The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

Energy Technology Data Exchange (ETDEWEB)

Barrett, T.R.; Harrison, R. [UKAEA, Nuclear Materials Control Dep., Dounreay (United Kingdom)

1997-07-01

The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

International Nuclear Information System (INIS)

Barrett, T.R.; Harrison, R.

1997-01-01

The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

Microstructural processes in irradiated materials

Energy Technology Data Exchange (ETDEWEB)

Byun, Thak Sang; Morgan, Dane; Jiao, Zhijie; Almer, Jonathan; Brown, Donald

2016-04-01

This is an editorial article (preface) for the publication of symposium papers in the Journal of Nuclear materials: These proceedings contain the papers presented at two symposia, the Microstructural Processes in Irradiated Materials (MPIM) and Characterization of Nuclear Reactor Materials and Components with Neutron and Synchrotron Radiation, held in the TMS 2015, 144th Annual Meeting & Exhibition at Walt Disney World, Orlando, Florida, USA on March 15–19, 2015.

Atomistic Simulations of Small-scale Materials Tests of Nuclear Materials

International Nuclear Information System (INIS)

Shin, Chan Sun; Jin, Hyung Ha; Kwon, Jun Hyun

2012-01-01

Degradation of materials properties under neutron irradiation is one of the key issues affecting the lifetime of nuclear reactors. Evaluating the property changes of materials due to irradiations and understanding the role of microstructural changes on mechanical properties are required for ensuring reliable and safe operation of a nuclear reactor. However, high dose of neuron irradiation capabilities are rather limited and it is difficult to discriminate various factors affecting the property changes of materials. Ion beam irradiation can be used to investigate radiation damage to materials in a controlled way, but has the main limitation of small penetration depth in the length scale of micro meters. Over the past decade, the interest in the investigations of size-dependent mechanical properties has promoted the development of various small-scale materials tests, e.g. nanoindentation and micro/nano-pillar compression tests. Small-scale materials tests can address the issue of the limitation of small penetration depth of ion irradiation. In this paper, we present small-scale materials tests (experiments and simulation) which are applied to study the size and irradiation effects on mechanical properties. We have performed molecular dynamics simulations of nanoindentation and nanopillar compression tests. These atomistic simulations are expected to significantly contribute to the investigation of the fundamental deformation mechanism of small scale irradiated materials

Evaluation of irradiated coating material specimens

International Nuclear Information System (INIS)

Lee, Yong Jin; Nam, Seok Woo; Cho, Lee Moon

2007-12-01

Evaluation result of irradiated coating material specimens - Coating material specimens radiated Gamma Energy(Co 60) in air condition. - Evaluation conditions was above 1 X 10 4 Gy/hr, and radiated TID 2.0 X 10 6 Gy. - The radiated coating material specimens, No Checking, Cracking, Flaking, Delamination, Peeling and Blistering. - Coating system at the Kori no. 1 and APR 1400 Nuclear power plant, evaluation of irradiated coating materials is in accordance with owner's requirement(2.0 X 10 6 Gy)

Analysis of irradiated materials

International Nuclear Information System (INIS)

Bellamy, B.A.

1988-01-01

Papers presented at the UKAEA Conference on Materials Analysis by Physical Techniques (1987) covered a wide range of techniques as applied to the analysis of irradiated materials. These varied from reactor component materials, materials associated with the Authority's radwaste disposal programme, fission products and products associated with the decommissioning of nuclear reactors. An invited paper giving a very comprehensive review of Laser Ablation Microprobe Mass Spectroscopy (LAMMS) was included in the programme. (author)

Capsule Development and Utilization for Material Irradiation Tests

International Nuclear Information System (INIS)

Kang, Young Hwan; Kim, B. G.; Joo, K. N.

2003-05-01

The objective of this project was to establish basic capsule irradiation technology using the multi-purpose research reactor [HANARO] to eventually support national R and D projects of advanced fuel and materials related to domestic nuclear power plants and next generation reactors. There are several national nuclear projects in KAERI, which require several irradiation tests to investigate in-pile behavior of nuclear reactor fuel and materials for the R and D of several types of fuels such as advanced PWR and DUPIC fuels and for the R and D of structural materials such as RPV(reactor pressure vessel) steel, Inconel, zirconium alloy, and stainless steel. At the moment, internal and external researchers in institutes, industries and universities are interested in investigating the irradiation characteristics of materials using the irradiation facilities of HANARO. For these kinds of material irradiation tests, it is important to develop various capsules using our own techniques. The development of capsules requires several leading-edge technologies and our own experiences related to design and fabrication. In the second phase from April 1,2000 to March 31, 2003, the utilization technologies were developed using various sensors for the measurements of temperature, pressure and displacement, and instrumented capsule technologies for the required fuel irradiation tests were developed. In addition, the improvement of the existing capsule technologies and the development of an in-situ measurable creep capsule for specific purposes were done to meet the various requirements of users

Physics and technology of nuclear materials

CERN Document Server

Ursu, Ioan

2015-01-01

Physics and Technology of Nuclear Materials presents basic information regarding the structure, properties, processing methods, and response to irradiation of the key materials that fission and fusion nuclear reactors have to rely upon. Organized into 12 chapters, this book begins with selectively several fundamentals of nuclear physics. Subsequent chapters focus on the nuclear materials science; nuclear fuel; structural materials; moderator materials employed to """"slow down"""" fission neutrons; and neutron highly absorbent materials that serve in reactor's power control. Other chapters exp

Components production and assemble of the irradiation capsule of the Surveillance Program of Materials of the nuclear power plant of Laguna Verde

International Nuclear Information System (INIS)

Medrano, A.

2009-01-01

To predict the effects of the neutrons radiation and the thermal environment about the mechanical properties of the reactor vessel materials of the nuclear power plant of Laguna Verde, a surveillance program is implemented according to the outlines settled by Astm E185-02 -Standard practice for design of surveillance programs for light-water moderated nuclear power reactor vessels-. This program includes the installation of three irradiation capsules of similar materials to those of the reactor vessels, these samples are test tubes for mechanical practices of impact and tension. In the National Institute of Nuclear Research and due to the infrastructure as well as of the actual human resources of the Pilot Plant of Nuclear Fuel Assembles Production it was possible to realize the materials rebuilding extracted in 2005 of Unit 2 of nuclear power plant of Laguna Verde as well as the production, assemble and reassignment of the irradiation capsule made in 2006. At the present time the surveillance materials extracted in 2008 of Unit 1 of the nuclear power plant of Laguna Verde are reconstituting and the components are manufactured for the assembles of the irradiation capsule that will be reinstalled in the reactor vessel in 2010. The purpose of the present work is to describe the necessary components as well as its disposition during the assembles of the irradiation capsule for the surveillance program of the reactors vessel of the nuclear power plant of Laguna Verde. (Author)

Microstructural processes in irradiated materials

Science.gov (United States)

Byun, Thak Sang; Morgan, Dane; Jiao, Zhijie; Almer, Jonathan; Brown, Donald

2016-04-01

These proceedings contain the papers presented at two symposia, the Microstructural Processes in Irradiated Materials (MPIM) and Characterization of Nuclear Reactor Materials and Components with Neutron and Synchrotron Radiation, held in the TMS 2015, 144th Annual Meeting & Exhibition at Walt Disney World, Orlando, Florida, USA on March 15-19, 2015.

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

High dose radiation damage in nuclear energy structural materials investigated by heavy ion irradiation simulation

International Nuclear Information System (INIS)

Zheng Yongnan; Xu Yongjun; Yuan Daqing

2014-01-01

Structural materials in ITER, ADS and fast reactor suffer high dose irradiations of neutrons and/or protons, that leads to severe displacement damage up to lOO dpa per year. Investigation of radiation damage induced by such a high dose irradiation has attracted great attention along with the development of nuclear energy facilities of new generation. However, it is deeply hampered for the lacking of high dose neutron and proton sources. Irradiation simulation of heavy ions produced by accelerators opens up an effective way for laboratory investigation of high dose irradiation induced radiation damage encountered in the ITER, ADS, etc. Radiation damage is caused mainly by atomic displacement in materials. The displacement rate of heavy ions is about lO 3 ∼10 7 orders higher than those of neutrons and protons. High displacement rate of heavy ions significantly reduces the irradiation time. The heavy ion irradiation simulation technique (HIIS) technique has been developed at China Institute of Atomic Energy and a series of the HIIS experiments have been performed to investigate radiation damage in stainless steels, tungsten and tantalum at irradiation temperatures from room temperature to 800 ℃ and in the irradiation dose region up to 100 dpa. The experimental results show that he radiation swelling peak for the modified stainless steel appears in the temperature region around 580 ℃ and the radiation damage is more sensitive to the temperature, the size of the radiation induced vacancy cluster or void increase with the increasing of the irradiation dose, and among the three materials the home-made modified stainless steel has the best radiation resistant property. (authors)

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

Recent irradiation tests for future nuclear system at HANARO

Energy Technology Data Exchange (ETDEWEB)

Cho, Man Soon; Choo, Kee Nam; Yang, Seong Woo; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-05-15

The capsule at HANARO is a device that evaluates the irradiation effects of nuclear materials and fuels, which can reproduce the environment of nuclear power plants and accelerate to reach to the end of life condition. As the integrity assessment and the extension of lifetime of nuclear power plants are recently considered as important issues in Korea, the requirements for irradiation test are gradually being increased. The capacity and capability irradiation tests at HANARO are becoming important because Korea strives to develop SFR (Sodium-cooled Fast Reactor) and VHTR (Very High Temperature Reactor) among the future nuclear system and to export the research reactors and to develop the fusion reactor technology.

Thermodynamics of nuclear materials

International Nuclear Information System (INIS)

Rand, M.H.

1975-01-01

A report is presented of the Fourth International Symposium on Thermodynamics of Nuclear Materials held in Vienna, 21-25 October 1974. The technological theme of the Symposium was the application of thermodynamics to the understanding of the chemistry of irradiated nuclear fuels and to safety assessments for hypothetical accident conditions in reactors. The first four sessions were devoted to these topics and they were followed by four more sessions on the more basic thermodynamics, phase diagrams and the thermodynamic properties of a wide range of nuclear materials. Sixty-seven papers were presented

Nuclear data for the production of radioisotopes in fusion materials irradiation facility

International Nuclear Information System (INIS)

Cheng, E.T.; Schenter, R.E.; Mann, F.M.; Ikeda, Y.

1991-01-01

The fusion materials irradiation facility (FMIF) is a neutron source generator that will produce a high-intensity 14-MeV neutron field for testing candidate fusion materials under reactor irradiation conditions. The construction of such a facility is one of the very important development stages toward realization of fusion energy as a practical energy source for electricity production. As a result of the high-intensity neutron field, 10 MW/m 2 or more equivalent neutron wall loading, and the relatively high-energy (10- to 20-MeV) neutrons, the FMIF, as future fusion reactors, also bears the potential capability of producing a significant quantity of radioisotopes. A study is being conducted to identify the potential capability of the FMIF to produce radioisotopes for medical and industrial applications. Two types of radioisotopes are involved: one is already available; the second might not be readily available using conventional production methods. For those radioisotopes that are not readily available, the FMIF could develop significant benefits for future generations as a result of the availability of such radioisotopes for medical or industrial applications. The current production of radioisotopes could help finance the operation of the FMIF for irradiating the candidate fusion materials; thus this concept is attractive. In any case, nuclear data are needed for calculating the neutron flux and spectrum in the FMIF and the potential production rates of these isotopes. In this paper, the authors report the result of a preliminary investigation on the production of 99 Mo, the parent radioisotope for 99m Tc

Complete Report on the Development of Welding Parameters for Irradiated Materials

Energy Technology Data Exchange (ETDEWEB)

Frederick, Greg [Electric Power Research Inst. (EPRI), Knoxville, TN (United States); Sutton, Benjamin J. [Electric Power Research Inst. (EPRI), Knoxville, TN (United States); Tatman, Jonathan K. [Electric Power Research Inst. (EPRI), Knoxville, TN (United States); Vance, Mark Christopher [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clark, Scarlett R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feng, Zhili [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chen, Jian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Tang, Wei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gibson, Brian T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-11-01

The advanced welding facility at the Radiochemical Engineering Development Center of Oak Ridge National Laboratory, which was conceived to enable research and development of weld repair techniques for nuclear power plant life extension, is now operational. The development of the facility and its advanced welding capabilities, along with the model materials for initial welding trials, were funded jointly by the U.S. Department of Energy, Office of Nuclear Energy, Light Water Reactor Sustainability Program, the Electric Power Research Institute, Long Term Operations Program and the Welding and Repair Technology Center, with additional support from Oak Ridge National Laboratory. Welding of irradiated materials was initiated on November 17, 2017, which marked a significant step in the development of the facility and the beginning of extensive welding research and development campaigns on irradiated materials that will eventually produce validated techniques and guidelines for weld repair activities carried out to extend the operational lifetimes of nuclear power plants beyond 60 years. This report summarizes the final steps that were required to complete weld process development, initial irradiated materials welding activities, near-term plans for irradiated materials welding, and plans for post-weld analyses that will be carried out to assess the ability of the advanced welding processes to make repairs on irradiated materials.

A neutron irradiator to perform nuclear activation

International Nuclear Information System (INIS)

Zamboni, C. B.; Zahn, G.S.; Figueredo, A. M. G.; Madi, T. F.; Yoriyaz, H.; Lima, R. B.; Shtejer, K.; Dalaqua Jr, L.

2001-01-01

The development of appropriate nuclear instrumentation to perform neutron activation analyze (NAA), using thermal and fast neutrons, can be useful to investigate materials outside the reactor premises. Considering this fact, a small size neutron irradiator prototype was developed at IPEN facilities (Instituto de Pesquisas Energeticas e Nucleares - Brazil). Basically, this prototype consists of a cylinder of 1200 mm long and 985 mm diameter (filled with paraffin) with two Am-Be sources (600GBq each) arranged in the longitudinal direction of its geometric center. The material to be irradiated is positioned at a radial direction of the cylinder between the two Am-Be sources. The main advantage of this irradiator is a very stable neutron flux eliminating the use of standard material (measure of the induced activity in the sample by comparative method). This way the process became agile, practical and economic, but quantities at mg levels of samples are necessary to achieve good sensitivity, when the material has a low microscopy neutron cross section. As fast and thermal neutron can be used, the flux distribution, for both, were calculated and the prototype performance is discussed

AGC 2 Irradiated Material Properties Analysis

Energy Technology Data Exchange (ETDEWEB)

Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-05-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

CONTRIBUTION OF HANARO IRRADIATION TECHNOLOGIES TO NATIONAL NUCLEAR R&D

Directory of Open Access Journals (Sweden)

KEE NAM CHOO

2014-08-01

Full Text Available HANARO is a multipurpose research reactor located at the Korea Atomic Energy Research Institute (KAERI. Since the commencement of its operation in 1995, various neutron irradiation facilities, such as rabbit irradiation facilities, fuel test loop (FTL facilities, capsule irradiation facilities, and neutron transmutation doping (NTD facilities, have been developed and actively utilized for various nuclear material irradiation tests requested by users from research institutes, universities, and industries. Most irradiation tests have been related to national R&D relevant to present nuclear power reactors such as the ageing management and safety evaluation of the components. Based on the accumulated experience as well as the sophisticated requirements of users, HANARO has recently supported national R&D projects relevant to new nuclear systems including the System-integrated Modular Advanced Reactor (SMART, research reactors, and future nuclear systems. This paper documents the current state and utilization of irradiation facilities in HANARO, and summarizes ongoing research efforts to deploy advanced irradiation technology.

Materials aging: first predictive modeling of iron under irradiation

International Nuclear Information System (INIS)

Anon.

2005-01-01

Researchers from the CEA-Bruyeres-le-Chatel have been able to quantitatively foresee for the very first time the evolution of irradiation defects inside a structural material. Their results, obtained with iron, will contribute to better understand the aging of the materials of today's nuclear power plants and of future nuclear systems. Short paper. (J.S.)

Radiation Effects in Nuclear Waste Materials

International Nuclear Information System (INIS)

Weber, William J.; Wang, Lumin; Hess, Nancy J.; Icenhower, Jonathan P.; Thevuthasan, Suntharampillai

2003-01-01

The objective of this project is to develop a fundamental understanding of radiation effects in glasses and ceramics, as well as the influence of solid-state radiation effects on aqueous dissolution kinetics, which may impact the performance of nuclear waste forms and stabilized nuclear materials. This work provides the underpinning science to develop improved glass and ceramic waste forms for the immobilization and disposition of high-level tank waste, excess plutonium, plutonium residues and scrap, other actinides, and other nuclear waste streams. Furthermore, this work is developing develop predictive models for the performance of nuclear waste forms and stabilized nuclear materials. Thus, the research performed under this project has significant implications for the immobilization of High-Level Waste (HLW) and Nuclear Materials, two mission areas within the Office of Environmental Management (EM). With regard to the HLW mission, this research will lead to improved understanding of radiation-induced degradation mechanisms and their effects on dissolution kinetics, as well as development of predictive models for waste form performance. In the Nuclear Materials mission, this research will lead to improvements in the understanding of radiation effects on the chemical and structural properties of materials for the stabilization and long-term storage of plutonium, highly-enriched uranium, and other actinides. The research uses plutonium incorporation, ion-beam irradiation, and electron-beam irradiation to simulate the effects of alpha decay and beta decay on relevant glasses and ceramics. The research under this project has the potential to result in improved glass and ceramic materials for the stabilization and immobilization of high-level tank waste, plutonium residues and scraps, surplus weapons plutonium, highly-enriched uranium, other actinides, and other radioactive materials

Radiation Effects in Nuclear Waste Materials

International Nuclear Information System (INIS)

Weber, William J.

2005-01-01

The objective of this project is to develop a fundamental understanding of radiation effects in glasses and ceramics, as well as the influence of solid-state radiation effects on aqueous dissolution kinetics, which may impact the performance of nuclear waste forms and stabilized nuclear materials. This work provides the underpinning science to develop improved glass and ceramic waste forms for the immobilization and disposition of high-level tank waste, excess plutonium, plutonium residues and scrap, other actinides, and other nuclear waste streams. Furthermore, this work is developing develop predictive models for the performance of nuclear waste forms and stabilized nuclear materials. Thus, the research performed under this project has significant implications for the immobilization of High-Level Waste (HLW) and Nuclear Materials, two mission areas within the Office of Environmental Management (EM). With regard to the HLW mission, this research will lead to improved understanding of radiation-induced degradation mechanisms and their effects on dissolution kinetics, as well as development of predictive models for waste form performance. In the Nuclear Materials mission, this research will lead to improvements in the understanding of radiation effects on the chemical and structural properties of materials for the stabilization and long-term storage of plutonium, highly-enriched uranium, and other actinides. The research uses plutonium incorporation, ion-beam irradiation, and electron-beam irradiation to simulate the effects of alpha decay and beta decay on relevant glasses and ceramics. The research under this project has the potential to result in improved glass and ceramic materials for the stabilization and immobilization of high-level tank waste, plutonium residues and scraps, surplus weapons plutonium, highly-enriched uranium, other actinides, and other radioactive materials

Nuclear reactor structural material forming less radioactive corrosion product

International Nuclear Information System (INIS)

Nakazawa, Hiroshi.

1988-01-01

Purpose: To provide nuclear reactor structural materials forming less radioactive corrosion products. Constitution: Ni-based alloys such as inconel alloy 718, 600 or inconel alloy 750 and 690 having excellent corrosion resistance and mechanical property even in coolants at high temperature and high pressure have generally been used as nuclear reactor structural materials. However, even such materials yield corrosion products being attacked by coolants circulating in the nuclear reactor, which produce by neutron irradiation radioactive corrosion products, that are deposited in primary circuit pipeways to constitute exposure sources. The present invention dissolves dissolves this problems by providing less activating nuclear reactor structural materials. That is, taking notice on the fact that Ni-58 contained generally by 68 % in Ni changes into Co-58 under irradiation of neutron thereby causing activation, the surface of nuclear reactor structural materials is applied with Ni plating by using Ni with a reduced content of Ni-58 isotopes. Accordingly, increase in the radiation level of the nuclear reactor structural materials can be inhibited. (K.M.)

In-service irradiated and aged material evaluations

International Nuclear Information System (INIS)

Haggag, F.M.; Nanstad, R.K.; Alexander, D.J.

