10-15

    In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased.

  17. The characteristics and mechanism of apoptosis induced by internal irradiation

    International Nuclear Information System (INIS)

    Hong Chengjiao; Zhang Junning; Zhu Shoupeng

    2001-01-01

    Apoptosis in tumor cells induced by radionuclides is likely the most effective way to cure cancer. In order to explore the possibility in clinic application, the characteristics and mechanism of apoptosis induced by internal irradiation were investigated. The apoptosis and expressions of bcl-2mRNA, bcl-2 and bax of K 562 cells following internal exposure with different accumulated absorbed doses of strontium-89 were studied. 6 h after irradiation, the characteristics of apoptosis and necrosis appeared in K 562 cells. The apoptosis and necrosis enhanced with the prolongation of internally contaminated time at 6 h, 9 h, 12 h, 24 h and 48 h. The expressions of bcl-2mRNA decreased at 12 h, most remarkably at 24 h. The expressions of bcl-2 decreased after irradiation whereas bax had no obvious changes. The results suggest that the apoptosis induced by internal exposure may be regulated by lower expressions of bcl-2mRNA and bcl-2, lower bcl-2/bax value

  18. Food Preservation by Irradiation. Vol. II. Proceedings of an International Symposium on Food Preservation by Irradiation

    International Nuclear Information System (INIS)

    1978-01-01

    In the task of alleviating the distress caused by the world-wide food shortage it is essential to preserve what has been grown and harvested in the fields. Clearly all suitable methods for preserving agricultural produce and food should be made use of. In this context treatment with ionizing radiation has proved its value as an environmentally clean, physical method of food preservation which is low in its energy requirement, but the volume of food being processed in this way is still low. The introduction of food irradiation on a global basis poses certain economic, legal, regulatory and health-related questions, the solution of which requires close international collaboration. Such collaboration between many international, intergovernmental and national organizations began over a decade ago. The need for dissemination and discussion of information gained through research and development work on this subject became apparent, and a number of inter-regional meetings were held. The last international symposium on the topic was held jointly by FAO and the IAEA in 1972 in Bombay. To review progress made since then, FAO and the IAEA, together with WHO, convened the present Symposium on 21-25 November 1977. It appeared timely to hold this Symposium for the following reasons: (1) Apart from significant scientific work reported in the literature, progress in other directions between 1972 and 1977 had also been made. For example, the number of food items authorized by governments, with or without restriction, had grown from 19 to 26, and the number of countries accepting one or more irradiated foods for human consumption had increased from 11 to 19. (2) Largely on the basis of the work of the International Project in the Field of Food Irradiation (Karlsruhe), already described at the Bombay Symposium, an international expert committee, jointly convened by FAO, the IAEA and WHO in August-September 1976, had made important statements on the philosophy of wholesomeness studies

  19. Food Preservation by Irradiation. Vol. I. Proceedings of an International Symposium on Food Preservation by Irradiation

    International Nuclear Information System (INIS)

    1978-01-01

    In the task of alleviating the distress caused by the world-wide food shortage it is essential to preserve what has been grown and harvested in the fields. Clearly all suitable methods for preserving agricultural produce and food should be made use of. In this context treatment with ionizing radiation has proved its value as an environmentally clean, physical method o f food preservation which is low in its energy requirement, but the volume of food being processed in this way is still low. The introduction o f food irradiation on a global basis poses certain economic, legal, regulatory and health-related questions, the solution o f which requires close international collaboration. Such collaboration between many international, intergovernmental and national organizations began over a decade ago. The need for dissemination and discussion o f information gained through research and development work on this subject became apparent, and a number of inter-regional meetings were held. The last international symposium on the topic was held jointly by FAO and the IAEA in 1972 in Bombay. To review progress made since then, FAO and the IAEA, together with WHO, convened the present Symposium on 21-25 November 1977. It appeared timely to hold this Symposium for the following reasons: (1) Apart from significant scientific work reported in the literature, progress in other directions between 1972 and 1977 had also been made. For example, the number of food items authorized by governments, with or without restriction, had grown from 19 to 26, and the number of countries accepting one or more irradiated foods for human consumption had increased from 11 to 19. (2) Largely on the basis of the work of the International Project in the Field of Food Irradiation (Karlsruhe), already described at the Bombay Symposium, an international expert committee, jointly convened by FAO, the IAEA and WHO in August-September 1976, had made important statements on the philosophy of wholesomeness

  20. IFMIF - International Fusion Materials Irradiation Facility Conceptual Design Activity/Interim Report

    International Nuclear Information System (INIS)

    Rennich, M.J.

    1995-12-01

    Environmental acceptability, safety, and economic viability win ultimately be the keys to the widespread introduction of fusion power. This will entail the development of radiation- resistant and low- activation materials. These low-activation materials must also survive exposure to damage from neutrons having an energy spectrum peaked near 14 MeV with annual radiation doses in the range of 20 displacements per atom (dpa). Testing of candidate materials, therefore, requires a high-flux source of high energy neutrons. The problem is that there is currently no high-flux source of neutrons in the energy range above a few MeV. The goal, is therefore, to provide an irradiation facility for use by fusion material scientists in the search for low-activation and damage-resistant materials. An accellerator-based neutron source has been established through a number of international studies and workshops' as an essential step for materials development and testing. The mission of the International Fusion Materials Irradiation Facility (IFMIF) is to provide an accelerator-based, deuterium-lithium (D-Li) neutron source to produce high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials up to about a full lifetime of anticipated use in fusion energy reactors. would also provide calibration and validation of data from fission reactor and other accelerator-based irradiation tests. It would generate material- specific activation and radiological properties data, and support the analysis of materials for use in safety, maintenance, recycling, decommissioning, and waste disposal systems

  1. Characterization of un-irradiated and irradiated reactor graphite; Karakterizacija neozracenog i ozracenog reaktorskog grafita

    Energy Technology Data Exchange (ETDEWEB)

    Marinkovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This report contains three parts: characterization of Yugoslav nuclear graphite development of methods and obtained results, characterization of un-irradiated and irradiated domestic nuclear graphite; calculation of electrical conductivity changes due to vacancies in the graphite crystal lattice.

  2. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  3. Tests for removal of Co-60 and Eu-154 from irradiated graphite in the TRIGA Reactor

    International Nuclear Information System (INIS)

    Arsene, Carmen

    2009-01-01

    The irradiated graphite in Romania is mainly generated in the thermal columns of TRIGA and WWER-S research reactors (about 9 tones). It was found that the radionuclide content of the graphite irradiated in the TRIGA research reactor is mainly due to C-14 (103 Bq/g), Eu-152 (600-700 Bq/g) and Co-60 (130-150 Bq/g) and low amounts of Eu-154 and Cs-137, depending on location in the thermal column and on irradiation history. In order to minimize the waste inventory and volume in view of their final disposal, in the present paper we show the results of experiments performed for developing and optimizing methods for the chemical decontamination of the irradiated graphite. These procedures are based on strong alkaline solutions for Eu-152 and strong acid solutions for Co-60. The influence of the process parameters on the decontamination factor is investigated. (authors)

  4. New temperature monitoring devices for high-temperature irradiation experiments in the high flux reactor Petten

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M.; Futterer, M. A.; Lapetite, J. M. [European Commission Joint Research Centre, Institute for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Fourrez, S. [THERMOCOAX SAS, BP 26, Planquivon, 61438 Flers Cedex (France); Morice, R. [Laboratoire National de Metrologie et d' Essais, 1 rue Gaston Boissier, 75724 Paris (France)

    2009-07-01

    Within the European High Temperature Reactor Technology Network (HTR-TN) and related projects a number of HTR fuel irradiations are planned in the High Flux Reactor Petten (HFR), The Netherlands, with the objective to explore the potential of recently produced fuel for even higher temperature and burn-up. Irradiating fuel under defined conditions to extremely high burn-ups will provide a better understanding of fission product release and failure mechanisms if particle failure occurs. After an overview of the irradiation rigs used in the HFR, this paper sums up data collected from previous irradiation tests in terms of thermocouple data. Some research and development work for further improvement of thermocouples and other on-line instrumentation will be outlined. (authors)

  5. Public information circular for shipments of irradiated reactor fuel. Revision 4

    International Nuclear Information System (INIS)

    1984-06-01

    This publication is the fifth in a series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. This publication contains basically three kinds of information: (1) routes recently approved (18 months) by the Commission for the shipment of irradiated reactor fuel; (2) information regarding any safeguards-significant incidents that may be (to date none have) reported during shipments along such routes; and (3) cumulative amounts of material shipped

  6. Recommendations on the measurement of irradiation received by the structural materials of reactors

    International Nuclear Information System (INIS)

    Genthon, J.P.; Mas, P.; Wright, S.B.; Zijp, W.L.

    1975-01-01

    The recommendations have been compiled by a working group Radiation Damage which has been set up by the Euratom Working Group for reactor Dosimetry. The parameters are indicated which must be defined for the characterisation of the neutron dose causing radiation-induced damage in construction materials important for reactor technique. Following an explanation of some theoretical aspects, practical guidelines for neutron metrology on irradiation of graphite and of metals are given. A thorough knowledge of the spectrum of the incident neutrons is required for a proper interpretation of the results of irradiation experiments

  7. Preliminary Options Assessment of Versatile Irradiation Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 x 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.

  8. International standardization of nuclear reactor designs - the way forward

    International Nuclear Information System (INIS)

    Raetzke, Christian

    2010-01-01

    The concept of 'International Standardization of Nuclear Reactor Designs' means that vendors could build their designs in every country without having to adapt it specifically to national safety requirements. Such standardization would have two main effects. It would greatly facilitate nuclear new build worldwide by giving greater efficiency and certainty to the national licensing procedures; by taking into account the fact that vendors, and nowadays also utilities, are active across borders; by helping developing countries to establish their nuclear new build programmes; and by reducing the strain on human resources on both the regulators' and the industry's side. The second valuable effect of standardization would be to further enhance safety by improving the exchange of construction and operating experience among a number of reactors belonging to fleets of the same design. The World Nuclear Association's CORDEL (Cooperation in Reactor Design Evaluation and Licensing) Group has developed a concept for implementation of international standardization of reactor designs. It has defined a number of steps to be taken by industry. At the same time, possibilities offered by national and international regulatory mechanisms would have to be fully made use of, and some changes in regulatory frameworks might be necessary. Some steps especially towards greater cooperation of regulators have already been taken; however, much still remains to be done. The concept of deploying standardized reactor designs across a number of countries supposes an alignment and, if possible, harmonization of national safety standards; a streamlining of national licensing procedures, making them more efficient and predictable; and the willingness of national regulators to take into account licensing done in other countries. In the end, this should lead to a mutual acceptance of design approvals or, in a more distant future, even to a multinational design approval process. All in all, the concept

  9. Participation of IRD/CNEN-Br in International Intercomparison of Criticality Accident Dosimetry Systems at Silene reactor, France

    International Nuclear Information System (INIS)

    Mauricio, Claudia Lucia P.; Fonseca, Evaldo S. da

    1996-01-01

    IRD has participated in an International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE reactor, France on June 1993. The dosemeters were irradiated on phantoms and free in air, in bare and lead shield reactor pulses, simulating different irradiation fields that can be found in criticality accidents. Comparing with the reference measurements, the calculated mean neutron kerma found by IRD was only 2% greater for lead shield and 14% greater for bare reactor. For gamma absorbed dose, the differences were, respectively + 22% and -9% for the dosemeters free in air and -19% and -9% for dosemeters on phantoms. IRD results are closer to the real values than the mean values measured by the participants. IRD results show a good performance if its simple criticality accident system. (author)

  10. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  11. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    West, C.D.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

  12. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    West, C D [comp.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN{sub 2} test, Source LH2-H{sub 2}O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface.

  13. Review of inspection of Torness reactor internals using remote techniques

    International Nuclear Information System (INIS)

    Dynan, J.

    1993-01-01

    Inspection of reactor internals is increasingly being achieved by deployment of cameras into areas of sometimes high radiation fields using manipulators. The manipulators have been developed over a number of years and from a variety of sources and those provided at Torness can maybe be seen to be the state of the art development of such machines for AGRS. Torness and Heysham 2 reactors were specifically designed for inspection using remote techniques and special facilities have been provided for this purpose. This paper is written by an operator and not a manipulator specialist and is intended to show what the operator needs versus what the designer gave him. (author)

  14. Ratcheting problems for ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Majumdar, S.

    1991-01-01

    Because of the presence of high cyclic thermal stress, pressure-induced primary stress, and disruption-induced high cyclic primary stress, ratcheting of the first wall poses a serious challenge to the designers of ITER (International Thermonuclear Experimental Reactor). Existing design tools such as the Bree diagram in the ASME Boiler and Pressure Vessels Code, are not directly applicable to ITER, because of important differences in geometry and loading modes. Available alternative models for ratcheting are discussed and new Bree diagrams, that are more relevant for fusion reactor applications, are proposed. 9 refs., 17 figs

  15. Summary of the fifth international conference on reactor shielding

    International Nuclear Information System (INIS)

    Roussin, R.W.; Abbott, L.S.; Bartine, D.E.

    1977-01-01

    The Fifth International Conference on Reactor Shielding was held April 18-23, 1977 in Knoxville, Tennessee. The meeting was the largest in the series and attracted participants from 34 countries. The 10 invited papers and 10 of the contributed papers, selected as being representative of the Conference by the Technical Program Committee, are published in this issue of ATOMKERNENERGIE. This collection of papers demonstrates that the field of nuclear reactor shielding has developed into a mature discipline while retaining a definite vitality. (orig.) [de

  16. Design of reactor internals in larger high-temperature reactors with spherical fuel elements

    International Nuclear Information System (INIS)

    Elter, C.

    1981-01-01

    In his paper, the author analyzes and summarizes the present state of the art with emphasis on the prototype reactor THTR 300 MWe, because in addition to spherical fuel elements, this type includes other features of future HTR design such as the same flow direction of cooland gas through the core. The paper on hand also elaborates design guidelines for reactor internals applicable with large HTR's of up to 1200 MWe. Proved designs will be altered so as to meet the special requirements of larger cores with spherical elements to be reloaded according to the OTTO principle. This paper is furthermore designed as a starting point for selective and swift development of reactor internals for large HTR's to be refuelled according to the OTTO principle. (orig./GL) [de

  17. Vibration behavior of PWR reactor internals Model experiments and analysis

    International Nuclear Information System (INIS)

    Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

    1975-01-01

    In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

  18. Moderator behaviour and reactor internals integrity at Atucha I NPP

    International Nuclear Information System (INIS)

    Berra, S.; Guala, M.; Herzovich, P.; Chocron, M.; Lorenzo, A.; Raffo Calderon, Ma. C. del; Urrutia, G.

    1996-01-01

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab

  19. Moderator behaviour and reactor internals integrity at Atucha I NPP

    Energy Technology Data Exchange (ETDEWEB)

    Berra, S; Guala, M; Herzovich, P [Central Nuclear Atucha I, Nucleoelectrica Argentina, Lima, Buenos Aires (Argentina); Chocron, M; Lorenzo, A; Raffo Calderon, Ma. C. del; Urrutia, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes

    1997-12-31

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab.

  20. Program of assessment of mechanical and corrosion mechanical properties of reactor internals materials due to operation conditions in WWERs

    International Nuclear Information System (INIS)

    Ruscak, M.; Zamboch, M.

    1998-01-01

    Reactor internals are subject to three principle operation influences: neutron and gamma irradiation, mechanical stresses, both static and dynamic, and coolant chemistry. Several cases of damage have been reported in previous years in both boiling and pressure water reactors. They are linked with the term of irradiation assisted stress corrosion cracking as a possible damage mechanism. In WWERs, the principal material used for reactor internals is austenitic titanium stabilized stainless steel 08Kh18N10T, however high strength steels are used as well. To assess the changes of mechanical properties and to determine whether sensitivity to intergranular cracking can be increased by high neutron fluences, the experimental program has been started. The goal is to assure safe operation of the internals as well as life management for all planned operation period. The program consists of tests of material properties, both mechanical and corrosion-mechanical. Detailed neutron fluxes calculation as well as stress and deformation calculations are part of the assessment. Model of change will be proposed in order to plan inspections of the facility. In situ measurements of internals will be used to monitor exact status of structure during operation. Tensile specimens manufactured from both base metal and model weld joint have been irradiated to the total fluences of 3-20 dpa. Changes of mechanical properties are tested by the tensile test, stress corrosion cracking tests are performed in the autoclave with water loop and active loading. Operation temperature, pressure and water chemistry are chosen for the tests. (author)

  1. Irradiation effects on Zr-2.5Nb in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

  2. International Conference on Physics and Technology of Reactors and Applications

    International Nuclear Information System (INIS)

    2007-01-01

    The first international conference on physics and technology of reactors and applications (PHYTRA 1) which took place in Marrakech (Morocco) from 14 to 16 March 2007, was designed to bring together scientists, teachers and students from universities, research centres and industry and other institutions to exchange knowledge and to discuss ideas and future issues. The programmes of the PHYTRA 1 conference covers a wide variety topics, the conference was organised in three plenary sessions, ten oral technical sessions and two poster sessions. The plenary sessions covers the following topics : The prospects of nuclear energy, The situation of nuclear sciences and energy in Morocco and Africa, and the new development in reactor physics and reactor design [fr

  3. Neutronic analysis for the fission Mo99 production by irradiation of leu targets in TRIGA 14 MW reactor

    International Nuclear Information System (INIS)

    Dulugeac, S. D.; Mladin, M.; Budriman, A. G.