1995-01-01

The objective of this task is to provide a direct assessment of actual material properties in irradiated components of nuclear reactors, including the effects of irradiation and aging. Four activities are currently in progress: (1) establishing a machining capability for contaminated or activated materials by completing procurement and installation of a computer-based milling machine in a hot cell; (2) machining and testing specimens from cladding materials removed from the Gundremmingen reactor to establish their fracture properties; (3) preparing an interpretive report on the effects of neutron irradiation on cladding; and (4) continuing the evaluation of long-term aging of austenitic structural stainless steel weld metal by metallurgically examining and testing specimens aged at 288 and 343 degrees C and reporting the results, as well as by continuing the aging of the stainless steel cladding toward a total time of 50,000 h

RADIATION EFFECTS IN NUCLEAR WASTE MATERIALS

International Nuclear Information System (INIS)

Weber, William J.

2000-01-01

The objective of this research was to develop fundamental understanding and predictive models of radiation effects in glasses and ceramics at the atomic, microscopic, and macroscopic levels, as well as an understanding of the effects of these radiation-induced solid-state changes on dissolution kinetics (i.e., radionuclide release). The research performed during the duration of this project has addressed many of the scientific issues identified in the reports of two DOE panels [1,2], particularly those related to radiation effects on the structure of glasses and ceramics. The research approach taken by this project integrated experimental studies and computer simulations to develop comprehensive fundamental understanding and capabilities for predictive modeling of radiation effects and dissolution kinetics in both glasses and ceramics designed for the stabilization and immobilization of high-level tank waste (HLW), plutonium residues and scraps, surplus weapons plutonium, other actinides, and other highly radioactive waste streams. Such fundamental understanding is necessary in the development of predictive models because all experimental irradiation studies on nuclear waste materials are ''accelerated tests'' that add a great deal of uncertainty to predicted behavior because the damage rates are orders of magnitude higher than the actual damage rates expected in nuclear waste materials. Degradation and dissolution processes will change with damage rate and temperature. Only a fundamental understanding of the kinetics of all the physical and chemical processes induced or affected by radiation will lead to truly predictive models of long-term behavior and performance for nuclear waste materials. Predictive models of performance of nuclear waste materials must be scientifically based and address both radiation effects on structure (i.e., solid-state effects) and the effects of these solid-state structural changes on dissolution kinetics. The ultimate goal of this

Microbial biofilm growth on irradiated, spent nuclear fuel cladding

International Nuclear Information System (INIS)

Bruhn, D.F.; Frank, S.M.; Roberto, F.F.; Pinhero, P.J.; Johnson, S.G.

2009-01-01

A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 x 10 3 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments

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

Irradiation creep of candidate materials for advanced nuclear plants

Energy Technology Data Exchange (ETDEWEB)

Chen, J., E-mail: jiachao.chen@psi.ch; Jung, P.; Hoffelner, W.

2013-10-15

In the present paper, irradiation creep results of an intermetallic TiAl alloy and two ferritic oxide dispersion strengthened (ODS) steels are summarized. In situ irradiation creep measurements were performed using homogeneous implantation with α- and p-particles to maximum doses of 0.8 dpa at displacement damage rates of 2–8 × 10{sup −6} dpa/s. The strains of miniaturized flat dog-bone specimens were monitored under uniaxial tensile stresses ranging from 20 to 400 MPa at temperatures of 573, 673 and 773 K, respectively. The effects of material composition, ODS particle size, and bombarding particle on the irradiation creep compliance was studied and results are compared to literature data. Evolution of microstructure during helium implantation was investigated in detail by TEM and is discussed with respect to irradiation creep models.

New JMTR irradiation test plan on fuels and materials

International Nuclear Information System (INIS)

Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi

2009-01-01

In order to maintain and enhance safety of light water reactors (LWRs) in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency (JAEA) have decided to refurbish the Japan Materials Testing Reactor (JMTR) and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of newdesign fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor (NSRR) and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses. (author)

Effects of irradiation temperature on polarisation and relaxation characteristics of polymeric materials

Energy Technology Data Exchange (ETDEWEB)

Bornstein, Marcel; Dutz, Hartmut; Goertz, Stefan; Reeve, Scott; Runkel, Stefan [Physikalisches Institut, Bonn Univ. (Germany)

2016-07-01

To achieve significant enhancement of polarisation of solid target materials one must use the principles of dynamic nuclear polarisation and utilise the coupling of the nuclear and electron spins. The unpaired electrons needed can be created as paramagnetic structural defects by irradiation of the material. Polyethylene and polypropylene materials were irradiated at various temperatures and subsequently polarised with microwaves of approximately 70 GHz at temperatures around 1 K. Additionally the samples were investigated with respect to the nature of the created paramagnetic defects using a X-band EPR spectrometer. It was found that the irradiation temperature has a significant effect on the polarisation values achieved and also on the relaxation times of the materials in the 2.5 T magnetic field. The EPR line shape is clearly dominated by the well known alkyl radical structure.

Developing Ultra-small Scale Mechanical Testing Methods and Microstructural Investigation Procedures for Irradiated Materials

Energy Technology Data Exchange (ETDEWEB)

Hosemann, Peter; Kaoumi, Djamel

2018-04-02

Nuclear materials are an essential aspect of nuclear engineering. While great effort is spent on designing more advanced reactors or enhancing a reactor’s safety, materials have been the bottleneck of most new developments. The designs of new reactor concepts are driven by neutronic and thermodynamic aspects, leading to unusual coolants (liquid metal, liquid salt, gases), higher temperatures, and higher radiation doses than conventional light water reactors have. However, any (nuclear) engineering design must consider the materials used in the anticipated application in order to ever be realized. Designs which may look easy, simple and efficient considering thermodynamics or neutronic aspects can show their true difficulty in the materials area, which then prevents them from being deployed. In turn, the materials available are influencing the neutronic and thermodynamic designs and therefore must be considered from the beginning, requiring close collaborations between different aspects of nuclear engineering. If a particular design requires new materials, the licensing of the reactor must be considered, but licensing can be a costly and time consuming process that results in long lead times to realize true materials innovation. Extensive materials evaluation and irradiation campaigns need to be conducted in order to introduce a new material in a nuclear system. For licensing purposes, standard materials testing is key. However, basic scientific studies on new materials or even already used materials have the potential to accelerate the process of materials development or foster predictability of materials that are already in service and therefore are essential in order not to face difficulties later in the development or service stage. Therefore a combination of engineering scale materials evaluation as well as basic scientific understanding of the materials property changes under service condition is key to address potential issues in the process. Ion

Calcium phosphate nuclear materials: apatitic ceramics for separated wastes

International Nuclear Information System (INIS)

Carpena, J.; Lacout, J.L.

2005-01-01

Is it feasible to elaborate conditioning materials for separated high activity nuclear wastes, as actinides or fission products? Specific materials have been elaborated so that the waste is incorporated within the crystalline structure of the most stable calcium phosphate, i.e. apatite. This mineral is able to sustain high irradiation doses assuming a well chosen chemical composition. Mainly two different ways of synthesis have been developed to produce hard apatite ceramics that can be used to condition nuclear wastes. Here we present a data synthesis regarding the elaboration of these apatite nuclear materials that includes experiments on crystallo-chemistry, chemical analysis, leaching and irradiation tests performed for the past fifteen years. (authors)

Radiation damage in nuclear waste materials

International Nuclear Information System (INIS)

Jencic, I.

2000-01-01

Final disposal of high-level radioactive nuclear waste is usually envisioned in some sort of ceramic material. The physical and chemical properties of host materials for nuclear waste can be altered by internal radiation and consequently their structural integrity can be jeopardized. Assessment of long-term performance of these ceramic materials is therefore vital for a safe and successful disposal. This paper presents an overview of studies on several possible candidate materials for immobilization of fission products and actinides, such as spinel (MgAl 2 O 4 ), perovskite (CaTiO 3 ), zircon (ZrSiO 4 ), and pyrochlore (Gd 2 Ti 2 O 7 and Gd 2 Zr 2 O 7 ). The basic microscopic picture of radiation damage in ceramics consists of atomic displacements and ionization. In many cases these processes result in amorphization (metaminctization) of irradiated material. The evolution of microscopic structure during irradiation leads to various macroscopic radiation effects. The connection between microscopic and macroscopic picture is in most cases at least qualitatively known and studies of radiation induced microscopic changes are therefore an essential step in the design of a reliable nuclear waste host material. The relevance of these technologically important results on our general understanding of radiation damage processes and on current research efforts in Slovenia is also addressed. (author)

A disposal centre for irradiated nuclear fuel: conceptual design study

International Nuclear Information System (INIS)

1980-09-01

This report describes a conceptual design of a disposal centre for irradiated nuclear fuel. The surface facilities consist of plants for the preparation of steel cylinders containing irradiated nuclear fuel immobilized in lead, shaft headframe buildings, and all necessary support facilities. The undergound disposal vault is located on one level at a depth of 1000 metres. The cylinders containing the irradiated fuel are emplaced on a one-metre thick layer of backfill material and then completely covered with backfill. All surface and subsurface facilities are described, operations and schedules are summarized, and cost estimates and manpower requirements are given. (auth)

DART, a BCA code to assess and compare primary irradiation damage in nuclear materials submitted to neutron and ion flux - 02002

International Nuclear Information System (INIS)

Luneville, L.; Simeone, D.

2016-01-01

When a material is subjected to a flux of high-energy particles, its constituent atoms can be knocked from their equilibrium positions with a wide range of energies, depending on the exact nature of the collision. The spectrum of damage energy, derived from the exact knowledge of the recoil spectra for each nuclear reaction occurring in the solid, constitutes a vital data set required for understanding how materials evolve under irradiation. The knowledge of such damage energy is relevant to compare the impact of different facilities on the structural behavior and relevant properties of materials. The DART code was developed for two distinct reasons: the first one was a correct determination of the Primary Knocked on Atoms (PKA) spectrum from reliable cross section data libraries and the second was a crude estimation of the damage energy induced by different irradiations. This last term can be a quick estimation of radiation damage produced in the same material by different nuclear plants and particle accelerators. Based on the Binary Collision Approximation, this code allows computing the primary spectra produced by neutrons, ions and electrons as well as the damage energy deposited by these particles in a poly atomic material. It is then a tool to compare radiation damage induced in nuclear reactors as well as in ion beam facilities. This brief paper is followed by the slides of the presentation

Conceptual Design Report for the Irradiated Materials Characterization Laboratory (IMCL)

Energy Technology Data Exchange (ETDEWEB)

Stephanie Austad

2010-06-01

This document describes the design at a conceptual level for the Irradiated Materials Characterization Laboratory (IMCL) to be located at the Materials and Fuels Complex (MFC) at the Idaho National Laboratory (INL). The IMCL is an 11,000-ft2, Hazard Category-2 nuclear facility that is designed for use as a state of the-art nuclear facility for the purpose of hands-on and remote handling, characterization, and examination of irradiated and nonirradiated nuclear material samples. The IMCL will accommodate a series of future, modular, and reconfigurable instrument enclosures or caves. To provide a bounding design basis envelope for the facility-provided space and infrastructure, an instrument enclosure or cave configuration was developed and is described in some detail. However, the future instrument enclosures may be modular, integral with the instrument, or reconfigurable to enable various characterization environments to be configured as changes in demand occur. They are not provided as part of the facility.

Materials Modification Under Ion Irradiation: JANNUS Project

International Nuclear Information System (INIS)

Serruys, Y.; Trocellier, P.; Ruault, M.-O.; Henry, S.; Kaietasov, O.; Trouslard, Ph.

2004-01-01

JANNUS (Joint Accelerators for Nano-Science and Nuclear Simulation) is a project designed to study the modification of materials using multiple ion beams and in-situ TEM observation. It will be a unique facility in Europe for the study of irradiation effects, the simulation of material damage due to irradiation and in particular of combined effects. The project is also intended to bring together experimental and modelling teams for a mutual fertilisation of their activities. It will also contribute to the teaching of particle-matter interactions and their applications. JANNUS will be composed of three accelerators with a common experimental chamber and of two accelerators coupled to a 200 kV TEM

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

Energy Technology Data Exchange (ETDEWEB)

Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science in Radiation and Dynamics Extremes

2016-09-26

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

Nuclear Fuels & Materials Spotlight Volume 4

Energy Technology Data Exchange (ETDEWEB)

I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

2014-04-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

Experimental study associated to irradiation of FBR structural material, (4)

International Nuclear Information System (INIS)

1976-01-01

The study presents one of the bases to evaluate the results of the post-irradiation tests to conduct the thermal control tests related to the second JMTR irradiation (70M-61P) of the demestic austenitic stainless steels for the structural material of the FBR performed by Power Reactor and Nuclear Fuel Development Corporation. The thermal control specimens were given the temperature history which simulated that of the irradiation temperature in vacuum by the electrical furnance, and then the tensile, fatigue and Charpy impact tests were performed. The changes of the material properties caused by the thermal history were investigated. (auth.)

JNC-JAERI united research report. A study on degradation of structural materials under irradiation environment in nuclear reactors

International Nuclear Information System (INIS)

Hoshiya, Taiji; Takaya, Shigeru; Nagae, Yuji; Aoto, Kazumi; Abe, Yasuhiro; Nakamura, Yasuo; Ueno, Fumiyoshi; Nemoto, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Ohmi, Masao; Saito, Junichi; Shimizu, Michio

2004-10-01

Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Energy Research Institute (JAERI) have started a JNC-JAERI united research program cooperatively in fiscal year 2003, which has been aimed for efficient progress and synergistic effect on the research activities of both Institutes in order to lead the facing task of unification between JNC and JAERI. This study has been chosen one of the united research themes because it has been common objective for both Institutes in the research field of structural materials such as Fast Breeder Reactor and Light Water Reactors components. The purpose of the study is to clarify damage mechanism of structural materials under irradiation, and then to develop the methods for damage evaluation and detection in earlier stage of progressing process of damage along grain boundaries. In fiscal year 2003, magnetic flux density distribution (JNC) and micro-corrosion (JAERI) measurement apparatus were newly developed and equipped in Hot Facilities in two Institutes, respectively. The former apparatus, supersensitive Flux Gate sensor was installed, could detector leaked magnetic flux from material damaged by neutron irradiation. The latter one, Atomic Force Microscope was installed, could detect grain boundary corrosion loss after an electrochemical corrosion test of irradiated material. These apparatus were designed and produced in consideration of radiation resistance and remote-controlled operation to equip in hot cells. As the results of preliminary studies using Ni ion irradiated specimen, damage detection by corrosion property in grain boundary was possible but magnetic property change could not detect. We will start the study on neutron irradiation damage by employing the two apparatus as the next step. (author)

Needs of in-situ materials testing under neutron irradiation

International Nuclear Information System (INIS)

Noda, K.; Hishinuma, A.; Kiuchi, K.

1989-01-01

Under neutron irradiation, the component atoms of materials are displaced as primary knock-on atoms, and the energy of the primary knock-on atoms is consumed by electron excitation and nuclear collision. Elementary irradiation defects accumulate to form damage structure including voids and bubbles. In situ test under neutron irradiation is necessary for investigating into the effect of irradiation on creep behavior, the electric properties of ceramics, transport phenomena and so on. The in situ test is also important to investigate into the phenomena related to the chemical reaction with environment during irradiation. Accelerator type high energy neutron sources are preferable to fission reactors. In this paper, the needs and the research items of in situ test under neutron irradiation using a D-Li stripping type high energy neutron source on metallic and ceramic materials are described. Creep behavior is one of the most important mechanical properties, and depends strongly on irradiation environment, also it is closely related to microstructure. Irradiation affects the electric conductibity of ceramics and also their creep behavior. In this way, in situ test is necessary. (K.I.)

Fundamental Technology Development for Radiation Damage in Nuclear Materials

International Nuclear Information System (INIS)

Kwon, Sang Chul; Kwon, J. H.; Kim, E. S. and others

2005-04-01

This project was performed to achieve technologies for the evaluation of radiation effects at materials irradiated at HANARO and nuclear power plants, to establish measurement equipment and software for the analysis of radiation defects and to set up facilities for the measurements of radiation damage with non-destructive methods. Major targets were 1) establishment of hot laboratories and remote handling facilities/ technologies for the radioactive material tests, 2) irradiation test for the simulation of nuclear power plant environment and measurement/calculation of physical radiation damage, 3) evaluation and analysis of nano-scale radiation damage, 4) evaluation of radiation embrittlement with ultrasonic resonance spectrum measurement and electromagnetic measurement and 5) basic research of radiation embrittlement and radiation damage mechanism. Through the performance of 3 years, preliminary basics were established for the application research to evaluation of irradiated materials of present nuclear power plants and GEN-IV systems. Particularly the results of SANS, PAS and TEM analyses were the first output in Korea. And computer simulations of radiation damage were tried for the first time in Korea. The technologies will be developed for the design of GEN-IV material

New materials in nuclear fusion reactors

International Nuclear Information System (INIS)

Iwata, Shuichi

1988-01-01

In the autumn of 1987, the critical condition was attained in the JET in Europe and Japanese JT-60, thus the first subject in the physical verification of nuclear fusion reactors was resolved, and the challenge to the next attainment of self ignition condition started. As the development process of nuclear fusion reactors, there are the steps of engineering, economical and social verifications after this physical verification, and in respective steps, there are the critical problems related to materials, therefore the development of new materials must be advanced. The condition of using nuclear fusion reactors is characterized by high fluence, high thermal flux and strong magnetic field, and under such extreme condition, the microscopic structures of materials change, and they behave much differently from usual case. The subjects of material development for nuclear fusion reactors, the material data base being built up, the materials for facing plasma and high thermal flux, first walls, blanket structures, electric insulators and others are described. The serious effect of irradiation and the rate of defect inducement must be taken in consideration in the structural materials for nuclear fusion reactors. (Kako, I.)

Comparison of material irradiation conditions for fusion, spallation, stripping and fission neutron sources

International Nuclear Information System (INIS)

Vladimirov, P.; Moeslang, A.

2004-01-01

Selection and development of materials capable of sustaining irradiation conditions expected for a future fusion power reactor remain a big challenge for material scientists. Design of other nuclear facilities either in support of the fusion materials testing program or for other scientific purposes presents a similar problem of irradiation resistant material development. The present study is devoted to an evaluation of the irradiation conditions for IFMIF, ESS, XADS, DEMO and typical fission reactors to provide a basis for comparison of the data obtained for different material investigation programs. The results obtained confirm that no facility, except IFMIF, could fit all user requirements imposed for a facility for simulation of the fusion irradiation conditions

Plan and reports of coupled irradiation (JRR-3 and JOYO of research reactors) and hot facilities work (WASTEF, JMTR-HL, MMF and FMF). R and D project on irradiation damage management technology for structural materials of long-life nuclear plant

International Nuclear Information System (INIS)

Matsui, Yoshinori; Yamamoto, Masaya; Yoshitake, Tsunemitsu; Yoshikawa, Katsunori; Iwamatsu, Shigemi; Ichikawa, Shoichi; Yamagata, Ichiro; Soga, Tomonori; Yonekawa, Minoru; Kitamura, Ryoichi; Miyake, Osamu; Takahashi, Hiroyuki; Ishikawa, Kazuyoshi; Kikuchi, Taiji; Usami, Koji; Endo, Shinya; Ichise, Kenichi; Numata, Masami; Onozawa, Atsushi; Aizawa, Masao; Kusunoki, Tsuyoshi; Nakata, Masahito; Abe, Kazuyuki; Ito, Kazuhiro; Takaya, Shigeru; Nagae, Yuji; Wakai, Eiichi; Aoto, Kazumi

2010-03-01

'R and D Project on Irradiation Damage Management Technology for Structural Materials of Long-life Nuclear Plant' was carried out from FY2006 in a fund of a trust enterprise of the Ministry of Education, Culture, Sports, Science and Technology. The coupled irradiations or single irradiation by JOYO fast reactor and JRR-3 thermal reactor were performed for about two years. The irradiation specimens are very important materials to establish of 'Evaluation of Irradiation Damage Indicator' in this research. For the acquisition of the examination specimens irradiated by the JOYO and JRR-3, we summarized about the overall plan, the work process and the results for the study to utilize these reactors and some facilities of hot laboratory (WASTEF, JMTR-HL, MMF and FMF) of the Oarai Research and Development Center and the Nuclear Science Research Institute in the Japan Atomic Energy Agency. (author)

Metallographic examination in irradiated materials examination facility

Energy Technology Data Exchange (ETDEWEB)

Choo, Yong Sun; Lee, Key Soon; Park, Dae Gyu; Ahn, Sang Bok; Yoo, Byoung Ok

1998-01-01

It is very important to have equipment of metallographic examination in hot-cell to observe the micro-structure of nuclear fuels and materials irradiated at nuclear power and/or research reactor. Those equipment should be operated by master-slave manipulators, so they are designed, manufactured and modified to make exercise easy and no trouble. The metallographic examination equipment and techniques as well as its operation procedure are described, so an operator can practice the metallography in hot-cell. (author). 5 refs., 7 tabs., 21 figs.

A Calculation of Nuclear Heating by Activation Product of Structure Materials for the In-core Irradiation Hole in the HANARO

International Nuclear Information System (INIS)

Noh, Tae Yang; Park, B. G.; Kim, M. S.

2016-01-01

Only delayed gamma heating is considered in this paper. Contribution of the delayed gamma heating is expected to be negligible for the reactor power. For the neutron irradiation, however, the contribution of delayed gamma heating is not negligible issue, and it should be evaluated for safety analysis. Additionally, in the case of temperature-sensitive irradiation targets, the delayed gamma heating should be evaluated precisely. For the HANARO, the delayed gamma heating has been evaluated by modifying the library data of the calculation code or by assuming the heating to be conservative value based on prompt gamma heating. For the method of modifying the library data, however, it should be able to estimate isotopes which contribute to heat generation exactly. And furthermore, it should be concerned to determine modified emission yield of gamma-rays depending on the half-life. For the method of assuming conservative value, it is hard to determine whether the assumed heating value is enough conservative or not. In this study, a methodology for evaluation of nuclear heating by structure materials irradiated for a long time is established with the ORIGEN and MCNP codes. And this method is applied to determine the nuclear heating of the RI capsule in the IR2 irradiation hole in the HANARO. In this paper, the methodology for evaluation of heat generation by irradiated structure materials was established by using the ORIGEN and MCNP codes. From this result, the contribution by farther structures was expected to be negligible. Meanwhile, heat generation by delayed gamma-ray was calculated less than 0.03% of heat generation by prompt radiations. The result of this study indicates that there are some remaining issues for the real situation of the neutron irradiation at HANARO.

Some elaborating methods of gamma scanning results on irradiated nuclear fuels

International Nuclear Information System (INIS)

Sternini, E.

1979-01-01

Gamma scanning, as a post-irradiation examination, is a technique which provides a large number of informations on irradiated nuclear fuels. Power profile, fission products distribution, average and local burn-up of single elements structural and nuclear behaviour of fuel materials are examples of the obtained informations. In the present work experimental methods and theoretical calculations used at the CNEN hot cell laboratory for the mentioned purposes are described. Errors arising from the application of the gamma scanning technique are also discussed

The problems of the usage of powerful electrons accelerators for the irradiation of nuclear power stations' equipment and materials

International Nuclear Information System (INIS)

Kovalinska, T.V.; Khalova, N.V.; Ostapenko, I.A.; Sakhno, V.I.; Zelinsky, A.G.; Shlapatska, V.V.

2012-01-01

The possibilities of making the qualification of the materials and equipment of nuclear power stations on modern electrons accelerators of high power are researched. The problems of using this powerful sources of radiation for modern methods of nondestructive control of functional characteristics of the equipment and materials are discussed. The purpose of researches is the determination of the possibility of such works from the point of view of radiation safety of the personnel and the environment. First of all, this problem is connected with the increase of the intensity of secondary irradiation in such processes. The character of secondary irradiation is researched, as well as the dynamics of its energetic spectrum in rooms of powerful industrial accelerator (with beam power of more than 20 kW and average energy of electrons of 1.6 MeV) in regimes of irradiation of the equipment with contents of heavy elements. The original way of solving this problem is suggested. Experimentally proved, that during the usage of the set of compensatory measures, it is also possible to make tests of NPPs' materials and equipment on industrial accelerators of high power

Irradiation facilities for materials research: IFMIF and small scale installations

International Nuclear Information System (INIS)

Perlado, J. M.; Victoria, M.

2007-01-01

The research of advance materials in nuclear fields such as new fission reactors (Generation-IV), Accelerator Driven Systems for Transmutation of Radioactive Wastes and Nuclear Fusion, is becoming very much common in the types of low activation and radiation resistant Materials. Ferritic-Martensitic Steels (based in 9-12 Cr) with or without Oxide Dispersion Techniques (Ytria Nanoparticles), Composites materials are becoming the new generation to answer requirements of high temperature, high radiation resistance of structural materials. Special dedication is appearing in general research programmes to this area of Materials. The understanding of their final performance needs a wider knowledge of the mechanisms of radiation damage in these materials from the atomistic scale to the macroscopic responses. New extensive campaigns are being funded to irradiate from simple elements to model alloys and finally the complex materials themselves. That sequence and its state of art will be presented One clear technique for that understanding is the Multi scale Modelling which includes simulation techniques from quantum mechanics, molecular dynamics, defects diffusion, mesoscopic modelling and finally the macroscopic constitutive relations for macroscopic analysis. However, in each one of these steps is necessary a systematic and well established program of experiments that combines the irradiation and the very detailed analysis with techniques such as Transmission Electron Microscope, Positron Annihilation, SIMS, Atom Probe, Nanoindebntation. A key aspect that wants to be presented in this work is the state of art and discussion of Irradiation Facilities for Materials studies. Those facilities goes from ion implantation sources, small accelerator, Experimental Reactors such High Flux Reactor, sophisticated Triple Beams Sources as JANNUS in France to generate at the same time displacements-hydrogen-helium, and projected very large neutron installation such as IFMIF. The role to

Nuclear Fuels & Materials Spotlight Volume 5

International Nuclear Information System (INIS)

Petti, David Andrew

2016-01-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system. • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.

Fusion materials irradiation test facility: description and status

International Nuclear Information System (INIS)

Trego, A.L.; Parker, E.F.; Hagan, J.W.

1982-01-01

The Fusion Materials Irradiation Test (FMIT) Facility will generate a high-flux, high-energy neutron source that will provide a fusion-like radiation environment for fusion reactor materials development. The neutrons will be produced in a nuclear stripping reaction by impinging a 35 MeV beam of deuterons from an Alvarez-type linear accelerator on a flowing lithium target. The target will be located in a test cell which will provide an irradiation volume of over 750l within which 10 cm 3 will have an average neutron flux of greater than 1.4 x 10 15 n/cm 2 -s and 500 cm 3 an average flux of greater than 2.2 by 10 14 n/cm 2- s with an expected availability factor greater than 65%. The projected fluence within the 10 cm 3 high flux region of FMIT will effect damage upon the materials test specimens to 30 dpa (displacements per atom) for each 90 day irradiation period. This irradiation flux volume will be at least 500 times larger than that of any other facility with comparable neutron energy and will fully meet the fusion materials damage research objective of 100 dpa within three years for the first round of tests

Stochastic simulation of destruction processes in self-irradiated materials

Directory of Open Access Journals (Sweden)

T. Patsahan

2017-09-01

Full Text Available Self-irradiation damages resulting from fission processes are common phenomena observed in nuclear fuel containing (NFC materials. Numerous α-decays lead to local structure transformations in NFC materials. The damages appearing due to the impacts of heavy nuclear recoils in the subsurface layer can cause detachments of material particles. Such a behaviour is similar to sputtering processes observed during a bombardment of the material surface by a flux of energetic particles. However, in the NFC material, the impacts are initiated from the bulk. In this work we propose a two-dimensional mesoscopic model to perform a stochastic simulation of the destruction processes occurring in a subsurface region of NFC material. We describe the erosion of the material surface, the evolution of its roughness and predict the detachment of the material particles. Size distributions of the emitted particles are obtained in this study. The simulation results of the model are in a qualitative agreement with the size histogram of particles produced from the material containing lava-like fuel formed during the Chernobyl nuclear power plant disaster.

Materials for the nuclear - Modelling and simulation of structure materials

International Nuclear Information System (INIS)

Berthoud, Georges; Ducros, Gerard; Feron, Damien; Guerin, Yannick; Latge, Christian; Limoge, Yves; Santarini, Gerard; Seiler, Jean-Marie; Vernaz, Etienne; Cappelaere, Chantal; Andrieux, Catherine; Athenes, Manuel; Baldinozzi, Guido; Bechade, Jean-Luc; Bonin, Bernard; Boutard, Jean-Louis; Brechet, Yves; Bruneval, Fabien; Carassou, Sebastien; Castelier, Etienne; Chartier, Alain; Clouet, Emmanuel; Marinica, Mihai-Cosmin; Crocombette, Jean-Paul; Dupuy, Laurent; Forget, Pierre; Fu, Chu Chun; Garnier, Jerome; Gelebart, Lionel; Henry, Jean; Jourdan, Thomas; Luneville, Laurence; Marini, Bernard; Meslin, Estelle; Nastar, Maylise; Onimus, Fabien; Poussard, Christophe; Proville, Laurent; Ribis, Joel; Robertson, Christian; Rodney, David; Roma, Guido; Sauzay, Maxime; Simeone, David; Soisson, Frederic; Tanguy, Benoit; Toffolon-Masclet, Caroline; Trocellier, Patrick; Van Brutzel, Laurent; Ventelon, Usa; Vincent, Ludovic; Willaime, Francois; Yvon, Pascal; Behar, Christophe; Provitina, Olivier; Lecomte, Michael; Forestier, Alain; Bender, Alexandra; Parisot, Jean-Francois; Finot, Pierre

2016-01-01

This collective publication proposes presentations of scientific approaches implemented to model and simulate the behaviour of materials submitted to irradiation, of associated experimental methods, and of some recent important results. After an introduction presenting the various materials used in different types of nuclear reactors (PWR, etc.), the effects of irradiation at the macroscopic or at the atomic scale, and the multi-scale (time and space) approach to the modelling of these materials, a chapter proposes an overview of modelling tools: multi-scale approach, electronic calculations for condensed matter, inter-atomic potentials, molecular dynamics simulation, thermodynamic and medium force potentials, phase diagrams, simulation of primary damages in reactor materials, kinetic models, dislocation dynamics, production of microstructures for simulation, crystalline visco-plasticity, homogenization methods in continuum mechanics, local approach and probabilistic approach in material fracture. The next part presents tools for experimental validation: tools for microscopic characterization or for mechanical characterization, experimental reactors and tests in atomic pile, tools for irradiation by charged particles. The next chapters presents different examples of thermodynamic and kinetic modelling in the case of various alloys (zirconium alloys, iron-chromium alloys, silicon carbide, austenitic alloys), of plasticity and failure modelling

Nuclear Fuels & Materials Spotlight Volume 5

Energy Technology Data Exchange (ETDEWEB)

Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-10-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system. • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.

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.)

Focusing mirrors for enhanced neutron radiography with thermal neutrons and application for irradiated nuclear fuel

Science.gov (United States)

Rai, Durgesh K.; Abir, Muhammad; Wu, Huarui; Khaykovich, Boris; Moncton, David E.

2018-01-01

Neutron radiography is a powerful method of probing the structure of materials based on attenuation of neutrons. This method is most suitable for materials containing heavy metals, which are not transparent to X-rays, for example irradiated nuclear fuel and other nuclear materials. Neutron radiography is one of the first non-distractive post-irradiated examination methods, which is applied to gain an overview of the integrity of irradiated nuclear fuel and other nuclear materials. However, very powerful gamma radiation emitted by the samples is damaging to the electronics of digital imaging detectors and has so far precluded the use of modern detectors. Here we describe a design of a neutron microscope based on focusing mirrors suitable for thermal neutrons. As in optical microscopes, the sample is separated from the detector, decreasing the effect of gamma radiation. In addition, the application of mirrors would result in a thirty-fold gain in flux and a resolution of better than 40 μm for a field-of-view of about 2.5 cm. Such a thermal neutron microscope can be useful for other applications of neutron radiography, where thermal neutrons are advantageous.

Erosion and corrosion of nuclear power plant materials

International Nuclear Information System (INIS)

1994-01-01

This conference is composed of 23 papers, grouped in 3 sessions which main themes are: analysis of corrosion and erosion damages of nuclear power plant equipment and influence of water chemistry, temperature, irradiations, metallurgical and electrochemical factors, flow assisted cracking, stress cracking; monitoring and control of erosion and corrosion in nuclear power plants; susceptibility of structural materials to erosion and corrosion and ways to improve the resistance of materials, steels, coatings, etc. to erosion, corrosion and cracking

Irradiation of Parts of the X-Gen Nuclear Fuel Assembly made by KNF in HANARO

International Nuclear Information System (INIS)

Choo, K. N.; Cho, M. S.; Shin, Y. T.; Kim, B. G.; Lee, S. H.; Eom, K. B.

2008-01-01

An instrumented capsule has been developed at HANARO (High flux Advanced Neutron Application ReactOr) for the neutron irradiation tests of materials. The capsule system has been actively utilized for the various material irradiation tests requested by users from research institutes, universities, and the industries. As a preliminary test, some specimens made of the parts of a nuclear fuel assembly were inserted in the 05M-07U instrumented capsule and successfully irradiated at HANARO. Based on the results and experience, a new irradiation capsule of 07M-13N was designed, fabricated, and irradiated at HANARO for the evaluation of the neutron irradiation properties of the parts of the X-Gen nuclear fuel assembly made by KNF (Korea Nuclear Fuel). Specimens such as bucking and spring test specimens of spacer grid, microstructure and tensile test specimens of welded parts, tensile, irradiation growth and spring test specimens made of HANA tube, Zirlo, Zircaloy-4 and Inconel-718 were placed in the capsule. The capsule was loaded into the CT test hole of HANARO of a 30MW thermal output and the specimens were irradiated at 295 - 460 .deg. C up to a fast neutron fluence of 1.2x10 21 (n/cm 2 ) (E>1.0MeV)

Gamma exposure rate estimation in irradiation facilities of nuclear research reactors

International Nuclear Information System (INIS)

Daoud, Adrian

2009-01-01

There are experimental situations in the nuclear field, in which dose estimations due to energy-dependent radiation fields are required. Nuclear research reactors provide such fields under normal operation or due to radioactive disintegration of fission products and structural materials activation. In such situations, it is necessary to know the exposure rate of gamma radiation the different materials under experimentation are subject to. Detectors of delayed reading are usually used for this purpose. Direct evaluation methods using portable monitors are not always possible, because in some facilities the entrance with such devices is often impracticable and also unsafe. Besides, these devices only provide information of the place where the measurement was performed, but not of temporal and spatial fluctuations the radiation fields could have. In this work a direct evaluation method was developed for the 'in-situ' gamma exposure rate for the irradiation facilities of the RA-1 reactor. This method is also applicable in any similar installation, and may be complemented by delayed evaluations without problem. On the other hand, it is well known that the residual effect of radiation modifies some properties of the organic materials used in reactors, such as density, colour, viscosity, oxidation level, among others. In such cases, a correct dosimetric evaluation enables in service estimation of material duration with preserved properties. This evaluation is for instance useful when applied to lubricating oils for the primary circuit pumps in nuclear power plants, thus minimizing waste generation. In this work the necessary elements required to estimate in-situ time and space integrated dose are also established for a gamma irradiated sample in an irradiation channel of a nuclear facility with zero neutron flux. (author)

Determination of burn-up of irradiated nuclear fuels using mass spectrometry

International Nuclear Information System (INIS)

Jagadish Kumar, S.; Telmore, V.M.; Shah, R.V.; Sasi Bhushan, K.; Paul, Sumana; Kumar, Pranaw; Rao, Radhika M.; Jaison, P.G.

2017-01-01

Burn-up defined as the atom percent fission, is a vital parameter used for assessing the performance of nuclear fuel during its irradiation in the reactor. Accurate data on the actinide isotopes are also essential for the reliable accountability of nuclear materials and for nuclear safeguards. Both destructive and non-destructive methods are employed in the post-irradiation analysis for the burn-up measurements. Though non-destructive methods are preferred from the point view of remote handling of irradiated fuels with high radioactivity, they do not provide the high accuracy as achieved by the chemical analysis methods. Thus destructive radiochemical and chemical analyses are still the established reference methods for accurate and reliable burn-up determination of irradiated nuclear fuels. In the destructive method, burn-up of irradiated nuclear fuel is determined by correlating the amount of a fission product formed during irradiation with that of heavy elements. Thus the destructive experimental determination of burn-up involves the dissolution of irradiated fuel samples followed by the separation and determination of heavy elements and fission product(s) to be used as burn-up monitor(s). Another approach for the experimental determination of burn-up is based on the changes in the abundances of the heavy element isotopes. A widely accepted method for burn-up determination is based on stable "1"4"8Nd and "1"3"9La as burn-up monitors. Several properties such as non-volatility, nearly same yields for thermal fissions of "2"3"5U and "2"3"9Pu etc justifies the selection of "1"4"8Nd as a burn-up monitor

Materials irradiation subpanel report to BESAC neutron sources and research panel

International Nuclear Information System (INIS)

Birtcher, R.C.; Goland, A.N.; Lott, R.

1992-01-01

The future success of the nuclear power option in the US (fission and fusion) depends critically on the continued existence of a healthy national materials-irradiation program. Consideration of the requirements for acceptable materials-irradiation systems in a new neutron source has led the subcommittee to identify an advanced steady-state reactor (ANS) as a better choice than a spallation neutron source. However, the subcommittee also hastens to point out that the ANS cannot stand alone as the nation's sole high-flux mixed-spectrum neutron irradiation source in the next century. It must be incorporated in a broader program that includes other currently existing neutron irradiation facilities. Upgrading and continuing support for these facilities must be planned. In particular, serious consideration should be given to converting the HFIR into a dedicated materials test reactor, and long-term support for several university reactors should be established

Tungsten - Yttrium Based Nuclear Structural Materials

Science.gov (United States)

Ramana, Chintalapalle; Chessa, Jack; Martinenz, Gustavo

2013-04-01

The challenging problem currently facing the nuclear science community in this 21st century is design and development of novel structural materials, which will have an impact on the next-generation nuclear reactors. The materials available at present include reduced activation ferritic/martensitic steels, dispersion strengthened reduced activation ferritic steels, and vanadium- or tungsten-based alloys. These materials exhibit one or more specific problems, which are either intrinsic or caused by reactors. This work is focussed towards tungsten-yttrium (W-Y) based alloys and oxide ceramics, which can be utilized in nuclear applications. The goal is to derive a fundamental scientific understanding of W-Y-based materials. In collaboration with University of Califonia -- Davis, the project is designated to demonstrate the W-Y based alloys, ceramics and composites with enhanced physical, mechanical, thermo-chemical properties and higher radiation resistance. Efforts are focussed on understanding the microstructure, manipulating materials behavior under charged-particle and neutron irradiation, and create a knowledge database of defects, elemental diffusion/segregation, and defect trapping along grain boundaries and interfaces. Preliminary results will be discussed.

Investigations on neutron irradiated 3D carbon fibre reinforced carbon composite material

Science.gov (United States)

Venugopalan, Ramani; Alur, V. D.; Patra, A. K.; Acharya, R.; Srivastava, D.

2018-04-01

As against conventional graphite materials carbon-carbon (C/C) composite materials are now being contemplated as the promising candidate materials for the high temperature and fusion reactor owing to their high thermal conductivity and high thermal resistance, better mechanical/thermal properties and irradiation stability. The current need is for focused research on novel carbon materials for future new generation nuclear reactors. The advantage of carbon-carbon composite is that the microstructure and the properties can be tailor made. The present study encompasses the irradiation of 3D carbon composite prepared by reinforcement using PAN carbon fibers for nuclear application. The carbon fiber reinforced composite was subjected to neutron irradiation in the research reactor DHRUVA. The irradiated samples were characterized by Differential Scanning Calorimetry (DSC), small angle neutron scattering (SANS), XRD and Raman spectroscopy. The DSC scans were taken in argon atmosphere under a linear heating program. The scanning was carried out at temperature range from 30 °C to 700 °C at different heating rates in argon atmosphere along with reference as unirradiated carbon composite. The Wigner energy spectrum of irradiated composite showed two peaks corresponding to 200 °C and 600 °C. The stored energy data for the samples were in the range 110-170 J/g for temperature ranging from 30 °C to 700 °C. The Wigner energy spectrum of irradiated carbon composite did not indicate spontaneous temperature rise during thermal annealing. Small angle neutron scattering (SANS) experiments have been carried out to investigate neutron irradiation induced changes in porosity of the composite samples. SANS data were recorded in the scattering wave vector range of 0.17 nm-1 to 3.5 nm-1. Comparison of SANS profiles of irradiated and unirradiated samples indicates significant change in pore morphology. Pore size distributions of the samples follow power law size distribution with

A Path Forward to Advanced Nuclear Fuels: Spectroscopic Calorimetry of Nuclear Fuel Materials

International Nuclear Information System (INIS)

Tobin, J.G.

2009-01-01

The goal is to relieve the shortage of thermodynamic and kinetic information concerning the stability of nuclear fuel alloys. Past studies of the ternary nuclear fuel UPuZr have demonstrated constituent redistribution when irradiated or with thermal treatment. Thermodynamic data is key to predicting the possibilities of effects such as constituent redistribution within the fuel rods and interaction with cladding materials

Evaluation method for change of concentration of nuclear fuel material

International Nuclear Information System (INIS)

Kiyono, Takeshi; Ando, Ryohei.