    2013-01-01

    Molybdenum production can be a solution for the future in the utilization of the Romanian TRIGA, taking into account the international market supply needs. Generally, two different techniques are available for Mo 99 production for use in medical Tc 99 generation.The first one is based on neutron irradiation of molybdenum targets of natural isotopic composition or enriched in Mo 98 . In a second process, Mo 99 is obtained as a result of the neutron induced fission of U 235 according to U 235 (n,f) Mo 99 . The objectives of the paper are related to Mo 99 production as a result of fission. Neutron physics parameters are determined and presented, such as: thermal flux axial distribution for the critical reactor at 10 MW inside the irradiation location; reactivity introduced by three Uranium foil containers; neutron fluxes and fission rates in the Uranium foils; released and deposited power in the Uranium foils; Mo 99 activity in the Uranium foils. (authors)

  4. Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

    2015-03-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

  5. Effect of irradiation-induced defects on fusion reactor ceramics

    International Nuclear Information System (INIS)

    Clinard, F.W. Jr.

    1986-01-01

    Structural, thermal, and electrical properties critical to performance of ceramics in a fusion environment can be profoundly altered by irradiation effects. Neutron damage may cause swelling, reduction of thermal conductivity, increase in dielectric loss, and either reduction or enhancement of strength depending on the crystal structure and defect content of the material. Absorption of ionizing energy inevitably leads to degradation of insulating properties, but these changes can be reduced by alterations in structural or compositional makeup. Assessment of the irradiation response of candidate ceramics Al 2 O 3 , MgAl 2 O 4 , SiC and Si 3 N 4 shows that each may find use in advanced fusion devices. The present understanding of irradiation-induced defects in ceramics, while far from complete, nevertheless points the way to methods for developing improved materials for fusion applications

  6. Proceedings 21. International Conference on Applied Physics of Condensed Matter and of the Scientific Conference Advanced Fast Reactors

    International Nuclear Information System (INIS)

    Vajda, J.; Jamnicky, I.

    2015-01-01

    The 21. International Conference on Applied Physics of Condensed Matter was held on 24-26 June, 2015 on Strbske Pleso, Strba, Slovakia. The Scientific Conference the Advanced Fast Reactors was part of the 21 st International Conference on APCOM 2015. The specialists discussed various aspects of modern problems in: Physical properties and structural aspects of solid materials and their influencing; Advanced fast reactors; Physical properties and structural aspects of solid materials and their influencing; Nuclear science and technology, influence of irradiation on physical properties of materials, radiation detection; Computational physics and theory of physical properties of matter; interdisciplinary physics of condensed matter; Nuclear science and technology, influence of irradiation on physical properties of materials, radiation detection; Optical phenomena in materials, photovoltaics and photonics, new principles in sensors and detection methods. Fifty seven contributions relevant of INIS interest has been inputted to INIS.

  7. Caramel fuel for research reactors: experience acquired in the fabrication, monitoring and irradiation of Osiris core

    International Nuclear Information System (INIS)

    Contenson, Ghislain de; Foulquier, Henri; Trotabas, Maria; Vignesoult, Nicole; Cerles, J.-M.; Delafosse, Jacques.

    1981-06-01

    A plate type nuclear fuel (Caramel fuel) has been developed in France in the framework of the various activities pursued in the design, fabrication and development of nuclear fuels by the CEA. This fuel can be adapted to various different categories of water cooled reactor (power reactors, marine propulsion reactors, urbain heating reactors, research reactors). The successful work conducted in this field led the realization of a complete core and reloads for the high performance research reactor, Osiris, at Saclay. The existing highly enriched U-Al alloy fuel was replaced by a non-proliferating low enrichment (7%) caramel fuel. This new core has been operating successfully since january 1980. A brief description of Caramel and its main advantages is given. The way in which it is fabricated is described together with the quality controls to which it is subjected. The qualification program and the main results deduced from it are also presented. The program used to monitor its in-pile behavior is described. The essential purpose of this program is to ensure the high performance of the fuel under irradiation. The successful operation of Osiris, which terminated 11 irradiation cycles on the 21st of April 1981 confirmed the correctness of the decisions made and the excellent performance that could be achieved with these fuel elements under the severe conditions encountered in a high performance research reactor [fr

  8. Qualification of the on-line power determination of fuel elements in irradiation devices in the BR2 reactor

    International Nuclear Information System (INIS)

    Vermeeren, L.; Dekeyser, J.; Gouat, P.; Kalcheva, S.; Koonen, E.; Kuzminov, V.; Verwimp, A.; Weber, M.

    2005-01-01

    Fuel irradiation tests require an on-line monitoring of the fuel power. In the BR2 reactor, this is performed by continuously measuring the enthalpy change in the coolant of the irradiation device and complementing this information with data on power losses, heating of structure parts and spatial power profiles from mock-up test experiments and from calculations. Since a few years Monte Carlo codes (MCNP) are used, describing the BR2 core in great detail for every reactor cycle with its specific core load, yielding not only reliable relative values, but also calculated absolute local power values in agreement with data from PIE analyses. Several methods were conceived to combine the experimental and calculated data for the on-line calculation of the local linear power in the fuel elements; their internal consistency and the consistency with gamma spectroscopy data and data from radiochemical fission product analysis was checked. The data show that fuel irradiations in BR2 can be performed in a well-controlled way, with an accurate and reliable on-line follow-up of the fuel power. (author)

  9. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

    2014-09-01

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  10. International standardization of safety requirements for fast reactors

    International Nuclear Information System (INIS)

    2011-06-01

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

  11. Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

    1990-05-01

    The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

  12. Dimensional control and check of field machining parts for reactor internals installation

    International Nuclear Information System (INIS)

    Zhang Caifang

    2010-01-01

    Some key issues of dimensional control for reactor internals installation are analyzed, and important technical requirements of crucial quality control elements on the measurement, machining, and checking of reactor internals filed machining parts are discussed. Moreover, provisions on quality control and risk prevention of reactor internals filed machining parts are presented in this paper. (author)

  13. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  14. 75 FR 62695 - Physical Protection of Irradiated Reactor Fuel in Transit

    Science.gov (United States)

    2010-10-13

    ... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule. SUMMARY: The... nuclear fuel in transit? H. Why require a telemetric position monitoring system or an alternative tracking... nuclear fuel in transit. The interim final rule added 10 CFR 73.37, ``Requirements for Physical Protection...

  15. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Science.gov (United States)

    2010-01-01

    ... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1...

  16. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Kocik, J.; Keilova, E.

    1993-01-01

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs

  17. Formation of the high-spin Hf-179m2 isomer in reactor irradiations

    Czech Academy of Sciences Publication Activity Database

    Karamian, S. A.; Carroll, J. J.; Adam, Jindřich; Kulagin, EN.; Shabalin, EP.

    2004-01-01

    Roč. 14, č. 4 (2004), s. 438-441 ISSN 1054-660X R&D Projects: GA MŠk(CZ) ME 134 Keywords : reactor irradiation * high-spin Hf-179m2 Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.836, year: 2004

  18. Tensile properties and microstructure of helium injected and reactor irradiated V-20 Ti

    International Nuclear Information System (INIS)

    Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

    1980-01-01

    The objective of this work was to determine the effect of preinjected helium followed by neutron irradiation on the mechanical properties and microstructure of V-20% Ti. These results will be used for the evaluation of the potential use of V-20% Ti in fusion reactor service

  19. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Science.gov (United States)

    2010-01-01

    ... fuel and nuclear waste. 71.97 Section 71.97 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... advance notification of transportation of nuclear waste was published in the Federal Register on June 30...

  20. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Kocik, J; Keilova, E [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding) to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs.

  1. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  2. Using TRIGA Mark II research reactor for irradiation with thermal neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Kolšek, Aljaž, E-mail: aljaz.kolsek@gmail.com; Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si; Trkov, Andrej, E-mail: andrej.trkov@ijs.si; Snoj, Luka, E-mail: luka.snoj@ijs.si

    2015-03-15

    Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 10{sup 15} neutrons/cm{sup 2} in irradiation time of 20 h.

  3. Preparations for the shipment of RA-3 reactor irradiated fuel

    International Nuclear Information System (INIS)

    Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

    2002-01-01

    During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

  4. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  5. Characterization of decommissioned reactor internals: Monte Carlo analysis technique

    International Nuclear Information System (INIS)

    Reid, B.D.; Love, E.F.; Luksic, A.T.

    1993-03-01

    This study discusses computer analysis techniques for determining activation levels of irradiated reactor component hardware to yield data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. The study recommends the Monte Carlo Neutron/Photon (MCNP) computer code as the best analysis tool for this application and compares the technique to direct sampling methodology. To implement the MCNP analysis, a computer model would be developed to reflect the geometry, material composition, and power history of an existing shutdown reactor. MCNP analysis would then be performed using the computer model, and the results would be validated by comparison to laboratory analysis results from samples taken from the shutdown reactor. The report estimates uncertainties for each step of the computational and laboratory analyses; the overall uncertainty of the MCNP results is projected to be ±35%. The primary source of uncertainty is identified as the material composition of the components, and research is suggested to address that uncertainty

  6. NRC data base for power reactor surveillance programs and for irradiation experiments results

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.

    1991-01-01

    The radiation damage of pressure vessel materials in nuclear reactors depends on many different factors, primarily fluence, fluence spectrum, fluence rate, irradiation temperature, and chemistry. These factors and, possibly, others such as heat treatment and type of flux used in weldments must be considered to reliably predict the pressure vessel embrittlement and to assure the safe operation of the reactor. Based on embrittlement predictions, decisions must be made concerning operating parameters, low-leakage fuel management, possible life extension, and the need for annealing of the pressure vessel. Large numbers of data obtained from surveillance capsules and test reactor experiments are needed, comprising many different materials and different irradiation conditions, to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. The US Nuclear Regulatory Agency has, therefore, sponsored a project to construct an Embrittlement Data Base (EDB) for a comprehensive collection of data concerning changes in material properties of pressure vessel steels due to neutron irradiation. A first version containing data from surveillance capsules of commercial power reactors, the Power Reactor Embrittlement Data Base (PR-EDB) Version 1, has been completed and is available to authorized users from the Radiation Shielding Information Center at the Oak Ridge National Laboratory. This document provides a discussion of the features of the current database. 1 fig

  7. 3. International conference on catalysis in membrane reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    The 3. International Conference on Catalysis in Membrane Reactors, Copenhagen, Denmark, is a continuation of the previous conferences held in Villeurbanne 1994 and Moscow 1996 and will deal with the rapid developments taking place within membranes with emphasis on membrane catalysis. The approx. 80 contributions in form of plenary lectures and posters discuss hydrogen production, methane reforming into syngas, selectivity and specificity of various membranes etc. The conference is organised by the Danish Catalytic Society under the Danish Society for Chemical Engineering. (EG)

  8. Contributions to the sixth international conference on fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-11-15

    The ICFRM series has documented progress in the field of fusion reactor materials since the first conference held in Tokyo in 1984. The conference series has continually increased its coverage to the point where it now includes the comprehensive range of materials science and technology areas that enable systems designers to meet the needs of current experiments and to present innovative solutions for future energy systems. This publication contains five contributions to the sixth international conference which have each been indexed separately.

  9. Combined development of international nuclear fusion test reactors

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    Ambassadors of the four most important partners (Common Market, Japan, USA and USSR) in the IAEA sponsored INTOR project, met on the 15 and 16 March 1987 in Vienna under the auspices of the IAEA. A press release was issued acknowledging the considerable technical progress made in magnetic nuclear fusion research. Future design concepts, assistance in research and development work and other activities towards the provision of an international experimental thermonuclear reactor were discussed. (G.T.H.)

  10. Gamma dose from activation of internal shields in IRIS reactor.

    Science.gov (United States)

    Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

  11. Gamma dose from activation of internal shields in IRIS reactor

    International Nuclear Information System (INIS)

    Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

  12. IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

    International Nuclear Information System (INIS)

    Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

    2002-01-01

    Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

  13. Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin

    International Nuclear Information System (INIS)

    Matos, J.E.

    1994-12-01

    This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy's Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006

  14. Specific features of reactor or cyclotron {alpha}-particles irradiated beryllium microstructure

    Energy Technology Data Exchange (ETDEWEB)

    Khomutov, A M [A.A.Bochvar All-Russia Research Inst. of Inorganic Materials (VNIINM), Moscow (Russian Federation); Gromov, B F; Karabanov, V N [and others

    1998-01-01

    Studies were carried out into microstructure changes accompanying helium swelling of Be reactor neutron irradiated at 450degC or {alpha}-particles implanted in cyclotron to reach the same volume accumulation of He (6-8 ncm{sup 3} He/cm{sup 3} Be). The microstructures of reactor irradiated and implanted samples were compared after vacuum anneal at 600-800degC up to 50h. The irradiated samples revealed the etchability along the grain boundaries in zones formed by adequately large equilibrium helium pores. The width of the zones increased with the annealing time and after 50h reached 30{mu}. Depleted areas 2-3{mu} dia were observed in some regions of near grain boundary zones. The roles of grain boundaries and manufacturing pores as vacancies` sources and helium sinks are considered. (author)

  15. Tensile properties and microstructure of helium-injected and reactor-irradiated V-20 Ti

    International Nuclear Information System (INIS)

    Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

    1981-01-01

    Mechanical properties and microstructure of vanadium-20% titanium were examined following helium-injection and reactor irradiation. Helium was injected at ambient temperature to concentrations of 90 and 200 at. ppM; neutron irradiation was at 400, 575, 625, and 700 0 C to fluence of 3 x 10 26 n/m 2 , E > 0.1 MeV. Cavities representing negligible volume swelling were observed in all helium-injected specimens. Degradation of mechanical properties, especially loss of ductility due to helium, occurred at temperatures of 625 and 700 0 C. The levels of helium produced in the fusion spectrum can be expected to alter the response of vanadium alloys from that observed in fast reactor irradiations

  16. Tensile properties and microstructure of helium-injected and reactor-irradiated V-20 Ti

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

    1981-01-01

    Mechanical properties and microstructure of vanadium-20% titanium were examined following helium-injection and reactor irradiation. Helium was injected at ambient temperature to concentrations of 90 and 200 at. ppM; neutron irradiation was at 400, 575, 625, and 700/sup 0/C to fluence of 3 x 10/sup 26/ n/m/sup 2/, E > 0.1 MeV. Cavities representing negligible volume swelling were observed in all helium-injected specimens. Degradation of mechanical properties, especially loss of ductility due to helium, occurred at temperatures of 625 and 700/sup 0/C. The levels of helium produced in the fusion spectrum can be expected to alter the response of vanadium alloys from that observed in fast reactor irradiations.