1997-01-01

The present invention provides a method of evaluating the change of concentration of compositions of nuclear fuel element materials loaded to a reactor along with neutron irradiation based on analytic calculation not relying on integration with time. Namely, the method of evaluating the change of concentration of nuclear fuel materials comprises evaluating the changing concentration of nuclear fuel materials based on nuclear fission, capturing of neutrons and radioactive decaying along with neutron irradiation. In this case, an optional nuclide on a nuclear conversion chain is determined as a standard nuclide. When the main fuel material is Pu-239, it is determined as the standard nuclide. The ratio of the concentration of the standard nuclide to that of the nuclide as an object of the evaluation can be expressed by the ratio of the cross sectional area of neutron nuclear reaction of the standard nuclide to the cross sectional area of the neutron nuclear reaction of the nuclide as the object of the evaluation. Accordingly, the concentration of the nuclide as the object of the evaluation can be expressed by an analysis formula shown by an analysis function for the ratio of the concentration of the standard nuclide to the cross section of the neutron nuclear reaction. As a result, by giving an optional concentration of the standard nuclide to the analysis formula, the concentration of each of other nuclides can be determined analytically. (I.S.)

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.

Radiation effects in nuclear materials: Role of nuclear and electronic energy losses and their synergy

Energy Technology Data Exchange (ETDEWEB)

Thomé, Lionel [Centre de Spectrométrie Nucléaire et de Spectrométrie de Masse, CNRS-IN2P3-Université Paris-Sud; Debelle, Aurelien [Universite Paris Sud, Orsay, France; Garrido, Frederico [Universite Paris Sud, Orsay, France; Mylonas, Stamatis [Universite Paris Sud, Orsay, France; Décamps, B. [Universite Paris Sud, Orsay, France; Bachelet, C. [Universite Paris Sud, Orsay, France; Sattonnay, G. [LEMHE/ICMMO, Université Paris-Sud, Bât. Orsay, France; Moll, Sandra [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette; Pellegrino, S. [French Atomic Energy Commission (CEA); Miro, S. [French Atomic Energy Commission (CEA); Trocellier, P. [French Atomic Energy Commission (CEA); Serruys, Y. [French Atomic Energy Commission (CEA); Velisa, G. [French Atomic Energy Commission (CEA); Grygiel, C. [CNRS, France; Monnet, I. [CIMAP, CEA-CNRS-Université de Caen, France; Toulemonde, Marcel [French Atomic Energy Commission (CEA), French National Centre for Scientific Research (CNRS)-ENSICAE; Simon, P. [CEMHTI, CNRS, France; Jagielski, Jacek [Institute for Electronic Materials Technology; Jozwik-Biala, Iwona [Institute for Electronic Materials Technology; Nowicki, Lech [Soltan Institute for Nuclear Studies, Swierk, Poland; Behar, M. [Instituto de Fisica, Universidade Federal do Rio Grande do Sul, Porto Alegre,; Weber, William J [ORNL; Zhang, Yanwen [ORNL; Backman, Marie [University of Tennessee, Knoxville (UTK); Nordlund, Kai [University of Helsinki; Djurabekova, Flyura [University of Helsinki

2013-01-01

Ceramic oxides and carbides are promising matrices for the immobilization and/or transmutation of nuclear wastes, cladding materials for gas-cooled fission reactors and structural components for fusion reactors. For these applications there is a need of fundamental data concerning the behavior of nuclear ceramics upon irradiation. This article is focused on the presentation of a few remarkable examples regarding ion-beam modifications of nuclear ceramics with an emphasis on the mechanisms leading to damage creation and phase transformations. Results obtained by combining advanced techniques (Rutherford backscattering spectrometry and channeling, X-ray diffraction, transmission electron microscopy, Raman spectroscopy) concern irradiations in a broad energy range (from keV to GeV) with the aim of exploring both nuclear collision (Sn) and electronic excitation (Se) regimes. Finally, the daunting challenge of the demonstration of the existence of synergistic effects between Sn and Se is tackled by discussing the healing due to intense electronic energy deposition (SHIBIEC) and by reporting results recently obtained in dual-beam irradiation (DBI) experiments.

Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels

Energy Technology Data Exchange (ETDEWEB)

Todd R. Allen, Director

2011-04-01

The Office of Science, Basic Energy Sciences, has funded the INL as one of the Energy Frontier Research Centers in the area of material science of nuclear fuels. This document is the required annual report to the Office of Science that outlines the accomplishments for the period of May 2010 through April 2011. The aim of the Center for Material Science of Nuclear Fuels (CMSNF) is to establish the foundation for predictive understanding of the effects of irradiation-induced defects on thermal transport in oxide nuclear fuels. The science driver of the center’s investigation is to understand how complex defect and microstructures affect phonon mediated thermal transport in UO2, and achieve this understanding for the particular case of irradiation-induced defects and microstructures. The center’s research thus includes modeling and measurement of thermal transport in oxide fuels with different levels of impurities, lattice disorder and irradiation-induced microstructure, as well as theoretical and experimental investigation of the evolution of disorder, stoichiometry and microstructure in nuclear fuel under irradiation. With the premise that thermal transport in irradiated UO2 is a phonon-mediated energy transport process in a crystalline material with defects and microstructure, a step-by-step approach will be utilized to understand the effects of types of defects and microstructures on the collective phonon dynamics in irradiated UO2. Our efforts under the thermal transport thrust involved both measurement of diffusive phonon transport (an approach that integrates over the entire phonon spectrum) and spectroscopic measurements of phonon attenuation/lifetime and phonon dispersion. Our distinct experimental efforts dovetail with our modeling effort involving atomistic simulation of phonon transport and prediction of lattice thermal conductivity using the Boltzmann transport framework.

Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels

International Nuclear Information System (INIS)

Allen, Todd R.

2011-01-01

The Office of Science, Basic Energy Sciences, has funded the INL as one of the Energy Frontier Research Centers in the area of material science of nuclear fuels. This document is the required annual report to the Office of Science that outlines the accomplishments for the period of May 2010 through April 2011. The aim of the Center for Material Science of Nuclear Fuels (CMSNF) is to establish the foundation for predictive understanding of the effects of irradiation-induced defects on thermal transport in oxide nuclear fuels. The science driver of the center's investigation is to understand how complex defect and microstructures affect phonon mediated thermal transport in UO2, and achieve this understanding for the particular case of irradiation-induced defects and microstructures. The center's research thus includes modeling and measurement of thermal transport in oxide fuels with different levels of impurities, lattice disorder and irradiation-induced microstructure, as well as theoretical and experimental investigation of the evolution of disorder, stoichiometry and microstructure in nuclear fuel under irradiation. With the premise that thermal transport in irradiated UO2 is a phonon-mediated energy transport process in a crystalline material with defects and microstructure, a step-by-step approach will be utilized to understand the effects of types of defects and microstructures on the collective phonon dynamics in irradiated UO2. Our efforts under the thermal transport thrust involved both measurement of diffusive phonon transport (an approach that integrates over the entire phonon spectrum) and spectroscopic measurements of phonon attenuation/lifetime and phonon dispersion. Our distinct experimental efforts dovetail with our modeling effort involving atomistic simulation of phonon transport and prediction of lattice thermal conductivity using the Boltzmann transport framework.

Residual stress improvement mechanism on metal material by underwater laser irradiation

International Nuclear Information System (INIS)

Sano, Yuji; Yoda, Masaki; Mukai, Naruhiko; Obata, Minoru; Kanno, Masanori

2000-01-01

Residual stress improvement technology for component surface by underwater pulsed laser irradiation has been developed as a method of preventing stress corrosion cracking (SCC) of core components in nuclear reactors. In order to optimize the laser irradiation conditions based on a complete understanding of the mechanism, the propagation of a shock wave induced by the impulse of laser irradiation and the dynamic response of the irradiated material were analyzed through time-dependent elasto-plastic calculations with a finite element program. The calculated results are compared with the measured results obtained by experiments in which laser pulses with an energy of 200 mJ are focused to a diameter of 0.8 mm on a water-immersed test piece of 20% cold-worked Type 304 austenitic stainless steel to simulate neutron irradiation hardening. A residual compressive stress, which is nearly equivalent to the yield stress of the processed material, remains on the material surface after passage of the shock wave with enough amplitude to induce a permanent strain. Multiple irradiation of laser pulses extends the stress-improved depth to about 1 mm, which would be the limit corresponding to the three-dimensional dispersion effect of the shock wave. (author)

Determination of material irradiation parameters. Required accuracies and available methods

International Nuclear Information System (INIS)

Cerles, J.M.; Mas, P.

1978-01-01

In this paper, the author reports some main methods to determine the nuclear parameters of material irradiation in testing reactor (nuclear power, burn-up, fluxes, fluences, ...). The different methods (theoretical or experimental) are reviewed: neutronics measurements and calculations, gamma scanning, thermal balance, ... The required accuracies are reviewed: they are of 3-5% on flux, fluences, nuclear power, burn-up, conversion factor, ... These required accuracies are compared with the real accuracies available which are at the present time of order of 5-20% on these parameters

Extreme Spectroscopy: In situ nuclear materials behavior from optical data

Energy Technology Data Exchange (ETDEWEB)

Guimbretiere, G.; Canizares, A.; Raimboux, N.; Omnee, R.; Duval, F.; Ammar, M.R.; Simon, P. [CNRS - UPR3079 CEMHTI, Universite d' Orleans, 45071Orleans cedex 2 (France); Desgranges, L.; Mohun, R. [CEA, DEN, DEC, F-13108 Saint-Paul-Lez-Durance (France); Jegou, C.; Magnin, M. [CEA/DTCD/SECM/LMPA, Marcoule 30207 Bagnols Sur Ceze (France); Clavier, N.; Dacheux, N. [ICSM-UMR5257 CEA/CNRS/UM2/ENSCM, Marcoule, BP17171, 30207 Bagnols sur Ceze (France)

2015-07-01

In the nuclear industry, materials are regularly exposed to high temperature or/and irradiation and a better knowledge and understanding of their behavior under such extreme conditions is a key-point for improvements and further developments. Nowadays, Raman spectroscopy begins to be well known as a promising technique in the post mortem and remote characterization of nuclear materials exposed to extreme conditions. On this topic, at ANIMMA 2013 conference, we have presented some results about its implementation in the study of model or real nuclear fuel. However, the strength of Raman spectroscopy as in situ characterization tool is mainly its ability to be implemented remotely through optical fibers. Aware of this, implementation of other optical techniques can be considered in order to gain information not only on the structural dynamics of materials but also on the electronic charge carrier populations. In this paper, we propose to present our last advances in Raman characterization of nuclear materials and enlarge to the in situ use of complementary optical spectroscopies. Emphasis will be made on the information that can be gained to the behavior of the model fuel depleted UO{sub 2} under extreme conditions of high temperature and ionic irradiation: - In Situ Raman identification of the radiolysis alteration products of UO{sub 2} in contact with water under ionic irradiation. - In Situ Raman recording of the damaged dynamic of UO{sub 2} under inert atmosphere. - In Situ Raman and photo-luminescence study of virgin and damaged UO2 at high temperature. - In Situ study of electronic charge carriers' behavior in U{sub x}Th{sub 1-x}O{sub 2} solid solutions by mean of Iono- and Thermo- luminescence under and post- ionic irradiation. (authors)

Available post-irradiation examination techniques at Romanian institute for nuclear research

International Nuclear Information System (INIS)

Parvan, Marcel; Sorescu, Antonius; Mincu, Marin; Uta, Octavian; Dobrin, Relu

2005-01-01

The Romanian Institute for Nuclear Research (INR) has a set of nuclear facilities consisting of TRIGA 14 MW(th) materials testing reactor and LEPI (Romanian acronym for post-irradiation examination laboratory) which enable to investigate the behaviour of the nuclear fuel and materials under various irradiation conditions. The available techniques of post-irradiation examination (PIE) and purposes of PIE for CANDU reactor fuel are as follows. 1) Visual inspection and photography by periscope: To examine the surface condition such as deposits, corrosion etc. 2) Eddy current testing: To verify the cladding integrity. 3) Profilometry and length measurement performed both before and after irradiation: To measure the parameters which highlight the dimensional changes i.e. diameter, length, diametral and axial sheath deformation, circumferential sheath ridging height, bow and ovality. 4) Gamma scanning and Tomography: To determine the burnup, axial and radial fission products activity distribution and to check for flux peaking and loading homogeneity. 5) Puncture test: To measure the pressure, volume and composition of fission gas and the inner free volume. 6) Optical microscopy: To highlight the structural changes and hydriding, to examine the condition of the fuel-sheath interface and to measure the oxide thickness and Vickers microhardness. 7) Mass spectrometry: To measure the burnup. 8) Tensile testing: To check the mechanical properties. So far, non-destructive and destructive post-irradiation examinations have been performed on a significant number of CANDU fuel rods (about 100) manufactured by INR and irradiated to different power histories in the INR 14 MW(th) TRIGA reactor. These examinations have been performed as part of the Romanian research programme for the manufacturing, development and safety of the CANDU fuel. The paper describes the PIE techniques and some results. (Author)

Neutron interrogation system using high gamma ray signature to detect contraband special nuclear materials in cargo

Science.gov (United States)

Slaughter, Dennis R [Oakland, CA; Pohl, Bertram A [Berkeley, CA; Dougan, Arden D [San Ramon, CA; Bernstein, Adam [Palo Alto, CA; Prussin, Stanley G [Kensington, CA; Norman, Eric B [Oakland, CA

2008-04-15

A system for inspecting cargo for the presence of special nuclear material. The cargo is irradiated with neutrons. The neutrons produce fission products in the special nuclear material which generate gamma rays. The gamma rays are detecting indicating the presence of the special nuclear material.

Reliability of structural materials in nuclear industry

International Nuclear Information System (INIS)

Pinard Legry, G.

1996-01-01

The reliability of nuclear installations is a fundamental point for the exploitation of nuclear energy. It requires an extensive knowledge of the behaviour of materials in the operating conditions and during the expected service life of the installations. In nuclear power plants multiple risks of failure can exist and are expressed by corrosion and deformation phenomena or by modification in the mechanical characteristics of materials. The knowledge of the evolution with time of a given material requires to take into account the data relative to the material itself, to its environment and to the physical conditions of this environment. The study of materials aging needs a more precise knowledge of the kinetics of phenomena at any scale and of their interactions, and a micro- or macro-modeling of their behaviour during long periods of time. This paper gives an overview of the aging phenomena that occur in the structural materials involved in PWR and fast neutron reactors: thermal aging, generalized corrosion, corrosion under constraint, intergranular corrosion, crack growth under loading, wear, irradiation etc.. (J.S.)

New trends in nuclear fuel experimental irradiation. Modern control and acquisition of the irradiation data

International Nuclear Information System (INIS)

Preda, M.; Ciocanescu, M.; Ana, E.M.

2010-01-01

With the irradiation devices used in the irradiation tests, the following experiments have been performed in TRIGA-SCN reactor: a) In capsule-type irradiation devices: - fission gases composition determination; - dimensional measurements; - fission gases pressure measurement; - power pre-ramp and ramp; - power cycling; - structural materials testing. b) In loop-type irradiation device: - power ramp; - multiple power ramps; - overpower. Aiming to develop irradiation tests for advanced nuclear fuel elements, it is mandatory to increase the maximum neutron flux in the core with about 20%. This will lead to reactor power increase up to 21 MW. This objective can be reached through: - increasing the number of fuel clusters in the reactor core; - using the 6x6 fuel cluster to replace the present 5x5 clusters; - relocation of the control rods. In this context, the new control system and the data acquisition system operates online and allows real-time data evaluation. (author)

What destiny could be given to the nuclear irradiated fuel

International Nuclear Information System (INIS)

Mundim, S.G.

1985-01-01

The uranium used in nuclear plants in the production of electric energy is not totally consumed. Part of the fuel that is left over is composed of radioactive material, that represents great danger to earth life. The destines that could be given to the irradiated fuel - reprocessing, provisional or definite storage - depend on the policy adopted by each country that enters the nuclear era, being involved in this increasing problem. (Author) [pt

Irradiation test on connector part for nuclear instrumentation of nuclear powered ship 'Mutsu'

International Nuclear Information System (INIS)

Kudo, Takahiro; Mizushima, Toshihiko; Tsunoda, Tsunemi; Nakazawa, Toshio

1991-01-01

The nuclear instrumetnation facility of the nuclear powered ship 'Mutsu' is composed of neutron detectors, signal cables and the circuits for measurement, and ocntinuously monitors neutron flux. Since this facility treats very faint signals, for the signal cables, coaxial cables and triple coaxial cables are used. The coaxial cables for the nuclear instrumentation are equipped with connectors at both ends, and those are called prefabricated cable. The prefabricated cables are connected to neutron detectors, and installed in the detection holes of the primary shielding tank in the containment vessel. Therefore, at the time of reactor operation, they are exposed to high radiation, and the deterioration of the characteristics of the prefabricated cables is feared. For the purpose of confirming that the part of deteriorating the insulation of the prefabricated cables is connectors, and clarifying the cause of the deterioration of insulation in connector part, the irradiation test of this time was carried out. The environment in which the prefabricated cables are laid, the specifications of the cables and connectors, the materials, gamma ray irradiation and the test results are reported. (K.I.)

Atomic scale modelling of materials of the nuclear fuel cycle

International Nuclear Information System (INIS)

Bertolus, M.

2011-10-01

This document written to obtain the French accreditation to supervise research presents the research I conducted at CEA Cadarache since 1999 on the atomic scale modelling of non-metallic materials involved in the nuclear fuel cycle: host materials for radionuclides from nuclear waste (apatites), fuel (in particular uranium dioxide) and ceramic cladding materials (silicon carbide). These are complex materials at the frontier of modelling capabilities since they contain heavy elements (rare earths or actinides), exhibit complex structures or chemical compositions and/or are subjected to irradiation effects: creation of point defects and fission products, amorphization. The objective of my studies is to bring further insight into the physics and chemistry of the elementary processes involved using atomic scale modelling and its coupling with higher scale models and experimental studies. This work is organised in two parts: on the one hand the development, adaptation and implementation of atomic scale modelling methods and validation of the approximations used; on the other hand the application of these methods to the investigation of nuclear materials under irradiation. This document contains a synthesis of the studies performed, orientations for future research, a detailed resume and a list of publications and communications. (author)

Materials for Nuclear Plants From Safe Design to Residual Life Assessments

CERN Document Server

Hoffelner, Wolfgang

2013-01-01

The clamor for non-carbon dioxide emitting energy production has directly  impacted on the development of nuclear energy. As new nuclear plants are built, plans and designs are continually being developed to manage the range of challenging requirement and problems that nuclear plants face especially when managing the greatly increased operating temperatures, irradiation doses and extended design life spans. Materials for Nuclear Plants: From Safe Design to Residual Life Assessments  provides a comprehensive treatment of the structural materials for nuclear power plants with emphasis on advanced design concepts.   Materials for Nuclear Plants: From Safe Design to Residual Life Assessments approaches structural materials with a systemic approach. Important components and materials currently in use as well as those which can be considered in future designs are detailed, whilst the damage mechanisms responsible for plant ageing are discussed and explained. Methodologies for materials characterization, material...

Selection of materials in nuclear fuel: present and future

International Nuclear Information System (INIS)

Munoz-Reja, C.; Fuentes, L.; Garcia de la Infanta, J. M.; Munoz Sicilia, A.

2013-01-01

One of the main aspects of the nuclear fuel is the selection of materials for the components. The operating conditions of the fuel elements impose a major challenge to materials: high temperature, corrosive aqueous environment, high mechanical properties, long periods of time under these extreme conditions and what is the differentiating factor; the effect of irradiation. The materials are selected to fulfill these severe requirements and also to be able to control and to predict its behavior in the working conditions. Their development, in terms of composition and processing, is based on the continuous follow-up of the operation behavior. Many of these materials are specific of the nuclear industry, such as the uranium dioxide and the zirconium alloys. This article presents the selection and development of the nuclear fuel materials as a function of the services requirements. It also includes a view of the new nuclear fuels materials that are being raised after Fukushima accident. (Author)

Nuclear Materials Stewardship Within the DOE Environmental Management Program

International Nuclear Information System (INIS)

Bilyeu, J. D.; Kiess, T. E.; Gates, M. L.