  17. Characterisation of the epithermal neutron irradiation facility at the Portuguese research reactor using MCNP.

    Science.gov (United States)

    Beasley, D G; Fernandes, A C; Santos, J P; Ramos, A R; Marques, J G; King, A

    2015-05-01

    The radiation field at the epithermal beamline and irradiation chamber installed at the Portuguese Research Reactor (RPI) at the Campus Tecnológico e Nuclear of Instituto Superior Técnico was characterised in the context of Prompt Gamma Neutron Activation Analysis (PGNAA) applications. Radiographic films, activation foils and thermoluminescence dosimeters were used to measure the neutron fluence and photon dose rates in the irradiation chamber. A fixed-source MCNPX model of the beamline and chamber was developed and compared to measurements in the first step towards planning a new irradiation chamber. The high photon background from the reactor results in the saturation of the detector and the current facility configuration yields an intrinsic insensitivity to various elements of interest for PGNAA. These will be addressed in future developments. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. Study on the behavior of irradiated light water reactor fuel during out-of-pile annealing

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Kanazawa, Hiroyuki; Uno, Hisao; Sasajima, Hideo

    1988-11-01

    Using the pre-irradiated light water reactor fuel (burnup: 35 MWd/kgU) and the slightly irradiated NSRR fuel (burnup: 5.6 x 10 -6 MWd/kgU), FP gas release rate up to the temperature of 2273 K was measured through out-of-pile annealing test. Results of this experiment were compared with those of ORNL annealing test (SFD/HI-test series) performed in USA. Obtained conclusions are: (1) Maximum release rate of Kr gas in light water reactor fuel was 6.4 % min -1 at temperature of 2273 K. This was in good agreement with ORNL data. FP gas release rate during annealing test was increased greatly with increasing fuel burnup and annealing temperature. (2) No FP was detected in NSRR slightly irradiated fuel up to the temperature of 1913 K. (author)

  19. Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

    International Nuclear Information System (INIS)

    Tachibana, Toshimichi

    1985-01-01

    The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

  20. Optimization of reactor pressure vessel internals segmentation in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

    2017-11-15

    One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

  1. International Reactor Physics Experiment Evaluation (IRPhE) Project

    International Nuclear Information System (INIS)

    2013-01-01

    The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  3. Aging management of PWR reactor internals in U.S. plants

    International Nuclear Information System (INIS)

    Amberge, K.J.; Demma, A.

    2015-01-01

    This paper describes the development, aging management strategies and inspection results of the Pressurized Water Reactor (PWR) vessel internals inspection and evaluation guidelines. The goal of these guidelines is to provide PWR owners with robust aging management strategies to monitor degradation of internals components to support life extension as well as the current period of operation and power up-rate activities. The implementation of these guidelines began in 2010 within the U.S. PWR fleet and several examinations have been performed since. Examples of inspection results are presented for selected vessel internals components and are compared with simulation results. In summary, to date there have been no observations of austenitic stainless steel stress corrosion cracking (SCC), which is consistent with expectations based on the current understanding of the mechanism. Observations of irradiation assisted stress corrosion cracking (IASCC) have been limited and only found in baffle former bolting. Additionally, no macroscopic effects or global observations of void swelling impacts on general conditions of reactor internal hardware have been observed. (authors)

  4. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for Kijang research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Man; Tahk, Young Wook; Jeong, Yong Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); and others

    2017-08-15

    The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm{sup 3}, was selected to achieve higher fuel efficiency and performance than are possible when using U{sub 3}Si{sub 2}/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm{sup 3}), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  5. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for kijang research reactor

    Directory of Open Access Journals (Sweden)

    Jong Man Park

    2017-08-01

    Full Text Available The construction project of the Kijang research reactor (KJRR, which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm3, was selected to achieve higher fuel efficiency and performance than are possible when using U3Si2/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm3, were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  6. International benchmark on the natural convection test in Phenix reactor

    International Nuclear Information System (INIS)

    Tenchine, D.; Pialla, D.; Fanning, T.H.; Thomas, J.W.; Chellapandi, P.; Shvetsov, Y.; Maas, L.; Jeong, H.-Y.; Mikityuk, K.; Chenu, A.; Mochizuki, H.; Monti, S.

    2013-01-01

    Highlights: ► Phenix main characteristics, instrumentation and natural convection test are described. ► “Blind” calculations and post-test calculations from all the participants to the benchmark are compared to reactor data. ► Lessons learned from the natural convection test and the associated calculations are discussed. -- Abstract: The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs

  7. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    Energy Technology Data Exchange (ETDEWEB)

    Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

    2013-07-01

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

  8. Seed irradiation facilities at TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Najzer, M.

    1972-01-01

    Fast neutrons and gamma-rays with their high and low LET respectively are excellent complementary tools for investigation of the effect of different types of mutations. TRIGA Irradiation Facility and Thermal Column Irradiation Facility were designed and installed for the first time in the TRIGA tank and thermal column respectively. The basic idea of design was the use of depleted uranium as gamma-ray and thermal neutron shield and simultaneously as thermal to fast neutron converter. Low LET radiation, due to direct and thermal neutron capture gamma-rays, is strongly attenuated while fast neutron flux is increased. GIF is made of a cadmium tube inserted in a graphite block. It is located in the central thermal column channel. The basic idea is to convert thermal neutrons to gamma-rays by capture in the cadmium

  9. Database of Samples Irradiated at Reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Muhd Husamuddin Abdul Khalil; Mohd Amin Sharifuldin Salleh; Julia Abdul Karim

    2011-01-01

    Evaluation has been made to data of irradiated samples for the type of sample requested for activation at RTP. Sample types are grouped with percentage of total throughputs to rule out the weight percent of every respective group. The database consists of radionuclide inventory of short and long half-life and high activity radionuclides such as Br and Au have been identified and that database has been constructed using a user-friendly Microsoft Access. Through this, trend of gamma exposure will easily be evaluated at experimental facilities and could ensure radiological effect towards safety and health is limited per Radiation Protection (Basic Safety Standard) Regulation 1988. This database places an important parameter to improve management system in acquiring information of the samples irradiated at RTP and will enhance the safety assurance and reliability of the experimental design basis. (author)

  10. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Hunn, John D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Ploger, Scott A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Morris, Robert N.; Baldwin, Charles A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Harp, Jason M.; Winston, Philip L. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Gerczak, Tyler J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Rooyen, Isabella J. van [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States)

    2016-09-15

    Highlights: • Post-irradiation examination was performed on AGR-1 coated particle fuel. • Cesium release from the particles was very low in the absence of failed SiC layers. • Silver release was often substantial, and varied considerably with temperature. • Buffer and IPyC layers were found to play a key role in TRISO coating behavior. • Fission products palladium and silver were found in the SiC layer of particles. - Abstract: The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of {sup 110m}Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10{sup −4} to 5 × 10{sup −4} for {sup 154}Eu and 8 × 10{sup −7} to 3 × 10{sup −5} for {sup 90}Sr. The average {sup 134}Cs fractional release from compacts was <3 × 10{sup −6} when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10{sup 5} in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving {sup 134}Cs fractional release in two capsules to approximately 10{sup −5}. Identification and characterization of these particles has provided unprecedented insight into

  11. Neutron irradiation embrittlement of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Steele, L.E.

    1975-01-01

    The reliability of nuclear power plants depends on the proper functioning of complex components over the whole life on the plant. Particular concern for reliability is directed to the primary pressure boundary. This report focuses on the portion of the primary system exposed to and significantly affected by neutron radiation. Experimental evidence from research programmes and from reactor surveillance programmes has indicated radiation embrittlement of a magnitude sufficient to raise doubts about reactor pressure vessel integrity. The crucial nature of the primary vessel function heightens the need to be alert to this problem, to which, fortunately, there are positive aspects: for example, steels have been developed which are relatively immune to radiation embrittlement. Further, awareness of such embrittlement has led to designs which can accomodate this factor. The nature of nuclear reactors, of the steels used in their construction, and of the procedures for interpreting embrittlement and minimizing the effects are reviewed with reference to the reactors that are expected to play a major role in electric power production from now to about the turn of the century. The report is intended as a manual or guidebook; the aim has been to make each chapter or major sub-division sufficiently comprehensive and self-contained for it to be understood and read independently of the rest of the book. At the same time, it is hoped that the whole is unified enough to make a complete reading useful and interesting to the several classes of reader that are involved with only specific aspects of the topic

  12. Microstructural stability of fast reactor irradiated 10 to 12% Cr ferritic-martensitic stainless steels

    International Nuclear Information System (INIS)

    Little, E.A.; Stoter, L.P.

    1982-01-01

    The strength and microstructural stability of three 10 to 12% Cr ferritic-martensitic stainless steels have been characterized following fast reactor irradiation to damage levels of 30 displacements per atom (dpa) at temperatures in the range 380 to 615 0 C. Irradiation results in either increases or decreases in room temperature hardness depending on the irradiation temperature. These strength changes can be qualitatively rationalized in terms of the combined effects of irradiation-induced interstitial dislocation loop formation and recovery of the dislocation networks comprising the initial tempered martensite structures. Precipitate evolution in the irradiated steels is associated with the nonequilibrium segregation of the elements nickel, silicon, molybdenum, chromium and phosphorus, brought about by solute-point defect interactions. The principal irradiation-induced precipitates identified are M 6 X, intermetallic chi and sigma phases and also α' (Cr-rich ferrite). The implications of the observed microstructural changes on the selection of martensitic stainless steels for fast reactor wrapper applications are briefly considered

  13. Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons

    Energy Technology Data Exchange (ETDEWEB)

    Stephenson, Kale J., E-mail: kalejs@umich.edu; Was, Gary S.

    2015-01-15

    Highlights: • Dislocation loops were the prominent defect, but neutron irradiation caused higher loop density. • Grain boundaries had similar amounts of radiation-induced segregation. • The increment in hardness and yield stress due to irradiation were very similar. • Relative IASCC susceptibility was nearly identical. • The effect of dislocation channel step height on IASCC was similar. - Abstract: The objective of this study was to compare the microstructures, microchemistry, hardening, susceptibility to IASCC initiation, and deformation behavior resulting from proton or reactor irradiation. Two commercial purity and six high purity austenitic stainless steels with various solute element additions were compared. Samples of each alloy were irradiated in the BOR-60 fast reactor at 320 °C to doses between approximately 4 and 12 dpa or by a 3.2 MeV proton beam at 360 °C to a dose of 5.5 dpa. Irradiated microstructures consisted mainly of dislocation loops, which were similar in size but lower in density after proton irradiation. Both irradiation types resulted in the formation of Ni–Si rich precipitates in a high purity alloy with added Si, but several other high purity neutron irradiated alloys showed precipitation that was not observed after proton irradiation, likely due to their higher irradiation dose. Low densities of small voids were observed in several high purity proton irradiated alloys, and even lower densities in neutron irradiated alloys, implying void nucleation was in process. Elemental segregation at grain boundaries was very similar after each irradiation type. Constant extension rate tensile experiments on the alloys in simulated light water reactor environments showed excellent agreement in terms of the relative amounts of intergranular cracking, and an analysis of localized deformation after straining showed a similar response of cracking to surface step height after both irradiation types. Overall, excellent agreement was observed

  14. Introduction [International Reactor Dosimetry File 2002 (IRDF-2002)

    International Nuclear Information System (INIS)

    Paviotti-Corcuera, R.; Zolnay, E.M.

    2006-01-01

    The most recently tested version of the International Reactor Dosimetry File, IRDF-90 Version 2 (IRDF-90.2), was released in 1993. Most of the evaluations used in this file were prepared in the mid-1980s, and in the meantime a large amount of new experimental data has become available, along with two new national reactor dosimetry libraries (the Russian Reactor Dosimetry File (RRDF-98) and the Japanese Evaluated Nuclear Data Library (JENDL/D-99)). The cross-sections and related uncertainties for several reactions in these libraries may be of better quality than the data in the older IRDF-90 file. These developments have resulted in different cross-section values being applied to the evaluation of experimental data, creating difficulties in comparing the results of reactor dosimetry calculations from the same types of nuclear facility. Therefore, there has been a strong demand from the reactor dosimetry community for an updated and standardized version of the IRDF. The IAEA has in the past supported similar efforts to improve the quality of data for reactor dosimetry applications. A major objective of the present data development project was to prepare and distribute a standardized, updated and tested reactor dosimetry cross-section library accompanied by uncertainty information (IRDF-2002) for use in service life assessments of nuclear power reactors. In order to achieve this objective, two technical meetings were organized. Both meetings were held at the IAEA in Vienna. The first meeting took place from 27 to 29 August 2002, the second from 1 to 3 October 2003. Recommendations were made concerning the following topics and the preparation of the library: reactions to be included, requirements for new evaluations or revisions, nuclear decay data, radiation damage data, testing of the data in benchmark fields and inclusion of computer codes. The participants emphasized that good quality nuclear data for reactor dosimetry are essential to improve assessments of the

  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. Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

    2003-01-01

    Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

  17. Onsite nondestructive examination techniques for irradiated water-cooled power reactor fuel

    International Nuclear Information System (INIS)

    1981-03-01

    The International Atomic Energy Agency, in response to the recommendations from several Member States, has prepared this Guidebook on Onsite Non-Destructive Techniques for Irradiated Water-Cooled Power Reactor Fuel with the assistance of a number of experts and organizations in this field. During the preparation of this report it became evident that a comparison between different techniques is a most difficult task and depends on a number of factors related to fuel design, plant characteristics and operating conditions. Consequently the emphasis of the report is on the survey of different techniques presently available. It is also to be noted that because the degree of development for any given technique varies significantly among organizations, it is understood that the report should not be used as consensus standard of the minimum capabilities for each class of techniques, nor does it give recommendations in the regulatory sense. Furthermore, the inclusion of some commercial pieces of equipment, services and other products are for illustrative purposes only and neither implies any preference by the Agency nor can the Agency be liable for any material presented in the report

  18. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels (Final Report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J. H.; Lee, B. S.; Chi, S. H.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Kwon, S. C.; Park, D. G.; Kang, Y. H.; Choo, K. N.; Oh, J. M.; Park, S. J.; Kim, B. K.; Shin, Y. T.; Cho, M. S.; Sohn, J. M.; Kim, D. S.; Choo, Y. S.; Ahn, S. B.; Oh, W. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-05-01

    Reactor pressure vessel materials, which were produced by a domestic company, Doosan Heavy Industries and construction Co., Ltd., have been evaluated using the neutron irradiation facility HANARO. For this evaluation, instrumented capsules were used for neutron irradiation of various kinds of specimens made of different heats of steels, which are VCD(Y4), VCD+Al(U4), Si+Al(Y5), U4 weld metal, and U4 HAZ, respectively. The fast neutron fluence levels ranged 1 to 5 (x10{sup 19} n/cm{sup 2}, E>1MeV) depending on the specimens and the irradiation temperature was controlled within 290{+-}10 deg C. The test results showed that, in the ranking of the material properties of the base metals, both before and after neutron irradiation, Y5 is the best, U4 the next and Y4 the last. Y4 showed a substantial change by neutron irradiation as well as the properties was worse than others in the unirradiated state. However, Y5, which showed the best properties in unirradiated state, was also the best in the resistance for irradiation embrittlement and one can hardly detect the property change after irradiation. The weldment showed a reasonably good resistance to irradiation embrittlement while the unirradiated properties were worse than base metals. The RPV steels are all expected to meet the screening criteria of the USNRC codes and regulations during the end of plant life. 39 refs., 42 figs., 27 tabs. (Author)

  19. Irradiated stainless steel material constitutive model for use in the performance evaluation of PWR pressure vessel internals

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, J.Y.; Dunham, R.S. [ANATECH (United States); Demma, A. [Electric Power Research Institute - EPRI (United States)

    2011-07-01

    Demonstration of component functionality requires analytical simulations of reactor internals behavior. Towards that aim, EPRI has undertaken the development of irradiated material constitutive model and damage criteria for use in global and local finite-element based functionality analysis methodology. The constitutive behavioral regimes of irradiated stainless steel types 316 and 304 materials included in the model consist of: elastic-plastic material response considering irradiation hardening of the stress-strain curve, irradiation creep, stress relaxation, and void swelling. IASCC and degradation of ductility with irradiation are the primary damage mechanisms considered in the model. The material behavior model development consists of two parts: the first part is a user-material subroutine that can interface with a general-purpose finite element computer program to adapt it to the special-purpose of functionality analysis of reactor internals. The second part is a user utility in the form of Excel Spread sheets that permit users to extract a given property, e.g. the elastic-plastic stress-strain curve, creep curve, or void-swelling curve, as function of the relevant independent variables. The development of the model takes full advantage of the significant work that has been undertaken within EPRI's Material Reliability Program (MRP) to improve the knowledge of the material properties of irradiated stainless steels. Data from EPRI's MRP database have been utilized to develop equations that characterize the yield strength, ultimate tensile strength, uniform elongation, total elongation, reduction in area, void swelling and irradiation creep of stainless steels in a PWR environment. It is noted that, while the development of the model's equations has been statistically faithful to the material database, approximations were introduced in the model to ensure appropriate conservatism in the model's application consistently with accepted

  20. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, St.

    2005-01-01

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  1. International project GT-MHR - New generation of nuclear reactors

    International Nuclear Information System (INIS)

    Vasyaev, A.; Kodochigov, N.; Kuzavkov, N.; Kuznetsov, L.