2002-01-01

The Department of Energy (DOE) Environmental Management (EM) Program has made significant progress in planning disposition of its excess nuclear materials and has recently completed several noteworthy studies. Since establishment in 1997, the EM Nuclear Material Stewardship Program has developed disposition plans for excess nuclear materials to support facility deactivation. All nuclear materials have been removed from the Miamisburg Environmental Management Project (Mound), and disposition planning is nearing completion for the Fernald Environmental Management Project and the Rocky Flats Environmental Technology Site. Only a few issues remain for materials at the Hanford and Idaho sites. Recent trade studies include the Savannah River Site Canyons Nuclear Materials Identification Study, a Cesium/Strontium Management Alternatives Trade Study, a Liquid Technical Standards Trade Study, an Irradiated Beryllium Reflectors with Tritium study, a Special Performance Assessment Required Trade Study, a Neutron Source Trade Study, and development of discard criteria for uranium. A Small Sites Workshop was also held. Potential and planned future activities include updating the Plutonium-239 storage study, developing additional packaging standards, developing a Nuclear Material Disposition Handbook, determining how to recover or dispose of Pu-244 and U-233, and working with additional sites to define disposition plans for their nuclear materials

Radiation damage studies of nuclear structural materials

International Nuclear Information System (INIS)

Barat, P.

2012-01-01

Maximum utilization of fuel in nuclear reactors is one of the important aspects for operating them economically. The main hindrance to achieve this higher burnups of nuclear fuel for the nuclear reactors is the possibility of the failure of the metallic core components during their operation. Thus, the study of the cause of the possibility of failure of these metallic structural materials of nuclear reactors during full power operation due to radiation damage, suffered inside the reactor core, is an important field of studies bearing the basic to industrial scientific views.The variation of the microstructure of the metallic core components of the nuclear reactors due to radiation damage causes enormous variation in the structure and mechanical properties. A firm understanding of this variation of the mechanical properties with the variation of microstructure will serve as a guide for creating new, more radiation-tolerant materials. In our centre we have irradiated structural materials of Indian nuclear reactors by charged particles from accelerator to generate radiation damage and studied the some aspects of the variation of microstructure by X-ray diffraction studies. Results achieved in this regards, will be presented. (author)

Materials modified by irradiation

International Nuclear Information System (INIS)

Chmielewski, A.G.

2007-01-01

Application of radiation in pharmaceutical sciences and cosmetology, polymer materials, food industry, environment, health camre products and packing production is described. Nano-technology is described more detailed, because it is less known as irradiation using technology. Economic influence of the irradiation on the materials value addition is shown

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.)

Segmented fuel irradiation program: investigation on advanced materials

International Nuclear Information System (INIS)

Uchida, H.; Goto, K.; Sabate, R.; Abeta, S.; Baba, T.; Matias, E. de; Alonso, J.

1999-01-01

The Segmented Fuel Irradiation Program, started in 1991, is a collaboration between the Japanese organisations Nuclear Power Engineering Corporation (NUPEC), the Kansai Electric Power Co., Inc. (KEPCO) representing other Japanese utilities, and Mitsubishi Heavy Industries, Ltd. (MHI); and the Spanish Organisations Empresa Nacional de Electricidad, S.A. (ENDESA) representing A.N. Vandellos 2, and Empresa Nacional Uranio, S.A. (ENUSA); with the collaboration of Westinghouse. The objective of the Program is to make substantial contribution to the development of advanced cladding and fuel materials for better performance at high burn-up and under operational power transients. For this Program, segmented fuel rods were selected as the most appropriate vehicle to accomplish the aforementioned objective. Thus, a large number of fuel and cladding combinations are provided while minimising the total amount of new material, at the same time, facilitating an eventual irradiation extension in a test reactor. The Program consists of three major phases: phase I: design, licensing, fabrication and characterisation of the assemblies carrying the segmented rods (1991 - 1994); phase II: base irradiation of the assemblies at Vandellos 2 NPP, and on-site examination at the end of four cycles (1994-1999). Phase III: ramp testing at the Studsvik facilities and hot cell PIE (1996-2001). The main fuel design features whose effects on fuel behaviour are being analysed are: alloy composition (MDA and ZIRLO vs. Zircaloy-4); tubing texture; pellet grain size. The Program is progressing satisfactorily as planned. The base irradiation is completed in the first quarter of 1999, and so far, tests and inspections already carried out are providing useful information on the behaviour of the new materials. Also, the Program is delivering a well characterized fuel material, irradiated in a commercial reactor, which can be further used in other fuel behaviour experiments. The paper presents the main

The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

International Nuclear Information System (INIS)

Toivonen, A.; Moilanen, P.; Pyykkoenen, M.; Taehtinen, S.; Rintamaa, R.; Saario, T.

1998-01-01

Environmentally assisted cracking (EAC) of core materials has become an increasingly important issue of downtime and maintenance costs in nuclear power plants. Small size specimens are necessary in stress corrosion testing of irradiated materials because of difficulties in handling high dose rate materials and because of restricted availability of the materials. The drawback of using small size specimens is that in some cases they do not fulfil the requirements of the relevant testing standards. Recently VTT has developed J-R testing with irradiated and non-irradiated sub size 3 PB specimens, both in inert and in LWR environments. Also, a new materials testing system which will enable simultaneous multiple specimen testing both in laboratory conditions and in operating reactor core is under development. The new testing system will utilize Charpy and sub size 3 PB specimens. The feasibility study of the system has been carried out using different materials. Fracture resistance curves of a Cu-Zr-Cr alloy are shown to be independent of the specimen geometry and size, to some extent. Results gained from tests in simulated boiling water reactor (BWR) water are presented for sensitized SIS 2333 stainless steel. The experimental results indicate that the size of the plastic zone or stress triaxiality must be further studied although no significant effect on the environmentally assisted crack growth rate was observed. (orig.)

The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

Energy Technology Data Exchange (ETDEWEB)

Toivonen, A.; Moilanen, P.; Pyykkoenen, M.; Taehtinen, S.; Rintamaa, R.; Saario, T. [Valtion Teknillinen Tutkimuskeskus, Espoo (Finland)

1998-11-01

Environmentally assisted cracking (EAC) of core materials has become an increasingly important issue of downtime and maintenance costs in nuclear power plants. Small size specimens are necessary in stress corrosion testing of irradiated materials because of difficulties in handling high dose rate materials and because of restricted availability of the materials. The drawback of using small size specimens is that in some cases they do not fulfil the requirements of the relevant testing standards. Recently VTT has developed J-R testing with irradiated and non-irradiated sub size 3 PB specimens, both in inert and in LWR environments. Also, a new materials testing system which will enable simultaneous multiple specimen testing both in laboratory conditions and in operating reactor core is under development. The new testing system will utilize Charpy and sub size 3 PB specimens. The feasibility study of the system has been carried out using different materials. Fracture resistance curves of a Cu-Zr-Cr alloy are shown to be independent of the specimen geometry and size, to some extent. Results gained from tests in simulated boiling water reactor (BWR) water are presented for sensitized SIS 2333 stainless steel. The experimental results indicate that the size of the plastic zone or stress triaxiality must be further studied although no significant effect on the environmentally assisted crack growth rate was observed. (orig.)

Passive nondestructive assay of nuclear materials

International Nuclear Information System (INIS)

Reilly, D.; Ensslin, N.; Smith, H. Jr.; Kreiner, S.

1991-03-01

The term nondestructive assay (NDA) is applied to a series of measurement techniques for nuclear fuel materials. The techniques measure radiation induced or emitted spontaneously from the nuclear material; the measurements are nondestructive in that they do not alter the physical or chemical state of the nuclear material. NDA techniques are characterized as passive or active depending on whether they measure radiation from the spontaneous decay of the nuclear material or radiation induced by an external source. This book emphasizes passive NDA techniques, although certain active techniques like gamma-ray absorption densitometry and x-ray fluorescence are discussed here because of their intimate relation to passive assay techniques. The principal NDA techniques are classified as gamma-ray assay, neutron assay, and calorimetry. Gamma-ray assay techniques are treated in Chapters 1--10. Neutron assay techniques are the subject of Chapters 11--17. Chapters 11--13 cover the origin of neutrons, neutron interactions, and neutron detectors. Chapters 14--17 cover the theory and applications of total and coincidence neutron counting. Chapter 18 deals with the assay of irradiated nuclear fuel, which uses both gamma-ray and neutron assay techniques. Chapter 19 covers perimeter monitoring, which uses gamma-ray and neutron detectors of high sensitivity to check that no unauthorized nuclear material crosses a facility boundary. The subject of Chapter 20 is attribute and semiquantitative measurements. The goal of these measurements is a rapid verification of the contents of nuclear material containers to assist physical inventory verifications. Waste and holdup measurements are also treated in this chapter. Chapters 21 and 22 cover calorimetry theory and application, and Chapter 23 is a brief application guide to illustrate which techniques can be used to solve certain measurement problems

Nuclear materials teaching and research at the University of California, Berkeley

International Nuclear Information System (INIS)

Olander, D.R.; Roberts, J.T.A.

1985-01-01

In academic nuclear engineering departments, research and teaching in the specialized subdiscipline of nuclear materials is usually a one-person or at best a two-person operation. These subcritical sizes invariably result in inadequate overall representation of the many topics in nuclear materials in the research program of the department, although broader coverage of the field is possible in course offerings. Even in course-work, the full range of materials problems important in nuclear technology cannot be dealt with in detail because the small number of faculty involved restricts staffing to as little as a single summary course and generally no more than three courses in this specialty. The contents of the two nuclear materials courses taught at the University of California at Berkeley are listed. Materials research in most US nuclear engineering departments focuses on irradiation effects on metals, but at UC Berkeley, the principal interest is in the high-temperature materials chemistry of UO 2 fuel and Zircaloy cladding

Characterization and testing of materials for nuclear reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2007-03-01

Nuclear techniques in general and neutrons based methods in particular have played and will continue to play an important role in research in materials science and technology. Today the world is looking at nuclear fission and nuclear fusion as the main sources of energy supply for the future. Research reactors have played a key role in the development of nuclear technology. A materials development programme will thus play a major role in the design and development of new nuclear power plants, for the extension of the life of operating reactors as well as for fusion reactors. Against this background, the IAEA had organized a Technical Meeting on Development, Characterization and Testing of Materials - With Special Reference to the Energy Sector under the activity on specific applications of research reactors. The meeting was held in Vienna, May 29- June 2, 2006. There was also participation by experts in techniques, complementary to neutrons. The participants for the technical meeting were experts in the utilization of nuclear techniques namely the high flux and medium flux research reactors, fusion research and positron annihilation. They presented the design, development and utilization of the facilities at their respective centres for materials characterization with main focus on materials for nuclear energy, both fission and fusion. In core irradiation of materials, development of instrument for residual stress measurement in large and / or irradiated specimen, neutron radiography for inspection of irradiated fuel, work on oxide dispersion strengthened (ODS) steels and SiC composites, relevant to future power systems were cited as application of nuclear techniques in fission reactors. The use of neutron scattering for helium bubbles in steel, application of positron annihilation to study helium bubbles in Cu, Ti-stabilized stainless steel and voidswelling studies etc. show that these techniques have an important role in the development of materials for energy

Dry storage of irradiated nuclear fuel

International Nuclear Information System (INIS)

Tolmie, R.D.

1983-01-01

In transferring radioactive material between the preparation and clean chambers of a dry storage complex, irradiated nuclear fuel is posted from the preparation chamber to a sealable canister supported in a closable bucket in the clean chamber, or a contaminated sealed canister is posted from a closed bucket in the clean chamber into the preparation chamber by using a facility comprising two coaxial tubes constituting a closable orifice between the two chambers, the tubes providing sealing means for the bucket, and masking means for the bucket and canister closures together with means for withdrawing the closures into the preparation chamber. (author)

Technical review on irradiation tests and post-irradiation examinations in JMTR

International Nuclear Information System (INIS)

2017-07-01

The Japan Materials Testing Reactor (JMTR) has been contributing to various R and D activities in the nuclear research such as the fundamental research of nuclear materials/ fuels, safety research and development of power reactors, radio isotope (RI) production since its beginning of the operation in 1968. Irradiation technologies and post irradiation examination (PIE) technologies are the important factors for irradiation test research. Moreover, these technologies induce the breakthrough in area of nuclear research. JMTR has been providing unique capabilities for the irradiation test research for about 40 years since 1968. In future, any needs for irradiation test research used irradiation test reactors will continue, such as R and D of generation 4 power reactors, fundamental research of materials/fuels, RI production. Now, decontamination and new research reactor construction are common issue in the world according to aging. This situation is the same in Japan. This report outlines irradiation and PIE technologies developed at JMTR in 40 years to contribute to the technology transfer and human resource development. We hope that this report will be used for the new research rector design as well as the irradiation test research and also used for the human resource development of nuclear engineers in future. (author)

Shield design for the Fusion Materials Irradiation Test facility

International Nuclear Information System (INIS)

Carter, L.L.; Mann, F.M.; Morford, R.J.; Wilcox, A.D.; Johnson, D.L.; Huang, S.T.

1983-03-01

The shield design for the Fusion Materials Irradiation Test facility is based upon one-, two- and three-dimensional transport calculations with experimental measurements utilized to refine the nuclear data including the neutron cross sections from 20 to 50 MeV and the gamma ray and neutron source terms. The high energy neutrons and deuterons produce activation products from the numerous reactions that are kinematically allowed. The analyses for both beam-on and beam-off (from the activation products) conditions have required extensive nuclear data libraries and the utilization of Monte Carlo, discrete ordinates, point kernel and auxiliary computer codes

Proceedings of the second international conference on advances in nuclear materials: abstract booklet and souvenir

International Nuclear Information System (INIS)

2011-01-01

Nuclear materials form special class of materials which either act as fuel for the nuclear reactors or form the structure of the reactors and the allied systems. The topics covered in this conference are: materials challenges for thermal and fast reactors, technological advances in nuclear fuels and components, materials for future reactors, fuel cycles and materials challenges, materials degradation and life management, advanced materials development, modelling and simulation, advanced materials- II, advanced materials for future reactors, development of advanced fuel and structural materials, zirconium alloy developments, irradiation effects and PIE, advanced nuclear fuels, corrosion and materials characterization. Papers relevant to INIS are indexed separately

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.

A short note on physical properties to irradiated nuclear fuel by means of X-ray diffraction and neutron scattering techniques

Energy Technology Data Exchange (ETDEWEB)

Abdullah, Yusof, E-mail: yusofabd@nuclearmalaysia.gov.my; Husain, Hishamuddin; Hak, Cik Rohaida Che; Alias, Nor Hayati; Yusof, Mohd Reusmaazran; Kasim, Norasiah Ab; Zali, Nurazila Mat [Malaysian Nuclear Agency, Bangi, Kajang 43000, Selangor (Malaysia); Mohamed, Abdul Aziz [College of Engineering, Universiti Tenaga National, Jalan Ikram-Uniten, 43000 Kajang, Selangor (Malaysia)

2015-04-29

For nuclear reactor applications, understanding the evolution of the fuel materials microstructure during irradiation are of great importance. This paper reviews the physical properties of irradiated nuclear fuel analysis which are considered to be of most importance in determining the performance behavior of fuel. X-rays diffraction was recognize as important tool to investigate the phase identification while neutron scattering analyses the interaction between uranium and other materials and also investigation of the defect structure.

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

Neutron irradiation effects on plasma facing materials

Science.gov (United States)

Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

2000-12-01

This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

Neutron irradiation effects on plasma facing materials

International Nuclear Information System (INIS)

Barabash, V.; Federici, G.; Roedig, M.; Snead, L.L.; Wu, C.H.

2000-01-01

This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed

Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Krishna, R. [Dalton Cumbrian Facility, Dalton Nuclear Institute, The University of Manchester, Westlakes Science & Technology Park, Moor Row, Whitehaven, Cumbria, CA24 3HA (United Kingdom); Jones, A.N., E-mail: Abbie.Jones@manchester.ac.uk [Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, M13 9PL (United Kingdom); McDermott, L.; Marsden, B.J. [Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, M13 9PL (United Kingdom)

2015-12-15

Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure. - Highlights: • Irradiated graphite

Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

International Nuclear Information System (INIS)

Krishna, R.; Jones, A.N.; McDermott, L.; Marsden, B.J.

2015-01-01

Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure. - Highlights: • Irradiated graphite exhibits

Spent fuel working group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

International Nuclear Information System (INIS)

1993-11-01

In a memo dated 19 August 1993, Secretary O'Leary assigned the Office of Environment, Safety and Health the primary responsibility to identify, characterize, and assess the safety, health, and environmental vulnerabilities of the DOE's existing storage conditions and facilities for the storage of irradiated reactor fuel and other reactor irradiated nuclear materials. This volume is divided into three major sections. Section 1 contains the Working Group Assessment Team reports on the following facilities: Hanford Site, INEL, SRS, Oak Ridge Site, West Valley Site, LANL, BNL, Sandia, General Atomics (San Diego), Babcock ampersand Wilcox (Lynchburg Technology Center), and ANL. Section 2 contains the Vulnerability Development Forms from most of these sites. Section 3 contains the documents used by the Working Group in implementing this initiative

Novel designs of continuous process for dissolution of irradiated nuclear fuels

International Nuclear Information System (INIS)

Rance, P.J.W.; Tinsley, T.P.; Polyakov, A.S.; Raginsky, L.S.; Morkovnikov, V.E.; Morozov, N.V.; Eliseev, S.P.

1998-01-01

A novel design of continuous dissolver for the dissolution of irradiated nuclear fuels is described. The development of the dissolver has resulted from a successful collaboration over the last four years between British Nuclear Fuels plc (UK) and the A.A. Bochvar All-Russia Research Institute of Inorganic Materials (Russia). An overview of the development work carried out on three different models is presented, and results from each of these are discussed. The dissolver provides many advantages over current designs of dissolvers. (author)

DBMS Development of Irradiated Materials and Spare parts on master-slave manipulator in IMEF

Energy Technology Data Exchange (ETDEWEB)

Choo, Y. S.; Kim, D. S.; Jung, Y. H.; Kim, H. M.; Yoo, B. O.; Baik, S. J.; Hong, K. P.; Ahn, S. B.; Ryu, W. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-12-15

The data of irradiated specimens(include nuclear fuel) which are transported from research reactor and commercial power reactor and the spare parts of the master-slave manipulator for the IMEF facility, which is operated since 1996, were controlled and managed through the Hangul and Excel software. But it is recommended to use a special program, which is developed for DBMS, for the beneficial control and systematic management of all irradiated specimens, especially assuming the increase of specimen's kind and amount by increasing customers in the near future. This report summarized the whole logical and physical processes and results about following items : - Management System of Irradiated Materials including nuclear fuel - Management System of spare parts for the master-slave manipulator.

Computational modeling of the behavior of nuclear materials (2). Molecular simulations for nuclear materials. Current situation and future perspective

International Nuclear Information System (INIS)

Okita, Taira; Itakura, Mitsuhiro

2017-01-01

Molecular simulations for nuclear materials aim to reproduce atomistic-scale phenomena induced by irradiation and infer the change in material properties. In the present work, recent progress in this field is presented. In particular, the following three topics are explained: (1) Quantification of lattice defects formation process induced by fast neutron collision. (2) Identification of dislocation-channeling mechanism induced by interactions between defect clusters and dislocations. (3) Modeling of the three dimensional movement of defect clusters using molecular dynamics and kinetic Monte Carlo simulations. (author)

Overall viscoplastic behavior of non-irradiated porous nuclear ceramics

International Nuclear Information System (INIS)

Monerie, Yann; Gatt, Jean-Marie

2006-01-01

This paper deals with the overall behavior of nonlinear viscous and porous nuclear ceramics. Bi-viscous isotropic porous materials are considered: the matrix is subjected to two power-law viscosities with different exponents related to two stationary temperature-activated creeping mechanisms (scattering-creep and dislocation-creep), and this matrix contains a low porosity volume fraction. The overall behavior of these types of composite materials is obtained with the help of quadratic strain-rate potentials combined with experimental-based coupling function depending on stress and temperature. For each creeping mechanism, the hollow sphere model of [Michel, J.-C., Suquet, P., 1992. The constitutive law of nonlinear viscous and porous materials. Journal of the Mechanics and Physics of Solids 40, 783-812] is used. Mechanical parameters of the resulting model are identified and validated in the particular case of non-irradiated uranium dioxide nuclear ceramics. This model predicts, under pure thermo-mechanical loading, a variation of the material volume and a variation of the porosity volume fraction (the so-called densification or swelling). (authors)

Potential use of nuclear irradiation technology for the microbial industry in Malaysia

International Nuclear Information System (INIS)

Raja, P.; Isyanti, I.

2002-01-01

The potential application of irradiation is numerous. The process seems technically feasible for sterilization of medical devices, pharmaceutical products, foods, cosmetics, packaging materials, scientific laboratory materials like petri dishes, growth media, complying with phytosanitary requirements for fruit and vegetables etc. An awareness of biofertilizer or microbial product has established itself to be a significant part of food production and consumption internationally as well as in Malaysia. People are discovering organic agriculture can serve their need for safe quality food, environment conservation and social accountability. One of the major technical constraints of viable microbial or biofertilizer product production is contaminated free carrier materials. In order to develop the good microbial product, we are practicing the gamma irradiation on our carrier materials and successfully established the product. The different carrier materials were packed and used 25 kGy Cobalt-60 source (JS 8900, SINAGAMA, MINT) for sterilization and simultaneously the materials were sterilized in steam. The cultured Trichoderma virile and T harzianum were mixed with sterilized carrier and observed the colony formin. unit (CFU) on regular intervals. The results showed the efficient colony forming unit on both sterilization method of materials. However, the nuclear irradiation is more cost effective and time saving than steam method. (Author)

Long-term radiation effects on commercial cable-insulating materials irradiated at CERN

International Nuclear Information System (INIS)

Maier, P.; Stolarz, A.

1983-01-01

Long-term irradiation damage tests have been carried out on a variety of flexible cable-insulating materials offered to CERN by different European cable manufacturers. Tensile test specimens were exposed for a maximum of three years in high-level radiation areas of the Super Proton Synchrotron (SPS) and for comparison at high dose rates in a nuclear reactor. The degradation of mechanical properties after irradiation in air depends not only on the total absorbed dose, but also on the dose rate for most of these polymer compounds. These dose-rate effects vary between material types and for different compounds. The results presented here illustrate the difference in radiation damage between short-term and long-term irradiation conditions in a typical service application for the various materials tested. They also allow safety factors to be estimated for the extrapolation of the limiting exposure in service from accelerated material tests in the range of dose rates covered. A discussion of the available models of the dose-rate effects results in a conservative estimate for extrapolation to low dose rates from measured values at intermediate dose rates of the order of 0.1 Gy/s. Based on short-term irradiation tests only, the safety factors to be applied depend on the end-point criterion used, and may vary between 1 and 10 for the range of dose rates and materials considered here. (orig.)

Ion-Neutron Irradiated BOR60 Sample Preparation and Characterization: Nuclear Science User Facility 2017 Milestone Report

Energy Technology Data Exchange (ETDEWEB)

Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Quinlan B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

This document outlines the results obtained by Oak Ridge National Laboratory (ORNL) in collaboration with the University of Michigan-led Consolidated Innovative Nuclear Research project, “Feasibility of combined ion-neutron irradiation for accessing high dose levels.” In this reporting period, neutron irradiated were prepared and shipped to the University of Michigan for subsequent ion irradiation. The specimens were returned to ORNL’s Low Activation Materials Development and Analysis facility, prepared via focused ion beam for examination using scanning/transmission electron microscopy (S/TEM), and then examined using S/TEM to measure the as-irradiated microstructure. This report briefly summarizes the S/TEM results obtained at ORNL’s Low Activation Materials Development and Analysis facility.

Optimization of the irradiation conditions of some control components and materials for the nuclear power plants and the radiation stability of certain types of plastic lubricants

International Nuclear Information System (INIS)

Pesek, M.; Rerichova, M.; Trebicky, V.; Chvojka, M.

1989-01-01

Fail-safe operation of various safeguard devices, operational and auxiliary equipments and control components, e.g. servomotors other engines and various appliances, is required for a safe operation of nuclear power plants. Non-metal materials, control components, motors and other appliances have to be tested and their properties evaluated after γ-irradiation with doses corresponding to the assumed long term radiation commitment and also to the irradiation caused by an eventual accident. The radiation stability of greases used in devices exposed to high doses of the ionizing radiation presents a rather serious and important problem. The results of some tests and the evaluation of the properties of irradiated plastic lubricants are described. (author)

Characterization of damaging in apatitic materials irradiated with heavy ions and thermally annealed

International Nuclear Information System (INIS)

Tisserand, R.

2004-12-01

Some minerals belonging to the family of apatite are seen to be potential candidates for use as conditioning matrices or transmutation targets for high level nuclear waste management. Indeed, studies of natural nuclear reactors (Oklo) highlighted the strong ability of these minerals to anneal irradiation damage. In order to determine the global behaviour of these materials, we performed a fundamental study on the evolution of irradiation damage induced by various heavy ions in two apatites: a natural phospho-calcic fluor-apatite from Durango and a synthetic sintered mono-silicated fluor-apatite, called britholite. The damage in these materials was measured by using channelling R.B.S. and X-ray diffraction respectively and by determining an amorphization effective radius Re. The results revealed a similar behaviour for both apatites according to the electronic energy deposit at the entrance of the material. In addition, the effect of an isothermal annealing at 300 C was quantified on a mono-silicated britholite previously irradiated with Kr ions. We highlighted in this case the return of the lattice parameters to their initial values, followed by a partial and slow rebuilding of the crystalline lattice versus the annealing time. Finally, we followed the changes in the morphology of etch pits in the Durango fluor-apatite after acid dissolution as a function of the energy deposit by the ions. We showed that the influence of crystallography leads quickly to opening angles close to 30 degrees. The calculation of etching velocities within the irradiated material highlighted that there is a range of deposit energy where the velocity ratio increases strongly before becoming constant. (author)

Nuclear materials

International Nuclear Information System (INIS)

1996-01-01

In 1998, Nuclear Regulatory Authority of the Slovak Republic (NRA SR) performed 38 inspections, 25 of them were performed in co-operation with IAEA inspectors. There is no fresh nuclear fuel at Bohunice A-1 NPP at present. Fresh fuel of Bohunice V-1 and V-2 NPPs is inspected in the fresh fuel storage.There are 327 fresh fuel assemblies in Mochovce NPP fresh fuel storage. In addition to that, are also 71 small users of nuclear materials in Slovakia. In most cases they use: covers made of depleted uranium for non-destructive works, detection of level in production plants, covers for therapeutical sources at medical facilities. In. 1995, NRA SR issued 4 new licences for nuclear material withdrawal. In the next part manipulation with nuclear materials, spent fuel stores and illegal trafficking in nuclear materials are reported

Microstructure of irradiated materials

International Nuclear Information System (INIS)

Robertson, I.M.

1995-01-01

The focus of the symposium was on the changes produced in the microstructure of metals, ceramics, and semiconductors by irradiation with energetic particles. the symposium brought together those working in the different material systems, which revealed that there are a remarkable number of similarities in the irradiation-produced microstructures in the different classes of materials. Experimental, computational and theoretical contributions were intermixed in all of the sessions. This provided an opportunity for these groups, which should interact, to do so. Separate abstracts were prepared for 58 papers in this book

Simulation of tensile stress-strain properties of irradiated type 316 SS by heavily cold-worked material

International Nuclear Information System (INIS)

Muto, Yasushi; Jitsukawa, Shiro; Hishinuma, Akimichi

1995-07-01

Type 316 stainless steel is one of the most promising candidate materials to be used for the structural parts of plasma facing components in the nuclear fusion reactor. The neutron irradiation make the material brittle and reduces its uniform elongation to almost zero at heavy doses. In order to apply such a material of reduced ductility to structural components, the structural integrity should be examined and assured by the fracture mechanics. The procedure requires a formulated stress-strain relationship. However, the available irradiated tensile test data are very limited at present, so that the cold-worked material was used as a simulated material in this study. Property changes of 316 SS, that is, a reduction of uniform elongation and an enhancement of yield stress are seemingly very similar for both the irradiated 316 SS and the cold-worked one. The specimens made of annealed 316 SS, 20% (or 15%) cold worked one and 40% cold worked one were prepared. After the formulation of stress strain behavior, the equation for the cold-worked 316 SS was fitted to the data on irradiated material under the assumption that the yield stress is the same for both materials. In addition, the upper limit for the plastic strain was introduced using the data on the irradiated material. (author)

The RaDIATE High-Energy Proton Materials Irradiation Experiment at the Brookhaven Linac Isotope Producer Facility

Energy Technology Data Exchange (ETDEWEB)

Ammigan, Kavin; et al.

2017-05-01

The RaDIATE collaboration (Radiation Damage In Accelerator Target Environments) was founded in 2012 to bring together the high-energy accelerator target and nuclear materials communities to address the challenging issue of radiation damage effects in beam-intercepting materials. Success of current and future high intensity accelerator target facilities requires a fundamental understanding of these effects including measurement of materials property data. Toward this goal, the RaDIATE collaboration organized and carried out a materials irradiation run at the Brookhaven Linac Isotope Producer facility (BLIP). The experiment utilized a 181 MeV proton beam to irradiate several capsules, each containing many candidate material samples for various accelerator components. Materials included various grades/alloys of beryllium, graphite, silicon, iridium, titanium, TZM, CuCrZr, and aluminum. Attainable peak damage from an 8-week irradiation run ranges from 0.03 DPA (Be) to 7 DPA (Ir). Helium production is expected to range from 5 appm/DPA (Ir) to 3,000 appm/DPA (Be). The motivation, experimental parameters, as well as the post-irradiation examination plans of this experiment are described.

Metallic structural materials in the nuclear environment: some problems illustrating new methods

International Nuclear Information System (INIS)

Brechet, Y.

2002-01-01

The structural components of the nuclear industry are submitted to a number of aggressions, mechanical, chemical and physical (irradiation). As a consequence, the problem of durability and ageing of such structures is a key issue. The understanding of the phenomena involved implies the description and modelling of atomic scale events (irradiation point defects) resulting in fluxes of matter (diffusion under irradiation), in the dynamic evolution of structural defects (dislocation loops, cavities,...), with major consequences on mechanical properties (yield stress, fracture behaviour), with, in addition, phenomena coupled between mechanical behaviour and chemical environment. It is therefore the totality of materials science which is involved in understanding the behaviour of metallic structural materials in the nuclear environment. The aim of the present paper is to illustrate some examples currently under investigation, and some of the new approaches involved in the understanding of mechanical behaviour (a scale transition from the atomic to the macroscopic). The input from large computer simulations as well as the value of simple 'back of the envelope' calculations, plus the need for cautious experimental studies will be illustrated. The theme of the ageing of materials, central to this paper, finds applications in many industrial situations. (author)

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)

Irradiation probe and laboratory for irradiated material evaluation

International Nuclear Information System (INIS)

Smutny, S.; Kupca, L.; Beno, P.; Stubna, M.; Mrva, V.; Chmelo, P.

1975-09-01

The survey and assessment are given of the tasks carried out in the years 1971 to 1975 within the development of methods for structural materials irradiation and of a probe for the irradiation thereof in the A-1 reactor. The programme and implementation of laboratory tests of the irradiation probe are described. In the actual reactor irradiation, the pulse tube length between the pressure governor and the irradiation probe is approximately 20 m, the diameter is 2.2 mm. Temperature reaches 800 degC while the pressure control system operates at 20 degC. The laboratory tests (carried out at 20 degC) showed that the response time of the pressure control system to a stepwise pressure change in the irradiation probe from 0 to 22 at. is 0.5 s. Pressure changes were also studied in the irradiation probe and in the entire system resulting from temperature changes in the irradiation probe. Temperature distribution in the body of the irradiation probe heating furnace was determined. (B.S.)

Comprehensive nuclear materials

CERN Document Server

Allen, Todd; Stoller, Roger; Yamanaka, Shinsuke

2012-01-01

Comprehensive Nuclear Materials encapsulates a panorama of fundamental information on the vast variety of materials employed in the broad field of nuclear technology. The work addresses, in five volumes, 3,400 pages and over 120 chapter-length articles, the full panorama of historical and contemporary international research in nuclear materials, from Actinides to Zirconium alloys, from the worlds' leading scientists and engineers. It synthesizes the most pertinent research to support the selection, assessment, validation and engineering of materials in extreme nuclear environments. The work discusses the major classes of materials suitable for usage in nuclear fission, fusion reactors and high power accelerators, and for diverse functions in fuels, cladding, moderator and control materials, structural, functional, and waste materials.

ORNL capability to conduct post irradiation examination of full-length commercial nuclear fuel rods

International Nuclear Information System (INIS)

Spellman, Donald J.

2007-01-01

Hot cells at Oak Ridge National Laboratory (ORNL) are nearing completion of a multi-year upgrade program to implement 21. century capabilities to meet the examination demands for higher burnup fuels and the future demands that will come from fuel recycling programs. Fuel reliability and zero tolerance for fuel failure is more than an industry goal. Fuel reliability is becoming a requirement that supports the renaissance of nuclear power generation. Thus, fuel development and management of new forms of waste that will come from programs such as the Global Nuclear Energy Partnership (GNEP) will require extensive use of the flexible, high-quality, technically advanced hot cells at ORNL. ORNL has the capability to perform post irradiation examination (PIE) of irradiated commercial nuclear fuel rods and the management structure to ensure a timely, cost-effective result. ORNL can: 1) Handle the transportation issues, 2) Perform macroscopic fuel rod examinations, 3) Perform microscopic fuel and clad examinations, and 4) Manage legacy material and waste disposal issues from PIE activities. All four of these items will be managed in a way that allows the customer day-to-day access to the results and data. Hot cell examination equipment that is necessary to determine the characteristics and performance of irradiated materials must operate in a hostile environment and is subject to long-term degradation that may result in reliability and quality assurance (QA) issues. ORNL has modernized its hot cell nuclear fuel examination equipment, installing state-of-the-art automated examination equipment and data gathering capabilities. ORNL is planning a major commitment to nuclear fuel examination and development, and future improvements will continue to be made over the next few years. (author)

Opening of new field in material science and technology by materials irradiation research

Energy Technology Data Exchange (ETDEWEB)

Kurishita, Hiroaki [Tohoku Univ., Sendai (Japan). Inst. for Materials Research

1998-03-01

It is believed that high energy particle irradiation causes severe degradation of materials, and great efforts have been made to reveal the underlying mechanism of such degradation. However, recent progress of the developments of irradiation rigs performed in the Japan Materials Testing Reactor (JMTR) and materials fabrication techniques has enabled to change our understanding of radiation effects on materials from the above pessimistic one to the very challenging one, i.e., irradiation has the beneficial effect of producing new phenomena and/or innovative materials that will not be available without irradiation. An example to be noted is that irradiation with neutrons in JMTR greatly improved the ductility of less ductile metals. This ductility improvement due to irradiation is directly opposite to irradiation embrittlement and is called radiation induced ductilization (RIDU). In this presentation the significance of RIDU and its mechanism will be stated. (author)

Spent Fuel Working Group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

International Nuclear Information System (INIS)

1993-11-01

The Secretary of Energy's memorandum of August 19, 1993, established an initiative for a Department-wide assessment of the vulnerabilities of stored spent nuclear fuel and other reactor irradiated nuclear materials. A Project Plan to accomplish this study was issued on September 20, 1993 by US Department of Energy, Office of Environment, Health and Safety (EH) which established responsibilities for personnel essential to the study. The DOE Spent Fuel Working Group, which was formed for this purpose and produced the Project Plan, will manage the assessment and produce a report for the Secretary by November 20, 1993. This report was prepared by the Working Group Assessment Team assigned to the Hanford Site facilities. Results contained in this report will be reviewed, along with similar reports from all other selected DOE storage sites, by a working group review panel which will assemble the final summary report to the Secretary on spent nuclear fuel storage inventory and vulnerability

Research on new materials development and elucidation of degradation phenomena in nuclear environments

International Nuclear Information System (INIS)

Terai, Takayuki

2003-01-01

Advanced materials technologies are required to develop the pressure boundary materials used in nuclear corrosive environments under irradiation. The outline of the materials research for overcoming the requirements is described. The applicability and fusibility of new materials technology and major characteristics are shown with respect to several examples. A new-type coating method of diamond-like carbide films by using RF magnetron sputtering, mechanistic analysis on tritium release behavior from molten lithium-tin alloy in-reactor, neutron irradiation assisted change in pinning properties of Bi 2 Sr 2 CaCu 2 O 8+x single crystals are discussed as typical outputs concerning the research. (author)

Irradiated film material and method of the irradiation

International Nuclear Information System (INIS)

1978-01-01

The irradiation of polymer film material is a strengthening procedure. To obtain a substantial uniformity in the radiation dosage profile, the film is irradiated in a trough having lateral deflection blocks adjacent to the film edges. These deflect the electrons towards the surface of the trough bottom for further deflection towards the film edge. (C.F.)

ATF Neutron Irradiation Program Technical Plan

Energy Technology Data Exchange (ETDEWEB)

Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

2016-03-01

The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post-irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.

Gamma irradiation tests of concrete material recommended for storage casks of spent nuclear fuel arising from Cernavoda NPP

International Nuclear Information System (INIS)

Dulama, M.; Deneanu, N.; Dulama, C.; Baboescu, E.

2001-01-01

Considerable effort is being devoted to the Romania's Nuclear Spent Fuel and Waste Management R and D Program to develop engineered barriers for the containment of nuclear fuel waste under conditions of deep geological disposal. Engineering practice suggests that the concrete should fulfil the requirements of long term physical stability and resistance to radiation. With an appropriate system of metal reinforcement, it should be possible to obtain the tensile and impact strength required, avoiding the risk of mechanical damage during handling and emplacement. In accordance with the concept developed by CITON-Bucharest, presently, the dry storage of spent nuclear fuel is thought by two choices: - The alternative of dry storage type MMB3; - The alternative of dry storage type TRANSTOR. By using ORIGEN and PELSHIE computer codes, we evaluated the gamma radiation dose absorbed by the concrete walls of the storage vault both in MMB3 and in TRANSTOR designing variants. The irradiation tests were performed at the Gamma Irradiation Facility of the Institute for Nuclear Research. (authors)

Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

Science.gov (United States)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2017-10-01

Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic

Post-irradiation examination and R and D programs using irradiated fuels at KAERI

International Nuclear Information System (INIS)

Chun, Yong Bum; Min, Duck Kee; Kim, Eun Ka and others

2000-12-01

This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

Post-irradiation examination and R and D programs using irradiated fuels at KAERI

International Nuclear Information System (INIS)

Chun, Yong Bum; So, Dong Sup; Lee, Byung Doo; Lee, Song Ho; Min, Duck Kee

2001-09-01

This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

The effect of neutron irradiation on the structure and properties of carbon-carbon composite materials

International Nuclear Information System (INIS)

Burchell, T.D.; Eatherly, W.P.; Robbins, J.M.; Strizak, J.P.

1991-01-01

Carbon-based materials are an attractive choice for fusion reactor plasma facing components (PFCs) because of their low atomic number, superior thermal shock resistance, and low neutron activation. Next generation plasma fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER), will require advanced carbon-carbon composite materials possessing extremely high thermal conductivity to manage the anticipated severe heat loads. Moreover, ignition machines such as ITER will produce high neutron fluxes. Consequently, the influence of neutron damage on the structure and properties of carbon-carbon composite materials must be evaluated. Data from an irradiation experiment are reported and discussed here. Fusion relevant graphite and carbon-carbon composites were irradiated in a target capsule in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). A peak damage dose of 1.59 dpa at 600 degrees C was attained. The carbon materials irradiated included nuclear graphite grade H-451 and one-, two-, and three-directional carbon-carbon composite materials. Dimensional changes, thermal conductivity and strength are reported for the materials examined. The influence of fiber type, architecture, and heat treatment temperature on properties and irradiation behavior are reported. Carbon-Carbon composite dimensional changes are interpreted in terms of simple microstructural models

High energy electron irradiation of flowable materials

International Nuclear Information System (INIS)

Offermann, B.P.

1975-01-01

In order to efficiently irradiate a flowable material with high energy electrons, a hollow body is disposed in a container for the material and the material is caused to flow in the form of a thin layer across a surface of the body from or to the interior of the container while the material flowing across the body surface is irradiated. (U.S.)

Effects of the temperature and the irradiation on the behaviour of chlorine 37 in nuclear graphite: consequences on the mobility of chlorine 36 in irradiated graphites

International Nuclear Information System (INIS)

Blondel, Antoine

2013-01-01

This thesis deals with the studies of the management of irradiated graphite wastes issued from the dismantling of the UNGG French reactors. This work focuses on the behavior of 36 Cl. This radionuclide is mainly issued through the neutron activation of 35 Cl by the reaction 35 Cl(n, γ) 36 Cl, pristine chlorine being an impurity of nuclear graphite, present at the level of some at.ppm. 36 Cl is a long lived radionuclide (about 300,000 years) and is highly soluble in water and mobile in concrete and clay. The solubilization of 36 Cl is controlled by the water accessibility into irradiated graphite pores as well as by factors related to 36 Cl itself such as its chemical speciation and its location within the irradiated graphite. Both speciation and chlorine location should strongly influence its behaviour and need to be taken into account for the choice of liable management options. However, data on radioactive chlorine features are difficult to assess in irradiated graphite and are mainly related to detection sensitivity problems. In this context, we simulated and evaluated the impact of the temperature, the irradiation and the radiolytic oxidation on the chlorine 36 behaviour. In order to simulate the presence of 36 Cl, we implanted 37 Cl into virgin nuclear graphite. Ion implantation has been widely used to study the lattice location, the diffusion and the release of fission and activation products in nuclear materials. Our results on the comparative effects of the temperature and the irradiation show that chlorine occurs in irradiated graphite on temperature and electronic and nuclear irradiation improve this effect. (author)

Sterilization by gamma irradiation

International Nuclear Information System (INIS)

Reyes Frias, L.