    2001-01-01

    Gas turbine-modular helium reactor (GT-MHR) is the reactor of new generation, which satisfies the requirements of the progressing large-scale nuclear power engineering. The activities in GT-MHR Project started in 1995. In 1997 the Conceptual Design was developed under four-side Agreement (MINATOM, General Atomics, FRAMATOME, Fuji Electric); it has passed through the internal and international reviews, has been approved and recommended for further development as one of new trends in creation of new generation plants. Starting from 1999, the activities in the development of the Preliminary Design of the plant were deployed under the Agreement between the Government of the United States of America and the Government of the Russian Federation on Scientific and Technical Cooperation in the Management of Plutonium That Has Been Withdrawn From Nuclear Military Programs dated July 24, 1998. The activities are established under the Contract between MINATOM and OKBM Russia, and under the General Agreement between Department of Energy (DOE), USA and OKBM. The GT-MHR Project is included into 'Development Strategy of Russian Nuclear Power in the first Half of the XXI-st Century' providing for 'the participation in an international project on the development and construction of GT-MHR nuclear power plant till year 2010 and 'operation of GT-MHR prototype unit and creation of fuel fabrication facility (within framework of International Project) till year 2030'. (author)

  2. Cutting technique for reactor internals by laser beam

    International Nuclear Information System (INIS)

    Matsumoto, O.; Sugihara, M.; Matsuda, K.; Miya, K.

    1990-01-01

    At present in Japan the verification tests on the commercial nuclear power reactor decommissioning technology are being conducted as the project of The Ministry of International Trade and Industry by Nuclear Power Engineering Test Center. This paper summarizes the interim results of the verification test for the reactor core internals decommissioning technology, which is being conducted from 1986 as a theme of the above project. All core internals to be studied here are made of stainless steel, and the maximum wall thickness is about 500mm (the maximum one to be cut is about 300mm) for the PWR's, and about 100mm for the BWR's. Though the plasma cutting, arc saw cutting method, etc. have been studied u p to now as the cutting technology for decommissioning these core internals, the authors are carrying out the development and verification test of the cutting technology with the laser beam, which is expected to increase its power in future and can be applied to various materials

  3. Irradiation effects on reactor structural materials. Semi-annual progress report, August 1974--February 1975

    International Nuclear Information System (INIS)

    Claudson, T.T.

    1975-03-01

    Data are reported on: effects of cold work on creep-fatigue of irradiated 304 and 316 stainless steel (ss); swelling of 304 and 316 ss irradiated with protons and fast neutrons; effects of hold time on fatigue crack propagation in neutron-irradiated 20 percent cold-worked 316 ss; radiation resistance of 0.03 percent Cu A533-B steel; microstructure of irradiated Inconel 718, Incoloy 800, PH13-8Mo, Mo, and Nb; dose dependence of 2.8-MeV Ni + ion damage (swelling) in Ni; notch ductility and strength of 316 ss submerged arc weld deposits; effects of microstructure of 316 ss on its irradiation response; in-reactor deformation of 20 percent cold-worked 316 ss; microstructure of HFIR-irradiated 316 ss; void microstructures of V bombarded by 46-MeV Ni 6+ ions (with and without preinjected helium) or 7.5-MeV Ta 3+ ions; swelling of Mo, Mo--0.5 Ti, Nb, Nb--1 Zr, W, and W--25 Re after fast neutron irradiation; swelling of V ion-irradiated Mo; creep of 20 percent cold-worked 316 ss at 850, 1000, and 1100 0 F; effects of fast neutrons on mechanical properties of 20 percent cold-worked 316 ss; notch effects in tensile behavior of irradiated, annealed 304 ss (EBR-II duct thimbles); equations for thermal creep in pressurized tubes of 20 percent cold-worked 316 ss; irradiation creep in cold-worked 316 ss; helium production cross sections in neutron-irradiated elements; and radiation effects on various alloys. (U.S.)

  4. Inspection and repair of reactor pressure vessel (RPV) internals

    International Nuclear Information System (INIS)

    Bohmann, W.; Poetz, F.; Nicolai, M.

    1996-01-01

    The past 10 years of operation of light water reactors were characterized by intensive inspection- and repair work on vital components. For boiling water reactors (BWR) it was typical to totally replace the piping system and for pressurized water reactors (PWR) it was the step to complete steam generator (SG) replacement - besides the development of increasingly diligent inspection and repair methods for SG tubes. It can be expected that in the 10 years to come the development of inspection- and repair methods will be aimed mainly at the core internals of BWR's as well as PWR's. Our prediction is that before the end of this decade a first complete replacement of these components will be performed. Already to date a broad range of techniques are available which enable the utilities to carry out inspections and repair of components of core internals in a relatively short time and acceptable expenses. Using examples such as Fuel Alignment Pin Inspection and Replacement, Baffle Former Bolt Inspection and Replacement, Core Barrel Former Bolt Inspection which are typical for PWR's we will in the following describe the existing methods, their development and - last but not least - their successful utilization. What is going to happen in the future? Ageing of the operating plants will continue, thus requesting the plant operators as well as the service companies to work on advanced technologies to fulfill the needs of the industry. (author)

  5. A description of the Canadian irradiation-research facility proposed to replace the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, A G; Lidstone, R F; Bishop, W E; Talbot, E F; McIlwain, H [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-31

    To replace the aging NRU reactor, AECL has developed the concept for a dual-purpose national Irradiation Research Facility (IRF) that tests fuel and materials for CANDU (CANada Deuterium Uranium) reactors and performs materials research using extracted neutron beams. The IRF includes a MAPLE reactor in a containment building, experimental facilities, and support facilities. At a nominal reactor power of 40 MW{sub t}, the IRF will generate powers up to 1 MW in natural-uranium CANDU bundles, fast-neutron fluxes up to 1.4 x 10{sup 18} n{center_dot}m{sup -2}{center_dot}s{sup -1} in Zr-alloy specimens, and thermal-neutron fluxes matching those available to the NRU beam tubes. (author). 9 refs., 5 tabs., 2 figs.

  6. A study on the irradiation effect of reactor materials using a cyclotron

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Joon Hwa; Ji, Se Hwan; Kang, Yung Hwan; Park, Duk Keun; Park, Jong Man; Lee, Bong Sang; Oh, Jong Myung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-12-01

    The objectives of the present study are to develop the simulation techniques of neutron irradiation through ion irradiation using a Cyclotron and small specimen techniques and to evaluate radiation effects of reactor materials. Effects of proton or neutron irradiation on domestic 12Cr-1MoV and SA508-3 steels were evaluated by small scale specimen test techniques, i.e, small punch and miniaturized tensile test. In order to study the radiation damage mechanism, irradiation effects of the steels were investigated by means of property change tests such as microstructure, physical and thermal properties. Feasibility study on application of a magnetic non-destructive methods to evaluate radiation effects on RPV materials was performed. 109 figs, 12 tabs, 102 refs. (Author).

  7. A study on the irradiation effect of reactor materials using a cyclotron

    International Nuclear Information System (INIS)

    Hong, Joon Hwa; Ji, Se Hwan; Kang, Yung Hwan; Park, Duk Keun; Park, Jong Man; Lee, Bong Sang; Oh, Jong Myung

    1994-12-01

    The objectives of the present study are to develop the simulation techniques of neutron irradiation through ion irradiation using a Cyclotron and small specimen techniques and to evaluate radiation effects of reactor materials. Effects of proton or neutron irradiation on domestic 12Cr-1MoV and SA508-3 steels were evaluated by small scale specimen test techniques, i.e, small punch and miniaturized tensile test. In order to study the radiation damage mechanism, irradiation effects of the steels were investigated by means of property change tests such as microstructure, physical and thermal properties. Feasibility study on application of a magnetic non-destructive methods to evaluate radiation effects on RPV materials was performed. 109 figs, 12 tabs, 102 refs. (Author)

  8. Interfacial degradation of organic composite material by irradiation in reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nishijima, Shigehiro; Nishiura, Tetsuya; Okada, Toichi [Osaka Univ., Ibaraki (Japan). Inst. of Scientific and Industrial Research

    1996-04-01

    Glass fiber reinforced plastics (GFRP) with many kinds of matrix resins were made of E glass treated with silane as the reinforced material. Degradation of shearing strength of GFRP irradiated at low temperature was determined. It was clear from the results of comparing the degradation process with the fractured surface that the degradation was very affected by the radiation resistance of the bonded part between resin and coupling agents. It means that we had to be careful in the choice of interfacial treatments and epoxy matrices corresponded to it. (S.Y.)

  9. Evaluation of the french test reactors irradiation embrittlement experiments

    International Nuclear Information System (INIS)

    Miannay, D.; Dussarte, D.; Soulat, P.

    1988-07-01

    The shifts of CV 41J energy index temperatures due to irradiation measured in France have been stored in a data bank and are analysed. According to a simple physically based model which is here-after verified, correlations are proposed for Base Metal (BM) and Weld Metal (WM). The achemical and phosphorus components of the chemical factor are equivalent. However, nickel and copper play a leading part in BM and WM respectively. The copper nickel interaction is not evident. These correlations are for cleavage fracture and not for intergranular fracture. This work is subject to revision and extension

  10. Accurate analysis of pins irradiated in fast neutron reactors

    International Nuclear Information System (INIS)

    Regnaud, F.; Guichard, C.

    1976-01-01

    A device was developed for the chemical analysis of fissile materials in a whole fuel pin in order to determine U and Pu with an accuracy of about 0.1% in samples from 100 to 200g, irradiated at burn-up between 50 and 100.000MWd/t. This unit is provided with two shielded α hot cells. A dissolving device was developed which enables the dissolution of residues. Pu is determined by oxydation-reduction titration and U by the isotopic dilution method [fr

  11. The behavior of 131I in polymetatelluric acid irradiated in the nuclear reactor

    International Nuclear Information System (INIS)

    Teofilovski, C.

    1966-01-01

    Polymetarelluric acid, whose composition is (H 2 TeO 4 ) n , is successfully used at he Institute as a target for obtaining 131 I in the reactor. It is prepared by hearing orthotelluric acid in air at 160 deg C of in a steam of water vapor at 208 deg C. Analysis of the valency states of 131 I in irradiated (H 2 TeO 4 ) n prepared in either of the above ways shows a variable ratio of reduced and oxidized forms. A considerable increase of the reduced forms with increasing integral thermal neutron flux during irradiation in the reactor in the given interval has also been observed. In order to explain the above phenomenon (H 2 TeO 4 ) n was irradiated in the reactor under different conditions, with measurement of the wall temperature of the quartz ampoules containing the target material. Yields of reduced and oxidized form of 131 I were determined immediately after irradiation and after annealing of the target at temperatures from 60 deg C to 150 deg C. A considerable decrease in the yield of the reduced forms of 131 I on target annealing above 100 deg C was observed (author)

  12. The behavior of {sup 131}I in polymetatelluric acid irradiated in the nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Teofilovski, C [Institute of Nuclear Sciences Boris Kidric, Hot Laboratory Department, Vinca, Beograd (Serbia and Montenegro)

    1966-01-15

    Polymetarelluric acid, whose composition is (H{sub 2}TeO{sub 4}){sub n}, is successfully used at the Institute as a target for obtaining {sup 131}I in the reactor. It is prepared by hearing orthotelluric acid in air at 160 deg C of in a steam of water vapor at 208 deg C. Analysis of the valency states of {sup 131}I in irradiated (H{sub 2}TeO{sub 4}){sub n} prepared in either of the above ways shows a variable ratio of reduced and oxidized forms. A considerable increase of the reduced forms with increasing integral thermal neutron flux during irradiation in the reactor in the given interval has also been observed. In order to explain the above phenomenon (H{sub 2}TeO{sub 4}){sub n} was irradiated in the reactor under different conditions, with measurement of the wall temperature of the quartz ampoules containing the target material. Yields of reduced and oxidized form of {sup 131}I were determined immediately after irradiation and after annealing of the target at temperatures from 60 deg C to 150 deg C. A considerable decrease in the yield of the reduced forms of {sup 131}I on target annealing above 100 deg C was observed (author)

  13. Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

    International Nuclear Information System (INIS)

    Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

    1985-01-01

    The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

  14. 8th International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

    Leotta, G G; Muon-catalyzed fusion and fusion with polarized nuclei

    1988-01-01

    The International School of Fusion Reactor Technology started its courses 15 years ago and since then has mantained a biennial pace. Generally, each course has developed the subject which was announced in advance at the closing of the previous course. The subject to which the present proceedings refer was chosen in violation of that rule so as to satisfy the recent and diffuse interest in cold fusion among the main European laboratories involved in controlled thermonuclear research (CTR). In the second half of 1986 we started to prepare a workshop aimed at assessing the state of the art and possibly of the perspectives of muon- catalyzed fusion. Research in this field has recently produced exciting experimental results open to important practical applications. We thought it worthwhile to consider also the beneficial effects and problems of the polarization ofthe nuclei in both cold and thermonuclear fusion. In preparing the 8th Course on Fusion Reactor Technology, it was necessary to abandon the tradi...

  15. Magnet systems for the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Miller, J.R.

    1988-01-01

    The definition phase for the International Thermonuclear Experimental Reactor (ITER) has been nearly completed, thus beginning a three-year design effort by teams from the European Community (EC), Japan, US, and USSR. Preliminary parameters for the superconducting magnet system have been established to guide more detailed design work. Radiation tolerance of the superconductors and insulators has been important because it sets requirements for the neutron-shield dimension and sensitively influences reactor size. Major levels of mechanical stress appear in the structural cases of the inboard legs of the toroidal-field (TF) coils. The winding packs of the TF coils include significant fractions of steel that provide support against in-plane separating loads, but they offer little support against out-of-plane loads unless shear-bonding of the conductors can be maintained. Heat removal from nuclear and ac loads has not limited the fundamental design, but it has nonnegligible economic consequences. 3 refs., 3 figs., 5 tabs

  16. Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Kristina, E-mail: kristina.lindgren@chalmers.se [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Boåsen, Magnus [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Stiller, Krystyna [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Vattenfall Ringhals AB, SE-430 22 Väröbacka (Sweden); Thuvander, Mattias [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden)

    2017-05-15

    Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58%) low Cu (0.04%) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance material has a lower cluster number density, but larger clusters. The resulting impact on the mechanical properties of these two effects cancel out, resulting in a measured hardness that seems to be on the same trend as the high flux material. The dispersed barrier hardening model with an obstacle strength factor of 0.15 was found to reproduce the increase in hardness. In the investigated high flux materials, the clusters' Cu content was higher. - Highlights: •Clustering in a low Cu, high Ni reactor pressure vessel steel weld is studied. •The clusters nucleate and grow during irradiation, and consist of Ni, Mn, Si, and Cu. •High flux neutron irradiated material is compared to surveillance material. •High flux was found to result in smaller clusters with a larger number density. •Hardness follows the same dependence on fluence, independent of flux.

  17. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    International Nuclear Information System (INIS)

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-01-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  18. Advance of investigation of irradiation embrittlement mechanism of nuclear reactor pressure vessel steels. History and future of irradiation embrittlement researches

    International Nuclear Information System (INIS)

    Ishino, Shiori

    2007-01-01

    The nuclear reactor pressure vessel is the most important component of LWR plants required to be safe. This paper describes contents of the title consisting of four chapters. The first chapter states the general theory of irradiation effects, irradiation embrittlement and decreasing of toughness, and some kinds of pressure vessel steels. The second chapter explains history of irradiation embrittlement investigations and the advance of research methods for experiments and calculation. The third chapter contains information of inner structure of irradiated materials and development of prediction equations, recent information of embrittlement mechanism and mechanism guided prediction method, USA model and Central Research Institute of Electric Power Industry (CRIEPI) model. The fourth chapter states recent problems from viewpoints of experimental and analytical approaches. Comparison of standards of LWR pressure vessel steels between Japan and USA, relation between the density of number of cluster and the copper content, effect of flux on clustering of copper atoms, and CRIEPI's way of approaching the prediction method are illustrated. (S.Y.)

  19. The irradiation creep in reactor graphites for HTR applications

    International Nuclear Information System (INIS)

    Veringa, H.J.; Blackstone, R.

    1976-01-01

    A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400 0 C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400 0 C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density. (author)

  20. Study by internal friction of curing low temperature irradiation defects in graphite

    International Nuclear Information System (INIS)

    Rouby, Dominique.

    1974-01-01

    Micromechanical properties and anelastic effects of neutrons irradiated graphites at 300 and 77 0 K are investigated by internal friction analysis and elasticity modulus variations. Defects created by irradiation are studied and evolution versus dose and annealing is followed [fr

  1. Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E.