1992-01-01

Since 1980 the National Institute of Nuclear Research counts with an Industrial Gamma Irradiator, for the sterilization of raw materials and finished products. Through several means has been promoted the use of this technology as alternative to conventional methods of sterilization as well as steam treatment and ethylene oxide. As a result of the made promotion this irradiator has come to its saturation limit being the sterilization irradiation one of the main services that National Institute of Nuclear Research offers to producer enterprises of disposable materials of medical use also of raw materials for the elaboration of cosmetic products and pharmaceuticals as well as dehydrated foods. It is presented the trend to the sterilization service by irradiation showed by the compilation data in a survey made by potential customers. (Author)

Technical limitations of nuclear fuel materials and structures

International Nuclear Information System (INIS)

Hansson, L.; Planman, T.; Vitikainen, E.

1993-05-01

This report gives a summary of the tasks carried out within the project 'Technical limitations of nuclear fuel materials and structures' which belongs to the Finnish national research programme called 'Systems behaviour and operational aspects of safety'. The duration of the project was three years from 1990 to 1992. Most western LWR utilities, including the two Finnish ones have an incentive to implement extended burnup fuel cycles in their nuclear power plants. The aim of this project has been authorities to support them in the assessment and licensing of new fuel designs and materials. The research work of the project was focused on collecting and qualifying fuel performance data and on performing laboratory tests on fresh and irradiated cladding and structural materials. Moreover, knowledge of the high burnup phenomena was obtained through participation in international research projects such as OECD Halden Project and several Studsvik projects. Experimental work within the framework of the VVER fuel cooperative effort was also continued. (orig.)

Special irradiation techniques

International Nuclear Information System (INIS)

Colomez, Gerard; Veyrat, J.F.

1981-01-01

Irradiation trials conducted on materials-testing reactors should provide a better understanding of the phenomena which characterize the working and evolution in time of electricity-generating nuclear reactors. The authors begin by outlining the objectives behind experimental irradiation (applied to the various nuclear chains) and then describe the special techniques deployed to achieve these objectives [fr

Analysis of nuclear material flow for experimental DUPIC fuel fabrication process at DFDF

International Nuclear Information System (INIS)

Lee, H. H.; Park, J. J.; Shin, J. M.; Lee, J. W.; Yang, M. S.; Baik, S. Y.; Lee, E. P.

1999-08-01

This report describes facilities necessary for manufacturing experiment for DUPIC fuel, manufacturing process and equipment. Nuclear material flows among facilities, in PIEF and IMEF, for irradiation test, for post examination of DUPIC fuel, for quality control, for chemical analysis and for treatment of radioactive waste have been analyzed in details. This may be helpful for DUPIC project participants and facility engineers working in related facilities to understand overall flow for nuclear material and radioactive waste. (Author). 14 refs., 15 tabs., 41 figs

Analysis of nuclear material flow for experimental DUPIC fuel fabrication process at DFDF

Energy Technology Data Exchange (ETDEWEB)

Lee, H. H.; Park, J. J.; Shin, J. M.; Lee, J. W.; Yang, M. S.; Baik, S. Y.; Lee, E. P

1999-08-01

This report describes facilities necessary for manufacturing experiment for DUPIC fuel, manufacturing process and equipment. Nuclear material flows among facilities, in PIEF and IMEF, for irradiation test, for post examination of DUPIC fuel, for quality control, for chemical analysis and for treatment of radioactive waste have been analyzed in details. This may be helpful for DUPIC project participants and facility engineers working in related facilities to understand overall flow for nuclear material and radioactive waste. (Author). 14 refs., 15 tabs., 41 figs.

Evaluation and development of advanced nuclear materials: IAEA activities

International Nuclear Information System (INIS)

Inozemtsev, V.; Basak, U.; Killeen, J.; Dyck, G.; Zeman, A.; )

2011-01-01

Economical, environmental and non-proliferation issues associated with sustainable development of nuclear power bring about a need for optimization of fuel cycles and implementation of advanced nuclear systems. While a number of physical and design concepts are available for innovative reactors, the absence of reliable materials able to sustain new challenging irradiation conditions represents the real bottle-neck for practical implementation of these promising ideas. Materials performance and integrity are key issues for the safety and competitiveness of future nuclear installations being developed for sustainable nuclear energy production incorporating fuel recycling and waste transmutation systems. These systems will feature high thermal operational efficiency, improved utilization of resources (both fissile and fertile materials) and reduced production of nuclear waste. They will require development, qualification and deployment of new and advanced fuel and structural materials with improved mechanical and chemical properties combined with high radiation and corrosion resistance. The extensive, diverse, and expensive efforts toward the development of these materials can be more effectively organized within international collaborative programmes with wide participation of research, design and engineering communities. IAEA carries out a number of international projects supporting interested Member States with the use of available IAEA program implementation tools (Coordinated Research Projects, Technical Meetings, Expert Reviews, etc). The presentation summarizes the activities targeting material developments for advanced nuclear systems, with particular emphasis on fast reactors, which are the focal topics of IAEA Coordinated Research Projects 'Accelerator Simulation and Theoretical Modelling of Radiation Effects' (on-going), 'Benchmarking of Structural Materials Pre-Selected for Advanced Nuclear Reactors', 'Examination of advanced fast reactor fuel and core

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

Nuclear Materials Characterization in the Materials and Fuels Complex Analytical Hot Cells

International Nuclear Information System (INIS)

Rodriquez, Michael

2009-01-01

As energy prices skyrocket and interest in alternative, clean energy sources builds, interest in nuclear energy has increased. This increased interest in nuclear energy has been termed the 'Nuclear Renaissance'. The performance of nuclear fuels, fuels and reactor materials and waste products are becoming a more important issue as the potential for designing new nuclear reactors is more immediate. The Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Analytical Laboratory Hot Cells (ALHC) are rising to the challenge of characterizing new reactor materials, byproducts and performance. The ALHC is a facility located near Idaho Falls, Idaho at the INL Site. It was built in 1958 as part of the former Argonne National Laboratory West Complex to support the operation of the second Experimental Breeder Reactor (EBR-II). It is part of a larger analytical laboratory structure that includes wet chemistry, instrumentation and radiochemistry laboratories. The purpose of the ALHC is to perform analytical chemistry work on highly radioactive materials. The primary work in the ALHC has traditionally been dissolution of nuclear materials so that less radioactive subsamples (aliquots) could be transferred to other sections of the laboratory for analysis. Over the last 50 years though, the capabilities within the ALHC have also become independent of other laboratory sections in a number of ways. While dissolution, digestion and subdividing samples are still a vitally important role, the ALHC has stand alone capabilities in the area of immersion density, gamma scanning and combustion gas analysis. Recent use of the ALHC for immersion density shows that extremely fine and delicate operations can be performed with the master-slave manipulators by qualified operators. Twenty milligram samples were tested for immersion density to determine the expansion of uranium dioxide after irradiation in a nuclear reactor. The data collected confirmed modeling analysis with very tight

Accountability of Radioactive Materials in Malaysian Nuclear Agency

International Nuclear Information System (INIS)

Noor Fadilla Ismail; Wan Saffiey Wan Abdullah; Khairuddin Mohamad Kontol; Azimawati Ahmad; Suzilawati Muhd Sarowi; Mohd Fazlie Abdul Rashid

2016-01-01

Radioactive materials possessed in Malaysian Nuclear Agency have many beneficial applications for research and development, calibration, tracer and irradiation. There are two types of radioactive materials which consist of sealed sourced and unsealed sourced shall be accounted for and secured at all the times by following the security aspect. The Health Physics Group in the Department of Radiation Safety and Health Division is responsible to manage the issues related to any accountability for all radioactive material purchased or received under the radioactive material protocol. The accountability of radioactive materials in Malaysian Nuclear Agency is very important to ensure the security and control the radioactive materials to not to be lost or fall into the hands of people who do not have permission to possess or use it. The accountability of radioactive materials considered as a mandatory to maintaining accountability by complying the requirements of the Atomic Energy Licensing Act 1984 (Act 304) and regulations made thereunder and the conditions of license LPTA / A / 724. In this report describes the important element of accountability of radioactive materials in order to enhances security standard by allowing tracking of the locations of sources and to reduce the risk of radioactive materials falling into the wrong hands. (author)

Materials behaviour in PWRs core

International Nuclear Information System (INIS)

Barbu, A.; Massoud, J.P.

2008-01-01

Like in any industrial facility, the materials of PWR reactors are submitted to mechanical, thermal or chemical stresses during particularly long durations of operation: 40 years, and even 60 years. Materials closer to the nuclear fuel are submitted to intense bombardment of particles (mainly neutrons) coming from the nuclear reactions inside the core. In such conditions, the damages can be numerous and various: irradiation aging, thermal aging, friction wear, generalized corrosion, stress corrosion etc.. The understanding of the materials behaviour inside the cores of reactors in operation is a major concern for the nuclear industry and its long term forecast is a necessity. This article describes the main ways of materials degradation without and under irradiation, with the means used to foresee their behaviour using physics-based models. Content: 1 - structures, components and materials: structure materials, nuclear materials; 2 - main ways of degradation without irradiation: thermal aging, stress corrosion, wear; 3 - main ways of degradation under irradiation: microscopic damaging - point defects, dimensional alterations, evolution of mechanical characteristics under irradiation, irradiation-assisted stress corrosion cracking (IASCC), synergies; 4 - forecast of materials evolution under irradiation using physics-based models: primary damage - fast dynamics, primary damage annealing - slow kinetics microstructural evolution, impact of microstructural changes on the macroscopic behaviour, insight on modeling methods; 5 - materials change characterization techniques: microscopic techniques - direct defects observation, nuclear techniques using a particle beam, global measurements, mechanical characterizations; 6 - perspectives. (J.S.)

Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

Science.gov (United States)

Norman, Eric B [Oakland, CA; Prussin, Stanley G [Kensington, CA

2009-01-27

A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

Minimizing material damage using low temperature irradiation

International Nuclear Information System (INIS)

Craven, E.; Hasanain, F.; Winters, M.

2012-01-01

Scientific advancements in healthcare driven both by technological breakthroughs and an aging and increasingly obese population have lead to a changing medical device market. Complex products and devices are being developed to meet the demands of leading edge medical procedures. Specialized materials in these medical devices, including pharmaceuticals and biologics as well as exotic polymers present a challenge for radiation sterilization as many of these components cannot withstand conventional irradiation methods. The irradiation of materials at dry ice temperatures has emerged as a technique that can be used to decrease the radiation sensitivity of materials. The purpose of this study is to examine the effect of low temperature irradiation on a variety of polymer materials, and over a range of temperatures from 0 °C down to −80 °C. The effectiveness of microbial kill is also investigated under each of these conditions. The results of the study show that the effect of low temperature irradiation is material dependent and can alter the balance between crosslinking and chain scission of the polymer. Low temperatures also increase the dose required to achieve an equivalent microbiological kill, therefore dose setting exercises must be performed under the environmental conditions of use. - Highlights: ► A study is performed to quantify low temperature irradiation effects on polymer materials and BIs. ► Low temperature irradiation alters the balance of cross-linking and chain scissoning in polymers. ► Low temperatures provide radioprotection for BIs. ► Benefits of low temperatures are application specific and must be considered when dose setting.

Irradiation plant for flowable material

International Nuclear Information System (INIS)

Bosshard, E.

1975-01-01

The irradiation plant can be used to treat various flowable materials including effluent or sewage sludge. The plant contains a concrete vessel in which a partition is mounted to form two coaxial irradiation chambers through which the flowable material can be circulated by means of an impeller. The partition can be formed to house tubes of radiation sources and to provide a venturi-like member about the impeller. The operation of the impeller is reversed periodically to assure movement of both heavy and light particles in the flow. (U.S.)

Diffusion, Thermal Properties and Chemical Compatibilities of Select MAX Phases with Materials For Advanced Nuclear Systems

Energy Technology Data Exchange (ETDEWEB)

Barsoum, Michel [Drexel Univ., Philadelphia, PA (United States); Bentzel, Grady [Drexel Univ., Philadelphia, PA (United States); Tallman, Darin J. [Drexel Univ., Philadelphia, PA (United States); Sindelar, Robert [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Garcia-Diaz, Brenda [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hoffman, Elizabeth [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-04-04

The demands of Gen IV nuclear power plants for long service life under neutron irradiation at high temperature are severe. Advanced materials that would withstand high temperatures (up to 1000+ ºC) to high doses in a neutron field would be ideal for reactor internal structures and would add to the long service life and reliability of the reactors. The objective of this work is to investigate the chemical compatibility of select MAX with potential materials that are important for nuclear energy, as well as to measure the thermal transport properties as a function of neutron irradiation. The chemical counterparts chosen for this work are: pyrolytic carbon, SiC, U, Pd, FLiBe, Pb-Bi and Na, the latter 3 in the molten state. The thermal conductivities and heat capacities of non-irradiated MAX phases will be measured.

Device for separating, purifying and recovering nuclear fuel material, impurities and materials from impurity-containing nuclear fuel materials or nuclear fuel containing material

International Nuclear Information System (INIS)

Sato, Ryuichi; Kamei, Yoshinobu; Watanabe, Tsuneo; Tanaka, Shigeru.

1988-01-01

Purpose: To separate, purify and recover nuclear fuel materials, impurities and materials with no formation of liquid wastes. Constitution: Oxidizing atmosphere gases are introduced from both ends of a heating furnace. Vessels containing impurity-containing nuclear fuel substances or nuclear fuel substance-containing material are continuously disposed movably from one end to the other of the heating furnace. Then, impurity oxides or material oxides selectively evaporated from the impurity-containing nuclear fuel substances or nuclear fuel substance-containing materials are entrained in the oxidizing atmosphere gas and the gases are led out externally from a discharge port opened at the intermediate portion of the heating furnace, filters are disposed to the exit to solidify and capture the nuclear fuel substances and traps are disposed behind the filters to solidify and capture the oxides by spontaneous air cooling or water cooling. (Sekiya, K.)

Steels and welding nuclear

International Nuclear Information System (INIS)

Sessa, M.; Milella, P.P.

1987-01-01

This ENEA Data-Base regards mechanical properties, chemical composition and heat treatments of nuclear pressure vessel materials: type A533-B, A302-B, A508 steel plates and forgings, submerged arc welds and HAZ before and after nuclear irradiation. Irradiation experiments were generally performed in high flux material test reactors. Data were collected from international available literature about water nuclear reactors pressure vessel materials embrittlement

Development of materials for the fusion nuclear energy system

International Nuclear Information System (INIS)

Park, J. Y.; Kim, S. H.; Jang, J. S.; Kim, W. J.; Jung, C. H.; Jun, B. H.; Maeng, W. Y.; Kwon, J. H.; Kim, H. P.; Hong, J. H.

2005-01-01

A state of the art on the nuclear material development has been reviewed based on the each component of the Tokamak typed fusion reactor. The current status of the development of structural materials such as FM steels, ODS steels, vanadium alloys and SiCf/SiC composites are introduced. The application of Li-based ceramics as a ceramic breeder and W-based alloys and C/C composites as plasma facing components for the divertor were also investigated, respectively. Some evaluation methods and results of the computational material simulation for irradiation damages and the compatibility between materials and coolant are described. Additionally, the material related research activities of ITER and ITER TBM and the collaboration activities on fusion materials between Japan and USA are briefly summarized

MAT-DB - A database for nuclear energy related materials data

International Nuclear Information System (INIS)

Over, H.H.

2009-01-01

The web-enabled materials database (Mat-DB) of JRC-IE has a long-term history in storing materials test data resulting from European and international research projects. The database structure and the user-guidance has bee permanently updated improved and optimized. The database is implemented in the secure ODIN portal: https://odin.jrc.ec.europa.eu of JRC-IE. This architecture guarantees fast access to confidential and public data and documentation which are stored in an inter-related document management database (DoMa). It is a part of JRC's nuclear knowledge management. Mat-DB hosts the whole pool of IAEA surveillance data of reactor pressure vessel materials from different nuclear power plants of the member states. Mat-DB contains also thousands of European GEN IV reactor systems related R and D materials data which are an important basis for the evaluating and extrapolating design data for candidate materials and setting up design rules covering high temperature exposure, irradiation and corrosion. Those data and rules would match also fusion related components. Mat-DB covers thermo-mechanical and thermo-physical properties data of engineering alloys at low, elevated and high temperatures for base materials and joints, including irradiated materials for nuclear fission and fusion applications, thermal barrier coated materials for gas turbines and properties of corroded materials. The corrosion part refers to weight gain/loss data of high temperature exposed engineering alloys and ceramic materials. For each test type the database structure reflects international test standards and recommendations. Mat-DB features an extensive library of evaluation programs for web-enabled assessment of uniaxial creep, fatigue, crack growth and high temperature corrosion properties. Evaluations can be performed after data retrieval or independently of Mat-DB by transferring other materials data in a given format to the programs. The fast evaluation processes help the user to

Calculation simulation of equivalent irradiation swelling for dispersion nuclear fuel

International Nuclear Information System (INIS)

Cai Wei; Zhao Yunmei; Gong Xin; Ding Shurong; Huo Yongzhong

2015-01-01

The dispersion nuclear fuel was regarded as a kind of special particle composites. Assuming that the fuel particles are periodically distributed in the dispersion nuclear fuel meat, the finite element model to calculate its equivalent irradiation swelling was developed with the method of computational micro-mechanics. Considering irradiation swelling in the fuel particles and the irradiation hardening effect in the metal matrix, the stress update algorithms were established respectively for the fuel particles and metal matrix. The corresponding user subroutines were programmed, and the finite element simulation of equivalent irradiation swelling for the fuel meat was performed in Abaqus. The effects of the particle size and volume fraction on the equivalent irradiation swelling were investigated, and the fitting formula of equivalent irradiation swelling was obtained. The results indicate that the main factors to influence equivalent irradiation swelling of the fuel meat are the irradiation swelling and volume fraction of fuel particles. (authors)

Study of candu fuel elements irradiated in a nuclear power plant

International Nuclear Information System (INIS)

Ionescu, S.; Uta, O.; Mincu, M.; Anghel, D.; Prisecaru, I.

2015-01-01

The object of this work is the behaviour of CANDU fuel elements after service in nuclear power plant. The results are analysed and compared with previous result obtained on unirradiated samples and with the results obtained on samples irradiated in the TRIGA reactor of INR Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor, the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) of INR Pitesti on samples from fuel elements after they were removed out of the nuclear power plant: - dimensional and macrostructural characterization; - microstructural characterization by metallographic analyses; - determination of mechanical properties; - fracture surface analysis by scanning electron microscopy (SEM). A full set of non-destructive and destructive examinations concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the cladding was performed. The obtained results are typical for CANDU 6-type fuel. The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for the improvement of the CANDU fuel. (authors)

10 CFR 74.51 - Nuclear material control and accounting for strategic special nuclear material.

Science.gov (United States)

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Nuclear material control and accounting for strategic special nuclear material. 74.51 Section 74.51 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL NUCLEAR MATERIAL Formula Quantities of Strategic Special Nuclear...

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.

Spent Fuel Working Group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

International Nuclear Information System (INIS)

1993-11-01

A self assessment was conducted of those Hanford facilities that are utilized to store Reactor Irradiated Nuclear Material, (RINM). The objective of the assessment is to identify the Hanford inventories of RINM and the ES ampersand H concerns associated with such storage. The assessment was performed as proscribed by the Project Plan issued by the DOE Spent Fuel Working Group. The Project Plan is the plan of execution intended to complete the Secretary's request for information relevant to the inventories and vulnerabilities of DOE storage of spent nuclear fuel. The Hanford RINM inventory, the facilities involved and the nature of the fuel stored are summarized. This table succinctly reveals the variety of the Hanford facilities involved, the variety of the types of RINM involved, and the wide range of the quantities of material involved in Hanford's RINM storage circumstances. ES ampersand H concerns are defined as those circumstances that have the potential, now or in the future, to lead to a criticality event, to a worker radiation exposure event, to an environmental release event, or to public announcements of such circumstances and the sensationalized reporting of the inherent risks

10 CFR 36.69 - Irradiation of explosive or flammable materials.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Irradiation of explosive or flammable materials. 36.69... IRRADIATORS Operation of Irradiators § 36.69 Irradiation of explosive or flammable materials. (a) Irradiation... cause radiation overexposures of personnel. (b) Irradiation of more than small quantities of flammable...