    1995-01-01

    The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab

  2. Study about internal friction in deformed - and irradiated pure titanium

    International Nuclear Information System (INIS)

    Miyada, L.T.

    1979-01-01

    Internal friction and modulus are measured in pure Ti at low temperature using an inverted torsion-pendulum at about 1 Hz. The presence of four relaxation peaks P' sub(d)(-140 0 C), P sub(d)(-101 0 C), P' sub(α)(-75 0 C) and P sub(α)(-50 0 C) has been found, and effects of plastic deformation, heat treatment and neutron irradiation on these peaks are investigated in detail. Activation energies and frequency factors of P sub(d) and Pα peaks are consistent with the data in higher frequency range reproted by other workers. The P sub(d) and P' sub(d) peaks grow after deformation and tend to decay after annealing at high temperatures or after neutron irradiation. Both peaks are resonably interpreted in terms of dislocation relaxation mechanisms (Bordoni type) arising from thermally activated motion of dislocations in different slip planes of h.c.f. structure. Peierls stress of dislocations giving rise to each peak have calculated based on Seeger's theory, and found to be consistent with that of f.c.c. metals. On the other hand, P sub(α) and P' sub(α) peaks grow significantly at the expense of P sub(d) and P' sub(d) peaks after neutron irradiation in deformed samples. The behaviour of these peaks as a function of irradiation dose and annealing temperatures strongly indicated that they are due to relaxations resulting from dislocations-point defects interactions (Hasiguti type). It is tentatively suggested that P sub(α) and P' sub(α) peaks are related with interactions of dislocations with divacancies and single vacancies, respectively. Application of Schiller's model showed a consistent result with regard to the P' sub(α) peak experimentally observed. (Author) [pt

  3. International centres of excellence based on research reactors

    International Nuclear Information System (INIS)

    Alldred, K.; Tozser, S.M.; Adelfang, P.

    2013-01-01

    A number of high flux research reactors were, or will be constructed. Each of these high flux facilities has the potential to be an important regional or International Centre of Excellence based on Research Reactors (ICERR) and scientific hub for research and materials investigations. Some are so organized currently, but for many there is a strongly national focus and scope for a significant expansion of their international role. There are manifold benefits of an expanded international role both for the ICERR's themselves and for the institutes that affiliate with them. These benefits include increased utilization and financial stability, increased international prestige, and enhanced scientific resources and capabilities. There are significant hurdles to obtaining the benefits from an expanded international role. For example, to achieve its full potential an ICERR must accommodate scientists from other nations, and include the plans and aspirations of the international community in the ICERR governance. The ICERR must also fully meet the national responsibilities for safety and security. Balancing these potentially conflicting requirements and finding a path through the organisational and legal issues is a significant challenge for any institute. The existing ICERR's therefore provide important case studies and examples of best practice that could inform the actions of other potential ICERR's. This paper describes an IAEA initiative to encourage and support the formation of new ICERR's, strengthen existing ones, and increase training resources available to Member States. The initiative will seek to share best practice and facilitate meetings and technical exchanges between the existing and potential ICERRs, and between the potential ICERR's and potential subscribing or affiliating institutes. (orig.)

  4. International Centers of Excellence based on Research Reactors

    International Nuclear Information System (INIS)

    Alldred, K.; Tozser, S. M.; Adelfang, P.

    2012-01-01

    A number of high flux research reactors were, or will be constructed. Each of these high flux facilities has the potential to be an important regional or International Centre of Excellence based on Research Reactors (ICERR) and scientific hub for research and materials investigations. Some are so organized currently, but for many there is a strongly national focus and scope for a significant expansion of their international role. There are manifold benefits of an expanded international role both for the ICERR's themselves and for the institutes that affiliate with them. These benefits include increased utilization and financial stability, increased international prestige, and enhanced scientific resources and capabilities. There are significant hurdles to obtaining the benefits from an expanded international role. For example, to achieve its full potential an ICERR must accommodate scientists from other nations, and include the plans and aspirations of the international community in the ICERR governance. The ICERR must also fully meet the national responsibilities for safety and security. Balancing these potentially conflicting requirements and finding a path through the organisational and legal issues is a significant challenge for any institute. The existing ICERR's therefore provide important case studies and examples of best practice that could inform the actions of other potential ICERR's. This paper describes an IAEA initiative to encourage and support the formation of new ICERR's, strengthen existing ones, and increase training resources available to Member States. The initiative will seek to share best practice and facilitate meetings and technical exchanges between the existing and potential ICERRs, and between the potential ICERR's and potential subscribing or affiliating institutes. (authors)

  5. Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor

    International Nuclear Information System (INIS)

    Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon

    2005-01-01

    A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams

  6. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Sorescu, Ion [Institute for Nuclear Research, Pitesti (Romania). TRIGA Reactor Loop Facility; Parvan, Marcel [Institute for Nuclear Research, Pitesti (Romania). Hot Cells Lab.

    2012-12-15

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO{sub 2} grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

  7. Treatment of opium alkaloid containing wastewater in sequencing batch reactor (SBR)-Effect of gamma irradiation

    International Nuclear Information System (INIS)

    Bural, Cavit B.; Demirer, Goksel N.; Kantoglu, Omer; Dilek, Filiz B.

    2010-01-01

    Aerobic biological treatment of opium alkaloid containing wastewater as well as the effect of gamma irradiation as pre-treatment was investigated. Biodegradability of raw wastewater was assessed in aerobic batch reactors and was found highly biodegradable (83-90% degradation). The effect of irradiation (40 and 140 kGy) on biodegradability was also evaluated in terms of BOD 5 /COD values and results revealed that irradiation imparted no further enhancement in the biodegradability. Despite the highly biodegradable nature of wastewater, further experiments in sequencing batch reactors (SBR) revealed that the treatment operation was not possible due to sludge settleability problem observed beyond an influent COD value of 2000 mg dm -3 . Possible reasons for this problem were investigated, and the high molecular weight, large size and aromatic structure of the organic pollutants present in wastewater was thought to contribute to poor settleability. Initial efforts to solve this problem by modifying the operational conditions, such as SRT reduction, failed. However, further operational modifications including addition of phosphate buffer cured the settleability problem and influent COD was increased up to 5000 mg dm -3 . Significant COD removal efficiencies (>70%) were obtained in both SBRs fed with original and irradiated wastewaters (by 40 kGy). However, pre-irradiated wastewater provided complete thebain removal and a better settling sludge, which was thought due to degradation of complex structure by radiation application. Degradation of the structure was observed by GC/MS analyses and enhancement in filterability tests.

  8. Treatment of opium alkaloid containing wastewater in sequencing batch reactor (SBR)-Effect of gamma irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Bural, Cavit B.; Demirer, Goksel N. [Middle East Technical University, Department of Environmental Engineering, 06531 Ankara (Turkey); Kantoglu, Omer [Turkish Atomic Energy Authority, Saraykoy Nuclear Research and Training Center, 06982, Kazan, Ankara (Turkey); Dilek, Filiz B., E-mail: fdilek@metu.edu.t [Middle East Technical University, Department of Environmental Engineering, 06531 Ankara (Turkey)

    2010-04-15

    Aerobic biological treatment of opium alkaloid containing wastewater as well as the effect of gamma irradiation as pre-treatment was investigated. Biodegradability of raw wastewater was assessed in aerobic batch reactors and was found highly biodegradable (83-90% degradation). The effect of irradiation (40 and 140 kGy) on biodegradability was also evaluated in terms of BOD{sub 5}/COD values and results revealed that irradiation imparted no further enhancement in the biodegradability. Despite the highly biodegradable nature of wastewater, further experiments in sequencing batch reactors (SBR) revealed that the treatment operation was not possible due to sludge settleability problem observed beyond an influent COD value of 2000 mg dm{sup -3}. Possible reasons for this problem were investigated, and the high molecular weight, large size and aromatic structure of the organic pollutants present in wastewater was thought to contribute to poor settleability. Initial efforts to solve this problem by modifying the operational conditions, such as SRT reduction, failed. However, further operational modifications including addition of phosphate buffer cured the settleability problem and influent COD was increased up to 5000 mg dm{sup -3}. Significant COD removal efficiencies (>70%) were obtained in both SBRs fed with original and irradiated wastewaters (by 40 kGy). However, pre-irradiated wastewater provided complete thebain removal and a better settling sludge, which was thought due to degradation of complex structure by radiation application. Degradation of the structure was observed by GC/MS analyses and enhancement in filterability tests.

  9. Review of recent studies on neutron irradiation embrittlement in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Sudo, Akira; Miyazono, Shohachiro

    1983-06-01

    Recent studies in foreign countries (USA, France, FRG and UK) on neutron irradiation embrittlement have been reviewed. These studies are classified into four areas, such as 1) effect of chemical composition on irradiation embrittlement sensitivity, 2) postirradiation heat treatment for embrittlement relief, 3) fracture toughness evaluation of irradiated materials based on fracture mechanics analysis, and 4) effect of irradiation on fatigue crack propagation behavior. The first area mainly includes the studies related to the effects of copper, phosphorus impurities and nickel alloying and synergistic effect of these components, and furthermore, evaluation of Regulatory Guide 1.99 Rev.l. Studies in the second area show the effects of annealing condition (temperature and time) and metallugical condition on embrittlement relief, and evaluation of periodic annealing in the period of irradiation as a promising method for embrittlement control. Studies in the third area show the correlation between fracture toughness and Cv notch ductility changes with neutron irradiation, and J-R curves of irradiated materials based on the elasto-plastic fracture mechanics. In the forth area, most of studies are investigated in air condition but a few studies in reactor-grade water at high temperature and pressure. (author)

  10. Change in physical properties of high density isotropic graphites irradiated in the ?JOYO? fast reactor

    Science.gov (United States)

    Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

    1995-08-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor "JOYO" to fluences from 2.11 to 2.86 × 10 26 n/m 2 ( E > 0.1 MeV) at temperatures from 549 to 597°C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens.

  11. Change in physical properties of high density isotropic graphites irradiated in the ''JOYO'' fast reactor

    International Nuclear Information System (INIS)

    Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

    1995-01-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor ''JOYO'' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597 C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (orig.)

  12. Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons

    Science.gov (United States)

    Stephenson, Kale J.; Was, Gary S.

    2015-01-01

    The objective of this study was to compare the microstructures, microchemistry, hardening, susceptibility to IASCC initiation, and deformation behavior resulting from proton or reactor irradiation. Two commercial purity and six high purity austenitic stainless steels with various solute element additions were compared. Samples of each alloy were irradiated in the BOR-60 fast reactor at 320 °C to doses between approximately 4 and 12 dpa or by a 3.2 MeV proton beam at 360 °C to a dose of 5.5 dpa. Irradiated microstructures consisted mainly of dislocation loops, which were similar in size but lower in density after proton irradiation. Both irradiation types resulted in the formation of Ni-Si rich precipitates in a high purity alloy with added Si, but several other high purity neutron irradiated alloys showed precipitation that was not observed after proton irradiation, likely due to their higher irradiation dose. Low densities of small voids were observed in several high purity proton irradiated alloys, and even lower densities in neutron irradiated alloys, implying void nucleation was in process. Elemental segregation at grain boundaries was very similar after each irradiation type. Constant extension rate tensile experiments on the alloys in simulated light water reactor environments showed excellent agreement in terms of the relative amounts of intergranular cracking, and an analysis of localized deformation after straining showed a similar response of cracking to surface step height after both irradiation types. Overall, excellent agreement was observed after proton and reactor irradiation, providing additional evidence that proton irradiation is a useful tool for accelerated testing of irradiation effects in austenitic stainless steel.

  13. Facts about food irradiation. A series of fact sheets from the International Consultative Group on Food Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-12-01

    The safety and benefits of foods processed by ionizing radiation are well documented. In an effort to provide governments, especially those of developing countries, with scientifically accurate information on issues of general interest to the public, the International Consultative Group on Food Irradiation (ICGFI), which was established under the aegis of the Food and Agriculture Organization of the United Nations (FAO), the World Health Organization (WHO), and the IAEA, decided at its 7th Annual Meeting in Rome, Italy, on October 1990, to issue a series of ''Fact Sheets'' on the subject. ICGFI, an inter-governmental body with a membership of 37 governments, has as one of its mandates the function to provide information to Member States of the FAO, WHO, and IAEA and to the three organizations themselves on the safe and proper use of food irradiation technology. The Fact Sheets included here cover issues relating to: status and trends; scientific and technical terms; food irradiation and radioactivity; chemical changes in irradiated food; nutritional quality of irradiated foods; genetic studies; microbiological safety of irradiated food; irradiation and food safety; irradiation and food additives and residues; packaging of irradiated foods; safety of irradiation facilities; controlling the process; food irradiation costs; and irradiated foods and the consumer. The Fact Sheets have been separately indexed and included in the INIS Database under Reference Numbers 23011206-23011217, 23011319 and 23012743. The Fact Sheets were first issued by the ICGFI Secretariat (Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture, Vienna, Austria) in May 1991.

  14. Facts about food irradiation. A series of fact sheets from the International Consultative Group on Food Irradiation

    International Nuclear Information System (INIS)

    1991-12-01

    The safety and benefits of foods processed by ionizing radiation are well documented. In an effort to provide governments, especially those of developing countries, with scientifically accurate information on issues of general interest to the public, the International Consultative Group on Food Irradiation (ICGFI), which was established under the aegis of the Food and Agriculture Organization of the United Nations (FAO), the World Health Organization (WHO), and the IAEA, decided at its 7th Annual Meeting in Rome, Italy, on October 1990, to issue a series of ''Fact Sheets'' on the subject. ICGFI, an inter-governmental body with a membership of 37 governments, has as one of its mandates the function to provide information to Member States of the FAO, WHO, and IAEA and to the three organizations themselves on the safe and proper use of food irradiation technology. The Fact Sheets included here cover issues relating to: status and trends; scientific and technical terms; food irradiation and radioactivity; chemical changes in irradiated food; nutritional quality of irradiated foods; genetic studies; microbiological safety of irradiated food; irradiation and food safety; irradiation and food additives and residues; packaging of irradiated foods; safety of irradiation facilities; controlling the process; food irradiation costs; and irradiated foods and the consumer. The Fact Sheets have been separately indexed and included in the INIS Database under Reference Numbers 23011206-23011217, 23011319 and 23012743. The Fact Sheets were first issued by the ICGFI Secretariat (Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture, Vienna, Austria) in May 1991

  15. Status of international cooperation in nuclear technology on testing/research reactors between JAEA and INP-NNC

    International Nuclear Information System (INIS)

    Kawamura, Hiroshi; Tsuchiya, Kunihiko; Takemoto, Noriyuki; Kimura, Akihiro; Tanimoto, Masataka; Izumo, Hironobu; Chakrov, Petr; Gizatulin, Shamil; Chakrova, Yelena; Ludmila, Chkushuina; Asset, Shaimerdenov; Nataliya, Romanova

    2012-02-01

    Based on the implementing arrangement between National Nuclear Center of the Republic of Kazakhstan (NNC) and the Japan Atomic Energy Agency (JAEA) for 'Nuclear Technology on Testing/Research Reactors' in cooperation in Research and Development in Nuclear Energy and Technology, four specific topics of cooperation (STC) have been carried out from June, 2009. Four STCs are as follows; (1) STC No.II-1 : International Standard of Instrumentation. (2) STC No.II-2 : Irradiation Technology of RI Production. (3) STC No.II-3 : Lifetime Expansion of Beryllium Reflector. (4) STC No.II-4 : Irradiation Technology for NTD-Si. The information exchange, personal exchange and cooperation experiments are carried out under these STCs. The status in the field of nuclear technology on testing/research reactors in the implementing arrangement is summarized, and future plans of these specific topics of cooperation are described in this report. (author)

  16. Doping of monocrystalline silicon with phosphorus by means of neutron irradiation at the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Carbonari, A.W.; Puget, M.A.C.

    1990-11-01

    The first neutron irradiation experiments with monocrystal silicon in the IEA-R1 research reactor of IPEN are related. The silicon is irradiated with phosphorus producing a N type semiconductor with a very small resistivity variation throughout the crystal volume. The neutrons induce nuclear reactions in Si-30 isotope and these atoms are then transformed in to phosphorous atoms. This process is known as Neutron Transmutation Doping. In order to irradiate the silicon crystals in the reactor, a specific device has been constructed, and it permits the irradiation of up to 2.5'' diameter monocrystals. (author)

  17. Effect of gamma rays on crystalline materials during irradiation in a reactor

    International Nuclear Information System (INIS)

    Nikolaenko, V.A.; Karpukhin, V.I.; Gordeev, V.G.

    1995-01-01

    The article presents and discusses the results of experiments to determine the effect of gamma rays on the change in the properties of diamond, graphite, and structural steel. The materials were irradiated in a VVER type reactor. For diamonds, the effect on the annealing of defects was investigated. As gamma ray intensity increased, the crystal lattice expansion and defect concentration increased. Graphite lattice expansion and the mechanical properties of structural steel were also examined. Graphite lattice expansion increased with increased neutron flux and decreased irradiation temperature. Changes in the impact toughness of structural steel correlated precisely to the gamma ray flux in the experiments. 6 refs., 3 figs

  18. Design considerations of the irradiation test vehicle for the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  19. Study of optical radiation from SiO2 during reactor irradiation

    International Nuclear Information System (INIS)

    Shikama, Tatsuo; Narui, Minoru; Kakuta, Tsunemi; Kayano, Hideo; Sagawa, Tsutomu; Sanada, Kazuo

    1994-01-01

    Optical radiations were observed from optical fibers under irradiation in a fission reactor, JMTR. A major part of the observed optical radiations is thought to be composed of broad optical radiation in the whole wavelength range studied in the present experiment. This broad optical radiation will be generated by the process of so-called Cherenkov radiation. Also, a sharp optical radiation peak was found at 1270 nm on a F-doped fiber. This peak is thought to relate with doped fluorine ions and ionizing gamma-ray irradiation. ((orig.))

  20. Simple computational modeling for human extracorporeal irradiation using the BNCT facility of the RA-3 Reactor

    International Nuclear Information System (INIS)

    Farias, Ruben; Gonzalez, S.J.; Bellino, A.; Sztenjberg, M.; Pinto, J.; Thorp, Silvia I.; Gadan, M.; Pozzi, Emiliano; Schwint, Amanda E.; Heber, Elisa M.; Trivillin, V.A.; Zarza, Leandro G.; Estryk, Guillermo; Miller, M.; Bortolussi, S.; Soto, M.S.; Nigg, D.W.

    2009-01-01

    We present a simple computational model of the reactor RA-3 developed using Monte Carlo transport code MCNP. The model parameters are adjusted in order to reproduce experimental measured points in air and the source validation is performed in an acrylic phantom. Performance analysis is carried out using computational models of animal extracorporeal irradiation in liver and lung. Analysis is also performed inside a neutron shielded receptacle use for the irradiation of rats with a model of hepatic metastases.The computational model reproduces the experimental behavior in all the analyzed cases with a maximum difference of 10 percent. (author)

  1. Correlation methodology for predicting in-service irradiation embrittlement of reactor pressure vessels

    International Nuclear Information System (INIS)

    Odette, G.R.

    1980-01-01

    Irradiation embrittlement of reactor pressure steels is the consequence of altered microstructure due to both irradiation and time-at-temperature. Relatively poor characterisation of the initial microstructure and chemistry, and inaccurate dosimetry and temperature control, as well as failure properly to correlate these variables, have all contributed to a very large scatter in the experimental embrittlement data base. This has made improvement of the basic understanding of embrittlement very difficult. Therefore, it is necessary to develop a more realistic approach to utilising the data base. This is discussed, and proposals are made. (author)

  2. Effects of nickel on irradiation embrittlement of light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    2005-06-01

    This TECDOC was developed under the IAEA Coordinated Research Project (CRP) entitled Effects of Nickel on Irradiation Embrittlement of Light Water Reactor Pressure Vessel (RPV) Steels. This CRP is the sixth in a series of CRPs to determine the influence of the mechanism and quantify the influence of nickel content on the deterioration of irradiation embrittlement of reactor pressure vessel steels of the Ni-Cr-Mo-V or Mn-Ni-Cr-Mo types. The scientific scope of the programme includes procurement of materials, determination of mechanical properties, irradiation and testing of specimens in power and/or test reactors, and microstructural characterization. Eleven institutes from eight different countries and the European Union participated in this CRP and six institutes conducted the irradiation experiments of the CRP materials. In addition to the irradiation and testing of those materials, irradiation experiments of various national steels were also conducted. Moreover, some institutes performed microstructural investigations of both the CRP materials and national steels. This TECDOC presents and discusses all the results obtained and the analyses performed under the CRP. The results analysed are clear in showing the significantly higher radiation sensitivity of high nickel weld metal (1.7 wt%) compared with the lower nickel base metal (1.2 wt%). These results are supported by other similar results in the literature for both WWER-1000 RPV materials, pressurized water reactor (PWR) type materials, and model alloys. Regardless of the increased sensitivity of WWER-1000 high nickel weld metal (1.7 wt%), the transition temperature shift for the WWER-1000 RPV design fluence is still below the curve predicted by the Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86). For higher fluence, no data were available and the results should not be extrapolated. Although manganese content was not incorporated directly in this CRP

  3. Design considerations of the irradiation test vehicle for the advanced test reactor

    International Nuclear Information System (INIS)

    Tsai, H.; Gomes, I.C.; Smith, D.L.

    1997-01-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements

  4. Vacuum brazing techniques for irradiation devices at TRIGA research reactor

    International Nuclear Information System (INIS)

    Savu, M.; Valeca, S. C.; Amzoi, A.

    2016-01-01

    Metallic thin-walled thermocouples are required for monitoring the temperature value for experiments that are conducted in a nuclear research reactor. The different location wall crossing is made by instrumented passage. Such a passage produced by vacuum brazing using a BNi-7 alloy, represents the proper way to obtain a sealed joint, which can withstand corrosion and high temperatures, having in the same time a small neutron cross section. This paper presents the brazing experiments of K-type thermocouples with stainless steel and Inconel 600 sheath. The sheaths brittleness, hardness changing in joint.s vicinity and structural modification emphasized by metallographic analysis are aspects treated by comparing different samples obtained in brazing laboratory. For finding the correct answer regarding the attenuation of negative effects which are occurring during brazing procedure using Inconel 600 - BNi-7 combination, one can assess both the adopted solution used in designing instrumented passage and thermal regime parameters and its precisely control. (authors)

  5. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    T. A. Tomberlin; S. B. Grover

    2004-11-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

  6. Direct digital control of furnaces irradiated in nuclear reactors

    International Nuclear Information System (INIS)

    Joumard, R.

    1969-01-01

    An experimental direct digital control system has been realised in the 'C.E.N.G.', in order to verify that a computer makes easier the control of the experiments done in the nuclear reactors and to solve the theoretical and technical difficulties. The regulation is applied to thermal processes. The sampled data systems theory permits to choose the type of an efficient and simple digital compensator, and to establish a diagram which gives the values of the correcting parameters (obtained by minimizing the difference between the output and the input when perturbations occur). The programme execute, in simultaneity, supervision and regulation. Complex possibilities of printing out measures and alarms existed. The computer works out an incremental correction which makes step motors to turn. These motors act on the heating organs. The theoretical values and answers have been confirmed. The accuracy was limited essentially by the input quantification (1/1000 th). The comfort of such a system has been noticeable. (author) [fr

  7. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    T. A. Tomberlin; S. B. Grover

    2004-01-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

  8. Irradiations under magnetic field. Measurement of resistivity sample irradiations between 100 and 500 deg C in a swimming-pool reactor

    International Nuclear Information System (INIS)

    Pauleve, J.; Marchand, A.; Blaise, A.

    1964-01-01

    An oven is described which enables the irradiation of small samples in the maximum neutron flux of a swimming-pool reactor of 15 MW (Siloe), at temperatures of between 100 and 500 deg.C defined to ± 0,5 deg.C, The oven is very simple from the technological point of view, and has a diameter of only 27 mm, This permits resistivity measurements to be carried out under irradiation in the reactor, or as another example, it enables irradiations in a magnetic field of 5000 oersteds, created by an immersed solenoid. (authors) [fr

  9. Combination Processes in Food Irradiation. Proceedings of an International Symposium on Combination Processes in Food Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1981-09-15

    Processes in Food Irradiation was held by the IAEA and FAO at the Bandaranaike Memorial International Conference Hall in Colombo on 24-28 November 1980, and the present volume contains the proceedings. One of the most effective means demonstrated of increasing the efficacy of irradiation in the control of food spoilage is the combination of a low irradiation dose with a mild heat treatment. Promising results were reported for the shelf-life extension of mangoes and papayas, and the disinfestation of dried dates. Commercial application of the heat-irradiation treatment for some fruits is expected to follow soon. The Symposium covered other topics, such as the mechanisms of sensitization of microorganisms by physical and chemical agents, improvement of the microbiological quality of foods by combination processes, and the aspects of the wholesomeness and legislation of the food irradiation process. A key issue in the general discussion was the recommendation on the acceptability of food irradiated up to an overall average dose of 10 kGy. This important recommendation had been achieved at a recently convened Joint FAO/IAEA/WHO Expert Committee on the Wholesomeness of Irradiated Foods (27 October - 3 November 1980, Geneva). The breakthrough on the toxicological acceptability constitutes a firm basis for going ahead speedily with the development of practical applications of food irradiation, which should take its rightful place among other food preservation methods in helping to provide more and better food to a world in need. The sponsoring organizations hope that the publication of these proceedings will encourage further research and development of food irradiation to the benefit of mankind.

  10. Combination Processes in Food Irradiation. Proceedings of an International Symposium on Combination Processes in Food Irradiation

    International Nuclear Information System (INIS)

    1981-01-01

    Processes in Food Irradiation was held by the IAEA and FAO at the Bandaranaike Memorial International Conference Hall in Colombo on 24-28 November 1980, and the present volume contains the proceedings. One of the most effective means demonstrated of increasing the efficacy of irradiation in the control of food spoilage is the combination of a low irradiation dose with a mild heat treatment. Promising results were reported for the shelf-life extension of mangoes and papayas, and the disinfestation of dried dates. Commercial application of the heat-irradiation treatment for some fruits is expected to follow soon. The Symposium covered other topics, such as the mechanisms of sensitization of microorganisms by physical and chemical agents, improvement of the microbiological quality of foods by combination processes, and the aspects of the wholesomeness and legislation of the food irradiation process. A key issue in the general discussion was the recommendation on the acceptability of food irradiated up to an overall average dose of 10 kGy. This important recommendation had been achieved at a recently convened Joint FAO/IAEA/WHO Expert Committee on the Wholesomeness of Irradiated Foods (27 October — 3 November 1980, Geneva). The breakthrough on the toxicological acceptability constitutes a firm basis for going ahead speedily with the development of practical applications of food irradiation, which should take its rightful place among other food preservation methods in helping to provide more and better food to a world in need. The sponsoring organizations hope that the publication of these proceedings will encourage further research and development of food irradiation to the benefit of mankind

  11. Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

    Science.gov (United States)

    Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

    2002-12-01

    This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

  12. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  13. MRP-227 Reactor vessel internals inspection planning and initial results at the Oconee nuclear station unit 2

    International Nuclear Information System (INIS)

    Davidsaver, S.B.; Fyfitch, S.; Whitaker, D.E.; Doss, R.L.

    2015-01-01

    The U.S. PWR industry has pro-actively developed generic inspection requirements and standards for reactor vessel (RV) internals. The Electric Power Research Institute (EPRI) Pressurized Water Reactor (PWR) Materials Reliability Program (MRP) has issued MRP-227-A and MRP-228 with mandatory and needed requirements based on the Nuclear Energy Institute (NEI) document NEI 03-08. The inspection and evaluation guidelines contained in MRP-227-A consider eight age-related degradation mechanisms: stress corrosion cracking (SCC), irradiation-assisted stress corrosion cracking (IASCC), wear, fatigue, thermal aging embrittlement, irradiation embrittlement, void swelling and irradiation growth, and thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep. This paper will discuss the decision planning efforts required for implementing the MRP-227-A and MRP-228 requirements and the results of these initial inspections at the Oconee Nuclear power station (ONS) units. Duke Energy and AREVA overcame a significant technology and NDE challenge by successfully completing the first-of-a-kind MRP-227-A scope requirements at ONS-1 in one outage below the estimated dose and with zero safety issues or events. This performance was repeated at ONS-2 a year later. The remote NDE tooling and processes developed to examine the MRP-227-A scope for ONS-1 and ONS-2 are transferable to other PWRs

  14. Upgraded prototype-reactor internal pump for ABWR

    International Nuclear Information System (INIS)

    Kumagai, Mikio; Amemori, Shiro; Saito, Takehiko

    1988-01-01

    In 1983, Toshiba, using their own technology, manufactured a commercial grade reactor internal pump (RIP). Recently, however, a licensing agreement with KSB of West Germany covering the RIP technology, has combined the know-how of KSB with Toshiba's technology to produce a truly high-quality prototype RIP. The pump produces the required coreflow for ABWR at low speed and with high efficiency, and simply by increasing the pump speed to the prior level, the coreflow can be further increased for such advantages as improved fuel cycle economy. Here, the advanced features and test results of the RIP are summarized. (author)

  15. Reactor internals vibration monitoring by neutron noise methods in PWRs

    International Nuclear Information System (INIS)

    Pazsit, I.; Por, G.; Lux, I.

    1983-01-01

    Certain elements of PWR cores such as control/fuel rods or cassettes, or other parts of reactor internals, often represent a vibration problem. Early analyses at operating PWR plant revealed that these vibrations can be detected by in-core neutron detectors, opening up the possibility of vibration monitoring and diagnostics by noise methods. Theoretical methods of calculating vibration induced neutron noise and its application to vibration diagnostics are summarized. Experiments to check theoretical conclusions are under way at the Central Research Institute for Physics, Budapest. (author)

  16. IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-03-01

    The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m{sup 2}, 20 dpa/y in Fe, in a volume of 500 cm{sup 3} and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

  17. IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

    International Nuclear Information System (INIS)

    2003-03-01

    The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m 2 , 20 dpa/y in Fe, in a volume of 500 cm 3 and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

  18. Post-irradiation examination of the vipac fuel assemblies IFA-104 and IFA-203, irradiated in the Halden boiling water reactor

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luyckx, H.J.B.

    1976-11-01

    Two seven pin assemblies, IFA-104 and IFA-203, have been irradiated in the Halding Boiling Water Reactor. The Zircaloy clad pins contained vibrationally compacted, sharp edged UO 2 particles with smear densities of 84-88% theoretical density and a fuel stack length of about 1520 mm. The IFA-104 pins operated satisfactorily during approximately 240 effective full power days at an average linear heat rating of about 340 W/cm, and a peak rating of about 480 W/cm. Ingress of water via a pressure transducer into one of the IFA-203 pins necessitated the premature termination of the irradiation after approximately 30 effective full power days at an average rating of about 460 W/cm, and a peak rating of about 550 W/cm. One IFA-104 pin and two IFA-203 pins failed early. The primary cause of these failures has been sought in the oxidation reaction of the tantalum tube which protected the W/Re thermocouples with the surrounding UO 2 or with moisture in the fuel at temperatures in excess of approximately 1700degC. The fuel centre temperature and the internal gas pressure measurements during the beginning of lifetime and the results of the post-irradiation chemical and microscopic analysis on fuel pin cross-sections have been correlated with corresponding data calculated on a Gapcon-Thermal computer programme written for pellet fuel pins. Fairly good agreement could be achieved between the measured data on the vipac pins and the calculated data of a pellet pin. During the corrosion of the Zircaloy-2 cladding of the IFA-104 pins which were autoclaved in 400degC steam prior to the irradiation, relatively large amounts of H 2 entered the cladding from the D 2 O coolant side. With 0.66 mole% H 2 O in the D 2 O the calculated hydrogen uptake preference ratio was 33

  19. Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

    1998-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10 19 n/cm 2 (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10 19 n/cm 2 (>l MeV). In both cases, irradiations were conducted at ∼290 C and annealing treatments were conducted at ∼454 C. The ORNL and RRC

  20. Using activation method to measure neutron spectrum in an irradiation chamber of a research reactor

    International Nuclear Information System (INIS)

    Zhou Xuemei; Liu Guimin; Wang Xiaohe; Li Da; Meng Lingjie

    2014-01-01

    Neutron spectrum should be measured before test samples are irradiated. Neutron spectrum in an irradiation chamber of a research reactor was measured by using activation method when the reactor is in normal operation under 2 MW. Sixteen kinds of non-fission foils (19 reaction channels) were selected, of which 10 were sensitive to thermal and intermediate energy regions, while the others were of different threshold energy and sensitive to fast energy regions. By measuring the foil radioactivity, the neutron spectrum was unfolded with the iterative methods SAND-II and MSIT. Finally, shielding corrections of group cross-section and main factors affecting the calculation accuracy were studied and the uncertainty of solution was analyzed using the Monte Carlo method in the process of SAND-II. (authors)

  1. Influence of neutron energy on formation of radioisotopes during the irradiation of targets in reactor

    Directory of Open Access Journals (Sweden)

    P. M. Vorona

    2011-09-01

    Full Text Available Method of calculation of nuclear transformations in irradiated targets is realized for selection of optimal conditions for accumulation of radioisotopes in reactor, taking into account contributions of different energy neutrons (thermal, resonance and fast. Wide potentialities of program complex MCNP-4C based on the method of statistical testing (Monte Carlo method were used. Positive in proposed method is that all calculations starting from spectra and fluxes of neutrons in reactor and completing by quantity of accumulating nuclei carry out within the framework of the same methodological approach. It was shown by the example of radioactive 98Mo production in Mo98Mo(n, γ99Mo reaction that for achievement of maximal yield of target radionuclide. it is necessary to irradiate start targets of Molybdenum in hard spectrum with essential contribution of resonance neutrons.

  2. Application of advanced irradiation analysis methods to light water reactor pressure vessel test and surveillance programs

    International Nuclear Information System (INIS)

    Odette, R.; Dudey, N.; McElroy, W.; Wullaert, R.; Fabry, A.

    1977-01-01

    Inaccurate characterization and inappropriate application of neutron irradiation exposure variables contribute a substantial amount of uncertainty to embrittlement analysis of light water reactor pressure vessels. Damage analysis involves characterization of the irradiation environment (dosimetry), correlation of test and surveillance metallurgical and dosimetry data, and projection of such data to service conditions. Errors in available test and surveillance dosimetry data are estimated to contribute a factor of approximately 2 to the data scatter. Non-physical (empirical) correlation procedures and the need to extrapolate to the vessel may add further error. Substantial reductions in these uncertainties in future programs can be obtained from a more complete application of available damage analysis tools which have been developed for the fast reactor program. An approach to reducing embrittlement analysis errors is described, and specific examples of potential applications are given. The approach is based on damage analysis techniques validated and calibrated in benchmark environments

  3. Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Israr, M.; Shami, Qamar-ud-din; Pervez, S.

    1997-11-01

    In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

  4. The dangers of irradiate uranium in nuclear reactors; Les dangers de l'uranium irradie dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Jammet, H; Joffre, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The danger of the uranium cans sur-activated by the use in the nuclear reactors is triple: - Irradiation from afar, during manipulations of the cans. - Contamination of air when decladding. - Contamination of air by fire of uranium in a reactor in function The first two dangers are usual and can be treated thanks to the rules of security in use in the atomic industry. The third has an accidental character and claimed for the use of special and exceptional rules, overflowing the industrial setting, to reach the surrounding populations. (authors) [French] Le danger des cartouches d'uranium suractive par utilisation dans les reacteurs nucleaires est triple: - Irradiation a distance, lors des manipulations des cartouches. - Contamination de l'air au moment de leur degainage. - Contamination de l'air par incendie d'uranium dans un reacteur en fonctionnement. Les deux premiers dangers sont habituels et peuvent etre traites grace aux regles de securite en usage dans l'industrie atomique. Le troisieme revet un caractere accidentel et reclame l'emploi de regles speciales et exceptionnelles, debordant le cadre industriel, pour atteindre celui des populations environnantes. (auteurs)

  5. Effects of neutron radiation and residual stresses on the corrosion of welds in light water reactor internals

    International Nuclear Information System (INIS)

    Schaaf, Bob van der; Gavillet, Didier; Lapena, Jesus; Ohms, Carsten; Roth, Armin; Dyck, Steven van

    2006-01-01

    After many years of operation in Light Water Reactors (LWR) Irradiation Assisted Stress Corrosion Cracking (IASCC) of internals has been observed. In particular the heat-affected zone (HAZ) has been associated with IASCC attack. The welding process induces residual stresses and micro-structural modifications. Neutron irradiation affects the materials response to mechanical loading. IASCC susceptibility of base materials is widely studied, but the specific conditions of irradiated welds are rarely assessed. Core component relevant welds of Type 304 and 347 steels have been fabricated and were irradiated in the High Flux Reactor (HFR) in Petten to 0.3 and 1 dpa (displacement per atom). In-service welds were cut from the thermal shield of the decommissioned BR-3 reactor. Residual stresses, measured using neutron diffraction, ring core tests and X-ray showed residual stress levels up to 400 MPa. Micro-structural characterization showed higher dislocation densities in the weld and HAZ. Neutron radiation increased the dislocation density, resulting in hardening and reduced fracture toughness. The sensitization degree of the welds, measured with the electrochemical potentio-dynamic reactivation method, was negligible. The Slow Strain Rate Tensile (SSRT) tests, performed at 290 deg. C in water with 200 ppb dissolved oxygen, (DO), did not reveal inter-granular cracking. Inter-granular attack of in-service steel is observed in water with 8 ppm (DO), attributed not only to IASCC, but also to IGSCC from thermal sensitization during fabrication. Stress-relieve annealing has caused Cr-grain boundary precipitation, indicating the sensitization. The simulated internal welds, irradiated up to 1.0 dpa, did not show inter-granular cracking with 8 ppm DO. (authors)

  6. Growth of optical transmission loss at 850 nm in silica core optical fibers during fission reactor irradiation

    International Nuclear Information System (INIS)

    Shikama, T.; Narui, M.; Sagawa, T.

    1998-01-01

    Pure, OH-doped and F-doped silica core optical fibers were irradiated in a fission reactor at 400±10 K using an electric heater at a reactor power greater than 10 MW (20% of the full power). The temperature was not controlled well at the early stage of the reactor startup, when the temperature was about 320-340 K. The optical fibers were irradiated with a fast neutron (E>1 MeV) flux of 3.2 x 10 17 n/cm 2 s and a gamma dose rate of 3 x 10 3 Gy/s for 527 h. Optical transmission loss at 850 nm was measured in situ during irradiation. A prompt increase in optical transmission loss was observed as irradiation started, which was probably due to dynamic irradiation effects caused by short-lived and transient defects and is probably recoverable when irradiation ceases. After the prompt increase in optical transmission loss, a so-called radiation hardening was observed in fibers containing OH. Radiation hardening was also observed in 900 ppm OH-doped fiber at the second startup. The optical transmission loss increased linearly with irradiation dose, denoted as the accumulated loss, which we believe is due to irradiation-induced long-lived defects. Accumulated loss dominates radiation-induced optical transmission loss in a fission reactor irradiation. (orig.)

  7. Ion irradiation studies of the origins of pressurized water reactor fuel assembly deformation

    International Nuclear Information System (INIS)

    Hengstler-Eger, Rosmarie Martina

    2012-01-01

    The presented thesis studies ion irradiation damage in Zr-based alloys for pressurized water reactors to explain the origins of unexpectedly high fuel assembly growth in some plants. Transmission electron microscopy was used to investigate the effects of temperature, dose, hydrogen content of the alloy and tensile stress. A clear correlation between the stress orientation towards the crystal lattice and the density of the dislocation loops which are responsible for increased growth was found.

  8. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    Roettger, H.; Hardt, P. von der; Tas, A.; Voorbraak, W.P.

    1981-01-01

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to January 1981

  9. NRX and NRU reactor research facilities and irradiation and examination charges

    International Nuclear Information System (INIS)

    1960-08-01

    This report details the irradiation and examination charges on the NRX and NRU reactors at the Chalk River Nuclear Labs. It describes the NRX and NRU research facilities available to external users. It describes the various experimental holes and loops available for research. It also outlines the method used to calculate the facilities charges and the procedure for applying to use the facilities as well as the billing procedures.

  10. Safety assessment of U–Mo fuel mini plates irradiated in HANARO reactor

    International Nuclear Information System (INIS)

    Jo, Daeseong; Kim, Haksung

    2015-01-01

    Highlights: • Neutronic and thermal-hydraulic analyses of U–Mo fuel irradiated in HANARO reactor. • A mock-up irradiation target was designed and tested to measure the flow rate. • During normal operation, boiling does not occur. • During limiting accidents, boiling occurs. However, fuel integrity is maintained. - Abstract: Neutronic and thermal hydraulic characteristics of U–Mo fuel mini plates irradiated in the HANARO reactor were analyzed for the safety assessment of these plates. A total of eight fuel plates were double-stacked; each stack contained three 8.0 gU/cc U–7Mo fuel plates and one 6.5 gU/cc U–7Mo fuel plate. The neutronic and thermal hydraulic analyses were carried out using the MCNP code and TMAP code, respectively. The core status used in the study was the equilibrium core, and four Control Absorber Rod (CAR) locations were considered: 350 mm, 450 mm, 550 mm, and 650 mm away from the bottom of the core. For the fuels in the lower stack, the maximum heat flux was found at the CAR located at 450 mm. For the fuels in the upper stack, the maximum heat flux was found at the CAR located at 650 mm. The axial power distributions for the upper and lower stacks were selected on the basis of thermal margin analyses. A mock-up irradiation target assembly was designed and tested at the out-of-pile test facility to measure the flow rate through the irradiation site, given that the maximum flow rate through the irradiation site at the HANARO reactor is limited to 12.7 kg/s. For conservative analyses, measurement and correlation uncertainties and engineering hot channel factors were considered. During normal operation, the minimum ONB temperature margins for the lower and upper stacks are 41.6 °C and 31.8 °C, respectively. This means that boiling does not occur. However, boiling occurs during the limiting accidents. Nevertheless, the fuel integrity is maintained since the minimum DNBR are 1.96 for the Reactivity Insertion Accident (RIA) and 2

  11. Progress toward international agreement to improve reactor safety

    International Nuclear Information System (INIS)

    Lieberman, J.I.; Graham, B.

    1993-01-01

    Representatives of nearly one-half of the 114 member states of the International Atomic Energy Agency (IAEA), including the United States, have participated in the development of an international nuclear safety conventions proposed multilateral treaty to improve civil nuclear power reactor safety. A preliminary draft of the convention has been developed (referred to as the draft convention for this report), but discussions are continuing, and when the final convention text will be completed and presented to IAEA member states for signature is uncertain. This report responds to the former and current Chairman's request that we provide information on the development of the nuclear safety convention, including a discussion of (1) the draft convention's scope and objectives, (2) how the convention will be implemented and monitored, (3) the views of selected country representatives on what provisions should be included in the draft convention, and (4) the convention's potential benefits and limitations

  12. Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

  13. International project on innovative nuclear reactors and fuel cycles

    International Nuclear Information System (INIS)

    Cherepnin, Yu.S.; Bezzubtsev, V.S.; Gabaraev, B.A.

    2002-01-01

    Positive changes are currently taking place in nuclear power in the world. Power generation at Nuclear Power Plants (NPPs) is increasing and new units construction and completion rates are growing in some of leading countries. Considerable efforts are made for improving the safety of operating NPPs, effective use of nuclear fuel and solving the spent nuclear fuel and radioactive waste problems. Simultaneously, work are undertaken to develop new reactor technologies to reduce the fundamental drawbacks of conventional nuclear power, namely: insufficient safety, spent fuel and waste handling problems, nuclear material proliferation risk and poor economic competitiveness as compared to fossil-fuel energy sources. One the most important events in this field is an international project implemented by three agencies (OECD-IEA, OECD-NEA, IAEA) for comparative evaluation of new projects, development of Generation IV reactors underway in the US in cooperation with a number of Western countries and, finally, the initiative by Russian President V.V. Putin for consolidation the efforts of interested countries under auspices of IAEA to solve the problem of energy support for sustainable development of humankind, radical solution of non-proliferation problems and environmental sanitation of the Planet of Earth. The 44-th General Conference of IAEA in September 2000 supported the Initiative of Russian President and called all interested countries to unite efforts under the Agency's auspices in the International Project on Innovative Nuclear Reactors and Fuel Cycles to consider and select the most acceptable nuclear technologies of the 21-st century with regard for the drawbacks of today's nuclear power. Main objectivities of INPRO: Promotion of the availability of nuclear power for sustainable satisfaction of the energy needs in 21-st century; Consolidation of efforts by all interested INPRO participating countries (both owners and users of technologies) for joint development of

  14. Meso-scale magnetic signatures for nuclear reactor steel irradiation embrittlement monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Suter, J. D., E-mail: pradeep.ramuhalli@pnnl.gov; Ramuhalli, P., E-mail: pradeep.ramuhalli@pnnl.gov; Hu, S.; Li, Y.; Jiang, W.; Edwards, D. J.; Schemer-Kohrn, A. L.; Johnson, B. R. [Pacific Northwest National Laboratory, 902 Battelle Blvd, Richland, WA 99352 (United States); McCloy, J. S., E-mail: john.mccloy@wsu.edu; Xu, K., E-mail: john.mccloy@wsu.edu [Washington State University, PO Box 642920, Pullman, WA 99164 (United States)

    2015-03-31

    Verifying the structural integrity of passive components in light water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the “state of health” of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of nondestructive evaluation technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results of integrating advanced material characterization techniques with meso-scale computational models. In the future, this will help to provide an interpretive understanding of the state of degradation in structural materials. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. Ongoing research is focused on extending the measurements and models on thin films to gain insights into the structural state of irradiated materials and the resulting impact on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

  15. Irradiation effects on material properties of steels used in nuclear reactors: a literature review

    International Nuclear Information System (INIS)

    Gerceker, N.; Dara, I. H.

    2001-01-01

    The structural materials of a nuclear power plant are of vital importance as they provide mechanical strength, structural support and physical containment for the primary reactor components as well as the nuclear power plant itself. These structural materials comprise mainly of metals and their alloys, ceramics and cermets. However, metals and their alloys are the most widely used materials and the irradiation effects are more pronounced on metallic materials as of their high temperature properties are more sensitive (with respect to ceramics and cermets) to any kind of external effects. The wholesale creation of effects on material properties has been studied for over four decades and it is not realistic to attempt to represent even a small part of the field in single poster paper. In the present contribution, a literature review of the irradiation effects on the material properties of different types of steel alloys will be given because steels are widely used as structural materials in reactors and therefore the irradiation effects on steels may be of paramount importance for reactor design, operation and safety concepts which will be discussed about radiation effects on material properties of steels will provide highlights to better understanding of the origins and development of radiation effects in materials

  16. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Messina, L. [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette (France); KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Olsson, P. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2017-02-15

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a “grey-alloy” approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  17. Methods and instrumentation for investigating Hall sensors during their irradiation in nuclear research reactors

    International Nuclear Information System (INIS)

    Bolshakova, I.; Holyaka, R.; Makido, E.; Marusenkov, A.; Shurygin, F.; Yerashok, V.; Moreau, P. J.; Vayakis, G.; Duran, I.; Stockel, J.; Chekanov, V.; Konopleva, R.; Nazarkin, I.; Kulikov, S.; Leroy, C.

    2009-01-01

    Present work discusses the issues of creating the instrumentation for testing the semiconductor magnetic field sensors during their irradiation with neutrons in nuclear reactors up to fluences similar to neutron fluences in steady-state sensor locations in ITER. The novelty of the work consists in Hall sensor parameters being investigated: first, directly during the irradiation (in real time), and, second, at high irradiation levels (fast neutron fluence > 10 18 n/cm 2 ). Developed instrumentation has been successfully tested and applied in the research experiments on radiation stability of magnetic sensors in IBR-2 (JINR, Dubna) and VVR-M (PNPI, Saint-Petersburg) reactors. The 'Remote-Rad' bench consists of 2 heads (head 1 and head 2) bearing investigated sensors put in a ceramic setting, of electronic unit, of personal computer and of signal lines. Each head contains 6 Hall sensors and a coil for generating test magnetic field. Moreover head 1 contains thermocouples for temperature measurement while the temperature of head 2 is measured by thermo-resistive method. The heads are placed in the reactor channel

  18. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  19. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  20. IAEA international database on irradiated nuclear graphite properties

    International Nuclear Information System (INIS)

    Burchell, T.D.; Clark, R.E.H.; Stephens, J.A.; Eto, M.; Haag, G.; Hacker, P.; Neighbour, G.B.; Janev, R.K.; Wickham, A.J.

    2000-02-01

    This report describes an IAEA database containing data on the properties of irradiated nuclear graphites. Development and implementation of the graphite database followed initial discussions at an IAEA Specialists' Meeting held in September 1995. The design of the database is based upon developments at the University of Bath (United Kingdom), work which the UK Health and Safety Executive initially supported. The database content and data management policies were determined during two IAEA Consultants' Meetings of nuclear reactor graphite specialists held in 1998 and 1999. The graphite data are relevant to the construction and safety case developments required for new and existing HTR nuclear power plants, and to the development of safety cases for continued operation of existing plants. The database design provides a flexible structure for data archiving and retrieval and employs Microsoft Access 97. An instruction manual is provided within this document for new users, including installation instructions for the database on personal computers running Windows 95/NT 4.0 or higher versions. The data management policies and associated responsibilities are contained in the database Working Arrangement which is included as an Appendix to this report. (author)

  1. Long-term irradiation effects on reactor-pressure vessel steels. Investigations on the nanometer scale

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, Arne

    2017-06-01

    The exposure of reactor pressure vessel (RPV) steels to neutron irradiation gives rise to irradiation-enhanced diffusion, a rearrangement of solute atoms and, consequently, a degradation of the mechanical properties. The increasing age of existing nuclear power plants raises new questions specific to long-term operation. Two of them are addressed in this thesis: flux effects and the late-blooming effect. Can low-flux irradiations up to a given fluence be reproduced by more rapid high-flux irradiations up to the same fluence? Can the irradiation response of RPV steels be extrapolated to higher fluences or are there unexpected ''late-blooming'' effects. Small-angle neutron scattering (SANS), atom-probe tomography (APT) and Vickers-hardness testing were applied. A novel Monte-Carlo based fitting algorithm for SANS data was implemented in order to derive statistically reliable characteristics of irradiation-induced solute-atom clusters. APT was applied in selected cases to gain additional information on the composition and the shape of clusters. Vickers hardness testing was performed on the SANS samples to link the nanometer-scale changes to irradiation hardening. The investigations on flux effects show that clusters forming upon high-flux irradiation are smaller and tend to have a higher number density compared to low-flux irradiations at a given neutron fluence. The measured flux dependence of the cluster-size distribution is consistent with the framework of deterministic growth (but not with coarsening) in combination with radiation-enhanced diffusion. Since the two effects on cluster-size and volume fraction partly cancel each other out, no significant effect on the hardening is observed. The investigations of a possible late-blooming effect indicate that the very existence (yes or no) of such an effect depends on the irradiation conditions. Irradiations at lower fluxes and a lower temperature (255 C) give rise to a significant increase of the

  2. Optimization of the irradiation beam in the BNCT research facility at IEA-R1 reactor

    International Nuclear Information System (INIS)

    Castro, Vinicius Alexandre de

    2014-01-01

    Boron Neutron Capture Therapy (BNCT) is a radiotherapeutic technique for the treatment of some types of cancer whose useful energy comes from a nuclear reaction that occurs when thermal neutron impinges upon a Boron-10 atom. In Brazil there is a research facility built along the beam hole number 3 of the IEA-R1 research reactor at IPEN, which was designed to perform BNCT research experiments. For a good performance of the technique, the irradiation beam should be mostly composed of thermal neutrons with a minimum as possible gamma and above thermal neutron components. This work aims to monitor and evaluate the irradiation beam on the sample irradiation position through the use of activation detectors (activation foils) and also to propose, through simulation using the radiation transport code, MCNP, new sets of moderators and filters which shall deliver better irradiation fields at the irradiation sample position In this work, a simulation methodology, based on a MCNP card, known as wwg (weight window generation) was studied, and the neutron energy spectrum has been experimentally discriminated at 5 energy ranges by using a new set o activation foils. It also has been concluded that the BNCT research facility has the required thermal neutron flux to perform studies in the area and it has a great potential for improvement for tailoring the irradiation field. (author)

  3. Staged deployment of the International Fusion Materials Irradiation Facility

    International Nuclear Information System (INIS)

    Takeuchi, H.; Sugimoto, M.; Nakamura, H.

    2001-01-01

    The International Fusion Materials Irradiation Facility (IFMIF) employs an accelerator based D-Li intense neutron source as defined in the 1995-96 Conceptual Design Activity (CDA) study. In 1999, IEA mandated a review of the CDA IFMIF design for cost reduction without change to its original mission. This objective was accomplished by eliminating the previously assumed possibility of potential upgrade of IFMIF beyond the user requirements. The total estimated cost was reduced from $797.2 M to $487.8 M. An option of deployment in 3 stages was also examined to reduce the initial investment and annual expenditures during construction. In this scenario, full performance is achieved gradually with each interim stage as follows. 1st Stage: 20% operation for material selection for ITER breeding blanket, 2nd Stage: 50% operation to demonstrate materials performance of a reference alloy for DEMO, 3rd Stage: full performance operation ( 2MW/m 2 at 500cm 3 ) to obtain engineering data for potential DEMO materials under irradiation up to 100-200 dpa. In summary, the new, reduced cost IFMIF design and staged deployment still satisfies the original mission. The estimated cost of the 1st Stage facility is only $303.6 M making it financially much more attractive. Currently, IFMIF Key Element Technology Phase (KEP) is underway to reduce the key technology risk factors. (author)

  4. Transmutation of waste actinides in thermal reactors: survey calculations of candidate irradiation schemes

    International Nuclear Information System (INIS)

    Gorrell, T.C.

    1978-11-01

    Actinide recycle and transmutation calculations were made for twelve specific thermal reactor environments. The calculations included H 2 O-moderated reactor lattices with enriched U, recycled Pu, and 233 ' 235 U-Th. In addition two D 2 O reactor cases were calculated. When all actinides were recycled into 235 U-enriched fuel, about 10 percent of the transuranic actinides were fissioned per 3-year fuel cycle. About 9 percent of the actinides were fissioned per 3-year fuel cycle when waste actinides (no U or Pu) were irradiated in separate target rods in a U-fuel assembly. When actinides were recycled in separate target assemblies, the fission rate was strongly dependent on the specific loading of the target. Fission rates of 5 to 10 percent per 3-year fuel cycle were observed

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

    International Nuclear Information System (INIS)

    Cowell, B.S.

    1997-06-01

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

  6. Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Adib, M.; Naguib, K.; Morcos, H.N

    2001-09-01

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.

  7. Comparative Analysis of Single and Dual Irradiation Pass of Deep Burn High Temperature Reactor Scenario

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Jo, Chang Keun; Noh, Jae Man

    2012-01-01

    A concept of a deep-burn (DB) of trans uranic (TRU) elements in a high temperature reactor (HTR) has been proposed and studied with a single irradiation pass. However, there is still a significant amount of TRU after burn in an HTR. Therefore, it is necessary to burn more TRU in a fast reactor (FR) with repeated reprocessing such as a pyro-process. In this study, the fuel cycle calculations are performed and the results are compared for a singlepass DB-HHR scenario and a dual-pass sodium-cooled fast reactor (SFR) scenario. For the analysis, front-end and back-end parameters are compared. The calculations were performed by the DANESS (Dynamic Analysis of Nuclear Energy System Strategies), which is an integrated system dynamic fuel cycle analysis code

  8. The metrological problems of irradiation embrittlement of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Vodenicharov, S.; Kamenova, Ts.

    1993-01-01

    Neutron irradiation of reactor pressure vessel steels increases the T k -values of transition temperature from ductile to brittle fracture. This effect is very important in emergency situations, when the water cooling injection in the reactor results in high thermal gradients. In such cases there is a risk from the appearance of a brittle fracture with catastrophic crack propagation speed at relatively low stresses. That is why the T k -value determination is very important for the safe operation of the reactor systems. Some advanced experimental methods for T k -testing and control have been discussed in the present article and the standards of different countries have been compared. The methods applying subsize specimens and welding-restored specimens have been reviewed. (author)

  9. The Effect of Topaz Irradiation to the Quality of Cooling Water Reactor GA Siwabessy

    International Nuclear Information System (INIS)

    Elisabeth Ratnawati; Kawkab Mustofa; Arif Hidayat

    2012-01-01

    Topaz irradiation which applied both inside and outside the reactor core is one utilization of the reactor GA Siwabessy. Topaz consists of silicon clusters containing a combination of aluminum, fluorine and hydroxyl, and impurities. The results of the qualitative analysis of the topaz before irradiation detected europium (Eu-152), potassium (K-40) and sodium (Na-24). While the post-irradiation of topaz detected europium (Eu), cobalt (Co), cesium (Cs), tantalum (Ta), scandium (Sc), iron (Fe), Selenium (Se) and potassium (K). These elements might affect the quality of the cooling water. But the results of the qualitative analysis that were carried out to the primary cooling water did not reveal any elements similar to the elements contained in topaz impurities. Most likely this is because most impurities have been caught by the resin trap in purification systems, because of the results of the analysis of the dirt on the resin trap contained elements similar to the impurities Fe and Co topaz. The purification system makes quality primary cooling water is maintained. From the result shows that chemically the quality of primary cooling water is not affected by the topaz irradiation. (author)

  10. Modeling of Microstructure Evolution in Austenitic Stainless Steels Irradiated Under Light Water Reactor Conditions

    International Nuclear Information System (INIS)

    Gan, J.; Stoller, R.E.; Was, G.S.

    1998-01-01

    A model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 approximately 325 C, up to approximately10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRS. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels

  11. Consequences of the improvement of fast reactor material behavior under irradiation on fuel element performance

    International Nuclear Information System (INIS)

    Leclere, J.; Dupouy, J.M.; Marcon, J.P.

    1979-01-01

    The most important problems in fast reactor fuel element come from the excessive swelling of the structural materials used. The limitations of irradiation time for a given reactor result from the cladding or hexagonal wrapper deformations. Irradiation creep plays a major role, either in inducing additional deformations, or in providing possible ways of accommodation of bending stresses. Progress has been made in designing swelling resistant and/or low irradiation creep modulus materials. For instance in FRANCE, annealed 316 SS has been eliminated from pin and subassembly, and replaced by cold worked 316; we are now considering introduction of stabilizing elements in 316 SS as a further improvement and studying different alloys (nickel alloys, or ferritic steels). It has to be checked that the improvement of irradiation characteristic is not counterbalanced by losses on other properties (embrittlement for instance). Considering that pushing off or eliminating a limit may lead to the onset of a new one, it is porposed to make a review of the consequences of substantial improvement of structural material behavior

  12. Development of a method of testing irradiation devices by gamma scanning inside a research nuclear reactor

    International Nuclear Information System (INIS)

    Michel, Francois.

    1975-01-01

    A tridimensional experiment of spectrometry of an irradiation device located inside the reactor Siloe at a place shielded against spurious radiations, is exposed. The automatic scanning system that was developed, makes it possible to fully analyze in about 24 hours, the irradiation device (fuel pin, coolant and structures). The process combined with a 'pre-processing' program allows first partial results to be simultaneously obtained, more refined results being achieved during the next week, using the 'heavy processing'. The irradiation of the device is only interrupted during the compelling duty shutdown of the reactor, the evolution of the device during the successive irradiation cycles being thus followed up without pertubing said evolution. The reproducibility was studied at a local stage for testing the collimation of the detection unit (1% accuracy) and for testing the whole set 'processing measurement and computation' (5% accuracy). The sensitivity has been illustrated by detecting and measuring local singularities inside fuel (pellets), determining the detection efficiency dependence on the radial distribution of fission products and measuring nucleides inside the coolant flow (limiting value 10 -7 ). The accuracy of the method is evaluated at 5% for relative measurement of an experimental device during its in-pile lifetime and as 10% for quantitative absolute measurements [fr

  13. Design of small-animal thermal neutron irradiation facility at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    Liu, H.B.

    1996-01-01

    The broad beam facility (BBF) at the Brookhaven Medical Research Reactor (BMRR) can provide a thermal neutron beam with flux intensity and quality comparable to the beam currently used for research on neutron capture therapy using cell-culture and small-animal irradiations. Monte Carlo computations were made, first, to compare with the dosimetric measurements at the existing BBF and, second, to calculate the neutron and gamma fluxes and doses expected at the proposed BBF. Multiple cell cultures or small animals could be irradiated simultaneously at the so-modified BBF under conditions similar to or better than those individual animals irradiated at the existing thermal neutron irradiation Facility (TNIF) of the BMRR. The flux intensity of the collimated thermal neutron beam at the proposed BBF would be 1.7 x 10 10 n/cm 2 ·s at 3-MW reactor power, the same as at the TNIF. However, the proposed collimated beam would have much lower gamma (0.89 x 10 -11 cGy·cm 2 /n th ) and fast neutron (0.58 x 10 -11 cGy·cm 2 /n th ) contaminations, 64 and 19% of those at the TNIF, respectively. The feasibility of remodeling the facility is discussed

  14. Influence of reactor irradiation on the protons intercalation and stability of barium cerates and strontium cerates

    International Nuclear Information System (INIS)

    Aksenova, T.I.; Khromushin, I.V.; Zhotabaev, Zh.R.; Kornienko, P.A.; Munasbaeva, K.K.

    2005-01-01

    The work is devoted to study of reactor irradiation influence on the gas-solid exchange processes in the high-temperature proton semiconductors on the base of cerates and strontium. A number of new regularities of influence of content of some proton semiconductors on the gas-solid exchange processes was established. It is shown, that increase of rate of cation doping rate leads to considerable lowering in its of carbonic gas content, and therefore to improvement their tribological properties. It is revealed, that irradiation of polycrystalline samples leads to growth of oxygen amount desorbed from samples, whereas irradiation of monocrystalline samples practically does not has effect on the desorbed oxygen amount. It was found, that character of relation of intercalated in the sample protons depend on sample doping rate

  15. Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

    Science.gov (United States)

    Shao, Lin; Gigax, Jonathan; Chen, Di; Kim, Hyosim; Garner, Frank A.; Wang, Jing; Toloczko, Mychailo B.

    2017-10-01

    Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and, in some cases, introduction of innovative irradiation techniques to alleviate these issues. We briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beam by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. By applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.

  16. The dynamics of immunologic reactions in rats affected by repeated external γ-irradiation and internal irradiation with radionuclides

    International Nuclear Information System (INIS)

    Shubik, V.M.; Livshits, R.E.

    1975-01-01

    Nonspecific factors of immunity and formation of autoantibodies in rats exposed to comparable doses of repeated external γ-irradiation and of internal irradiation with Cs 137 , Sr 90 and I 131 chronically administered to animals have been studied comparatively. No essential variations have been found in changes induced in the immunologic reactions by chronic external and even internal γ-irradiation. Certain peculiarities have been revealed in the character of changes in the immunologic reactions depending on biophysical properties of the radionuclides used in the experiments

  17. Accelerator conceptual design of the international fusion materials irradiation facility

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, M.; Kinsho, M. [Japan Atomic Energy Res. Inst., Tokai, Ibaraki (Japan). Intense Neutron Source Lab.; Jameson, R.A.; Blind, B. [Los Alamos National Lab., NM (United States); Teplyakov, V. [Institute for High Energy Physics, Moscow (Russian Federation); Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J. [Northrop Grumman Corp., Bethpage, NY (United States); Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K. [Johann Wolfgang Goethe Univ., Frankfurt (Germany). Inst. fur Angewandte Phys.; Ferdinand, R.; Lagniel, J.-M. [CEA Saclay LNS, Gif-sur-Yvette (France); Miyahara, A. [Teikyo Univ., Tokyo (Japan); Olivier, M. [CEA DSM, Saclay, Gif-sur-Yvette (France); Piechowiak, E. [Northrop Grumman Corp., Baltimore, MD (United States); Tanabe, Y. [Toshiba Corp., Tsurumi-ku, Yokohama (Japan)

    1998-10-01

    The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.) 8 refs.

  18. Accelerator conceptual design of the international fusion materials irradiation facility

    International Nuclear Information System (INIS)

    Sugimoto, M.; Kinsho, M.; Teplyakov, V.; Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J.; Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K.; Miyahara, A.; Olivier, M.; Piechowiak, E.; Tanabe, Y.

    1998-01-01

    The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.)

  19. The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

    2012-01-01

    This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

  20. Process and apparatus for adjusting a new upper reactor internals in a reactor vessel of a PWR

    International Nuclear Information System (INIS)

    Frizot, A.; Cadaureille, G.; Lalere, C.; Machuron, J.Y.

    1987-01-01

    On the new upper reactor internals is mounted devices for alignment and clearances, before introducing in the reactor vessel. After introducing alignment and clearances are measured. Adjustment pieces are provided for optimum clearances and alignment and fixed after removal from vessel. Decontamination is made by using water jets prior to fitting recess parts [fr