Research on technology of evaluating thermal property data of nuclear power materials

International Nuclear Information System (INIS)

Imai, Hidetaka; Baba, Tetsuya; Matsumoto, Tsuyoshi; Kishimoto, Isao; Taketoshi, Naoyuki; Arai, Teruo

1997-01-01

For the materials of first wall and diverter of nuclear fusion reactor, in order to withstand steady and unsteady high heat flux load, excellent thermal characteristics are required. It is strongly demanded to measure such thermal property values as heat conductivity, heat diffusivity, specific heat capacity, emissivity and so using small test pieces up to higher than 2000degC. As the materials of nuclear reactors are subjected to neutron irradiation, in order to secure the long term reliability of the materials, it is very important to establish the techniques for forecasting the change of the thermal property values due to irradiation effect. Also the establishment of the techniques for estimating the thermal property values of new materials like low radioactivation material is important. In National Research Laboratory of Metrology, the research on the advancement of the measuring technology for high temperature thermal properties has resulted in the considerably successful development of such technologies. In this research, the rapid measurement of thermal property values up to superhigh temperature with highest accuracy, the making of thermal property data set of high level, the analysis and evaluation of the correlation of material characters and thermal property values, and the development of the basic techniques for estimating the thermal property values of solid materials are aimed at and advanced. These are explained. (K.I.)

LAMI - a planned Brazilian facility to investigate the mechanical and physical properties of structural materials under irradiation

International Nuclear Information System (INIS)

Andrade, Arnaldo H.P.; Lobo, Raquel M.

2011-01-01

The LAMI (Laboratorio de Materiais Irradiados) is a hot laboratory designed to the characterization of irradiated structural material and will constitute one of the main installations of the Brazilian Multipurpose Reactor (RMB). The strong points of LAMI are: to contribute, through theoretical and experimental investigations, to the development of knowledge in materials science in order to be able to predict the evolution of the physical and mechanical material properties under service conditions (irradiation, thermomechanical solicitation, influence of the environment, etc); to characterize the properties of the materials used in the nuclear industry in order to determine their performance and to be able to predict their life expectancy; to establish, maintain and make use of the database generated by these data and to provide expertise on industrial components, in particular to investigate strain or rupture mechanisms. The test materials can be irradiated or not, and originate from surveillance programs, experimental neutron irradiations or simulated irradiation with charged particles. The main line of LAMI will have 10 shielded hot cells. The building also will have an area dedicated to micro and nano structural materials analysis. The mechanical characterization to be carried out within LAMI includes mechanical tests on irradiated materials, comprehension of behavior and damage processes and the incorporation of the test data results in a data bank for capitalization of test results. Planned materials to be tested are going to be metallic alloys used in industrial and experimental reactor: pressure vessel steels, internal stainless steels, austeno-ferritic steels, zirconium alloys and aluminum alloys. (author)

Workshop on materials irradiation effects and applications 2012

International Nuclear Information System (INIS)

Xu, Qiu; Sato, Koichi; Yoshiie, Toshimasa

2013-01-01

For the study of the material irradiation effects, irradiation fields with improved control capabilities, advanced post irradiation experiments and well developed data analyses are required. This workshop aims to discuss new results and to plan the future irradiation research in the KUR. General meeting was held from December 14, 2012 to December 15, 2012 with 44 participants and 28 papers were presented. Especially recent experimental results using irradiation facilities in the KUR such as Materials Controlled Irradiation Facility, Low Temperature Loop and LINAC, and results of computer simulation, and fruitful discussions were performed. This volume contains the summary and selected transparencies presented in the meeting. (author)

Characterization of semiconductor and frontier materials by nuclear microprobe technology

International Nuclear Information System (INIS)

Zhu Jieqing; Li Xiaolin; Yang Changyi; Lu Rongrong; Wang Jiqing; Guo Panlin

2002-01-01

The nuclear microprobe technology is used to characterize the properties of semiconductor and other frontier materials at the stages of their synthesis, modification, integration and application. On the basis of the beam current being used, the analytical nuclear microprobe techniques being used in this project can be divided into two categories: high beam current (PIXE, RBS, PEB) or low beam current (IBIC, STIM) techniques. The material properties measured are the thickness and composition of a composite surface on a SiC ceramic, the sputtering-induced surface segregation and depth profile change in a Ag-Cu binary alloy, the irradiation effects on the CCE of CVD diamond, the CCE profile at a polycrystalline CVD diamond film and a GaAs diode at different voltage biases and finally, the characterization of individual sample on an integrated material chip. (author)

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

Present and future status of distributed database for nuclear materials (Data-Free-Way)

International Nuclear Information System (INIS)

Fujita, Mitsutane; Xu, Yibin; Kaji, Yoshiyuki; Tsukada, Takashi

2004-01-01

Data-Free-Way (DFW) is a distributed database for nuclear materials. DFW has been developed by three organizations such as National Institute for Materials Science (NIMS), Japan Atomic Energy Research Institute (JAERI) and Japan Nuclear Cycle Development Institute (JNC) since 1990. Each organization constructs each materials database in the strongest field and the member of three organizations can use these databases by internet. Construction of DFW, stored data, outline of knowledge data system, data manufacturing of knowledge note, activities of three organizations are described. On NIMS, nuclear reaction database for materials are explained. On JAERI, data analysis using IASCC data in JMPD is contained. Main database of JNC is experimental database of coexistence of engineering ceramics in liquid sodium at high temperature' and 'Tensile test database of irradiated 304 stainless steel' and 'Technical information database'. (S.Y.)

Self-organization in irradiated materials

International Nuclear Information System (INIS)

Gerasimenko, N.N.; Dzhamanbalin, K.K.; Medetov, N.A.

2003-01-01

Full text: By the present time a great deal of experimental material concerning self-organization in irradiated materials is stored. It means that in different materials (single crystal and amorphous semiconductor, metals, polymers) during one process of irradiation with accelerated particles or energetic quanta the structure previously disordered can be reordered to the previous or different order. These processes are considered separately from the processes of radiation-stimulated ordering when the renewal of the structure occurs as the result of extra irradiation, sometimes accompanied with another influence (heating, lighting, application of mechanical tensions). The processes of reordering are divided into two basic classes: the reconstruction of crystalline structure (1) and the formation of space-ordered system (2). The processes of ordering are considered with the use of synergetic approach and are analyzed conformably to the concrete conditions of new order appearance process realization in order to reveal the self-organization factor's role. The concrete experimental results of investigating of the radiation ordering processes are analyzed for different materials: semiconductor, metals, inorganic dielectrics, polymers. The ordering processes are examined from the point of their possible use in the technology of creating nano-dimensional structures general and quantum-dimensional ones in particular

Material science as basis for nuclear medicine: Holmium irradiation for radioisotopes production

Science.gov (United States)

Usman, Ahmed Rufai; Khandaker, Mayeen Uddin; Haba, Hiromitsu; Otuka, Naohiko

2018-05-01

Material Science, being an interdisciplinary field, plays important roles in nuclear science. These applications are seen in weaponry, armoured vehicles, accelerator structure and development, semiconductor detectors, nuclear medicine and many more. Present study presents the applications of some metals in nuclear medicine (radioisotope production). The charged-particle-induced nuclear reactions by using cyclotrons or accelerators have become a very vital feature of the modern nuclear medicine. Realising the importance of excitation functions for the efficient production of medical radionuclides, some very high purity holmium metals are generally prepared or purchased for bombardment in nuclear accelerators. In the present work, various methods to obtain pure holmium for radioisotope production have been discussed while also presenting details of our present studies. From the experimental work of the present studies, some very high purity holmium foils have been used in the work for a comprehensive study of residual radionuclides production cross-sections. The study was performed using a stacked-foil activation technique combined with γ-ray spectrometry. The stack was bombarded with 50.4 MeV alpha particle beam from AVF cyclotron of RI Beam Factory, Nishina Centre for Accelerator-Based Science, RIKEN, Japan. The work produced thulium radionuclides useful in nuclear medicine.

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.)

Hot cell works and related irradiation tests in fission reactor for development of new materials for nuclear application

International Nuclear Information System (INIS)

Shikama, Tatsuo

1999-01-01

Present status of research works in Oarai Branch, Institute for Materials Research, Tohoku University, utilizing Japan Materials Testing Reactor and related hot cells will be described.Topics are mainly related with nuclear materials studies, excluding fissile materials, which is mainly aiming for development of materials for advanced nuclear systems such as a nuclear fusion reactor. Conflict between traditional and routined procedures and new demands will be described and future perspective is discussed. (author)

Corrosion degradation of materials in nuclear reactors and its control

International Nuclear Information System (INIS)

Kain, Vivekanand

2016-01-01

As in every industry, nuclear industry also faces the challenge of corrosion degradation due to the exposure of the materials to the working environment. The aggressiveness of the environment is enhanced by the presence of radiation and high temperature and high-pressure environment. Radiation has influence on both the materials (changes in microstructure and microchemistry) and the aqueous environment (radiolysis producing oxidizing conditions). A survey of all the light water reactors in the world showed that stress corrosion cracking (SCC) and flow accelerated corrosion (FAC) account for more than two third of all the corrosion degradation cases. This paper visits these two forms of corrosion in nuclear power plants and illustrates cases from Indian nuclear power plants. Remedial measures against these two forms of corrosion that are possible to be employed and the actual measures employed in Indian nuclear power plants are discussed. Key features of SCC in different types of nuclear power plants are discussed. Main reasons for irradiation assisted stress corrosion cracking (IASCC) are presented and discussed. The signature patterns of single and dual phase FAC captured from components replaced from Indian nuclear power plants are presented. The development of a correlation between the scallop size and rate of single phase FAC - based on the database developed in Indian nuclear power plants is presented. Based on these two forms of degradation in nuclear reactors, design of materials that would resist these forms of degradation is presented. (author)

Materials of 15. autumn school on irradiated food

International Nuclear Information System (INIS)

1994-01-01

The ionizing radiation use for food preservation has been shown on the background of other methods. Several aspects connected with food irradiation have been discussed. Among them the legal aspects and recommendations have been performed. The healthy aspects from the view point of the radiolysis of main components of irradiated food have been presented. The broad review of physical, chemical and biological methods for identification of irradiated food products has been done. The accelerator pilot plant for food irradiation working at the Institute of Nuclear Chemistry and Technology, Warsaw, has been presented as well

Development of Micro-welding Technology of Cladding Tube with Temperature Sensor for Nuclear Fuel Irradiation Test

Energy Technology Data Exchange (ETDEWEB)

Kim, Soo Sung; Lee, C. Y.; Kim, W. K.; Lee, J. W.; Lee, D. Y

2006-01-15

Laser welding technology is widely used to fabricate some products of nuclear fuel in the nuclear industry. Especially, micro-laser welding is one of the key technology to be developed to fabricate precise products of fuel irradiation test. We have to secure laser welding technology to perform various instrumentations for fuel irradiation test. The instrumented fuel irradiation test at a research reactor is needed to evaluate the performance of the developed nuclear fuel. The fuel elements can be designed to measure the center line temperature of fuel pellets during the irradiation test by using temperature sensor. The thermal sensor was composed of thermocouple and sensor sheath. Micro-laser welding technology was adopted to seal between seal tube and sensor sheath with thickness of 0.15mm. The soundness of weld area has to be confirmed to prevent fission gas of the fuel from leaking out of the element during the fuel irradiation test. In this study, fundamental data for micro-laser welding technology was proposed to seal temperature sensor sheath of the instrumented fuel element. And, micro-laser welding for dissimilar metals between sensor sheath and seal tube was characterized by investigating welding conditions. Moreover, the micro-laser welding technology is closely related to advanced industry. It is expected that the laser material processing technology will be adopted to various applications in the industry.

Nuclear prehistory influence on irradiated metallic iron phase composition

International Nuclear Information System (INIS)

Alekseev, I.E.

2007-01-01

With application of different Moessbauer spectroscopy applications the phase composition of metallic iron after irradiation by both neutrons and charged particles were studied. Irradiation conditions, method of targets examination and phase composition of samples after irradiation were presented in tabular form. It is shown, that phase composition of irradiated metal is defined by nuclear prehistory. So, in a number of cases abnormals (stabilization of high- and low-temperature structural phases of iron at room temperature after irradiation end) were revealed

Minimizing material damage using low temperature irradiation

Science.gov (United States)

Craven, E.; Hasanain, F.; Winters, M.

2012-08-01

Scientific advancements in healthcare driven both by technological breakthroughs and an aging and increasingly obese population have lead to a changing medical device market. Complex products and devices are being developed to meet the demands of leading edge medical procedures. Specialized materials in these medical devices, including pharmaceuticals and biologics as well as exotic polymers present a challenge for radiation sterilization as many of these components cannot withstand conventional irradiation methods. The irradiation of materials at dry ice temperatures has emerged as a technique that can be used to decrease the radiation sensitivity of materials. The purpose of this study is to examine the effect of low temperature irradiation on a variety of polymer materials, and over a range of temperatures from 0 °C down to -80 °C. The effectiveness of microbial kill is also investigated under each of these conditions. The results of the study show that the effect of low temperature irradiation is material dependent and can alter the balance between crosslinking and chain scission of the polymer. Low temperatures also increase the dose required to achieve an equivalent microbiological kill, therefore dose setting exercises must be performed under the environmental conditions of use.

Use of the SPIRAL 2 facility for material irradiations with 14 MeV energy neutrons

International Nuclear Information System (INIS)

Mosnier, A.; Ridikas, D.; Ledoux, X.; Pellemoine, F.; Anne, R.; Huguet, Y.; Lipa, M.; Magaud, P.; Marbach, G.; Saint-Laurent, M.G.; Villari, A.C.C.

2005-01-01

The primary goal of an irradiation facility for fusion applications will be to generate a material irradiation database for the design, construction, licensing and safe operation of a fusion demonstration power station (e.g., DEMO). This will be achieved through testing and qualifying material performance under neutron irradiation that simulates service up to the full lifetime anticipated in the power plant. Preliminary investigations of 14 MeV neutron effects on different kinds of fusion material could be assessed by the SPIRAL 2 Project at GANIL (Caen, France), aiming at rare isotope beams production for nuclear physics research with first beams expected by 2009. In SPIRAL 2, a deuteron beam of 5 mA and 40 MeV interacts with a rotating carbon disk producing high-energy neutrons (in the range between 1 and 40 MeV) via C (d, xn) reactions. Then, the facility could be used for 3-4 months y -1 for material irradiation purposes. This would correspond to damage rates in the order of 1-2 dpa y -1 (in Fe) in a volume of ∼10 cm 3 . Therefore, the use of miniaturized specimens will be essential in order to effectively utilize the available irradiation volume in SPIRAL 2. Sample package irradiation temperature would be in the range of 250-1000 deg. C. The irradiation level of 1-2 dpa y -1 with 14 MeV neutrons (average energy) may be interesting for micro-structural and metallurgical investigations (e.g., mini-traction, small punch tests, etc.) and possibly for the understanding of specimen size/geometric effects of critical material properties. Due to the small test cell volume, sample in situ experiments are not foreseen. However, sample packages would be, if required, available each month after transfer in a special hot cell on-site

Austin: austenitic steel irradiation E 145-02 Irradiation Report

International Nuclear Information System (INIS)

Genet, F.; Konrad, J.

1987-01-01

Safety measures for nuclear reactors require that the energy which might be liberated in a reactor core during an accident should be contained within the reactor pressure vessel, even after very long irradiation periods. Hence the need to know the mechanical properties at high deformation velocity of structure materials that have received irradiation damage due to their utilization. The stainless steels used in the structures of reactors undergo damage by both thermal and fast neutrons, causing important changes in the mechanical properties of these materials. Various austenitic steels available as structural materials were irradiated or are under irradiation in various reactors in order to study the evolution of the mechanical properties at high deformation velocity as a function of the irradiation damage rate. The experiment called AUSTIN (AUstenitic STeel IrradiatioN) 02 was performed by the JRC Petten Establishment on behalf of Ispra in support of the reactor safety programme

Low cycle fatigue of irradiated LMFBR materials

International Nuclear Information System (INIS)

Blackburn, L.D.

1976-01-01

A review of low cycle fatigue data on irradiated LMFBR materials was conducted and extensive graphical representations of available data are presented. Representative postirradiation tensile properties of annealed 304 and 316 SS are selected and employed in several predictive methods to estimate irradiated material fatigue curves. Experimental fatigue data confirm the use of predictive methods for establishing conservative design curves over the range of service conditions relevant to such CRBRP components as core former, fixed radial shielding, core barrel, lower inlet module and upper internals structures. New experimental data on fatigue curves and creep-fatigue interaction in irradiated 20 percent cold worked (CW) 316 SS and Alloy 718 would support the design of removable radial shielding and upper internals in CRBRP. New experimental information on notched fatigue behavior and cyclic stress-strain curves of all these materials in the irradiated condition could provide significant design data

An analysis of irradiation creep in nuclear graphites

International Nuclear Information System (INIS)

Neighbour, G.B.; Hacker, P.J.

2002-01-01

Nuclear graphite under load shows remarkably high creep ductility with neutron irradiation, well in excess of any strain experienced in un-irradiated graphite (and additional to any dimensional changes that would occur without stress). As this behaviour compensates, to some extent, some other irradiation effects such as thermal shutdown stresses, it is an important property. This paper briefly reviews the approach to irradiation creep in the UK, described by the UK Creep Law. It then offers an alternative analysis of irradiation creep applicable to most situations, including HTR systems, using AGR moderator graphite as an example, to high values of neutron fluence, applied stress and radiolytic weight loss. (authors)

Safeguards for special nuclear materials

International Nuclear Information System (INIS)

Carlson, R.L.

1979-12-01

Safeguards, accountability, and nuclear materials are defined. The accuracy of measuring nuclear materials is discussed. The use of computers in nuclear materials accounting is described. Measures taken to physically protect nuclear materials are described

SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material Irradiation, Annealing and Re-Embrittlement

International Nuclear Information System (INIS)

Van Walle, E.; Chaouadi, R.; Puzzolante, J.L.; Fabry, A.; Van de Velde, J.

1998-01-01

The contribution of the Belgian Nuclear Research Centre SCK-CEN to the IAEA Round Robin Exercise on WWER-440 RPV weld material is reported. The objective of this contribution is twofold: (1) to gain experience in the field of the testing of WWER-440 steels; (2) to analyse the round-robin data according to in-house developed on used models in order to check their validity and applicability. Results from testing on unirradiated material are reported including data obtained from chemical analysis, Charpy-V impact testing, tensile testing and fracture toughness determination. Finally, irradiation strategies that can be used in the program to obtain irradiated, irradiated-annealed and irradiated-annealed-reirradiated conditions are outlined

AGC-2 Irradiation Report

Energy Technology Data Exchange (ETDEWEB)

Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-06-01

The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the

Thermodynamics of nuclear materials

International Nuclear Information System (INIS)

1979-01-01

Full text: The science of chemical thermodynamics has substantially contributed to the understanding of the many problems encountered in nuclear and reactor technology. These problems include reaction of materials with their surroundings and chemical and physical changes of fuels. Modern reactor technology, by its very nature, has offered new fields of investigations for the scientists and engineers concerned with the design of nuclear fuel elements. Moreover, thermodynamics has been vital in predicting the behaviour of new materials for fission as well as fusion reactors. In this regard, the Symposium was organized to provide a mechanism for review and discussion of recent thermodynamic investigations of nuclear materials. The Symposium was held in the Juelich Nuclear Research Centre, at the invitation of the Government of the Federal Republic of Germany. The International Atomic Energy Agency has given much attention to the thermodynamics of nuclear materials, as is evidenced by its sponsorship of four international symposia in 1962, 1965, 1967, and 1974. The first three meetings were primarily concerned with the fundamental thermodynamics of nuclear materials; as with the 1974 meeting, this last Symposium was primarily aimed at the thermodynamic behaviour of nuclear materials in actual practice, i.e., applied thermodynamics. Many advances have been made since the 1974 meeting, both in fundamental and applied thermodynamics of nuclear materials, and this meeting provided opportunities for an exchange of new information on this topic. The Symposium dealt in part with the thermodynamic analysis of nuclear materials under conditions of high temperatures and a severe radiation environment. Several sessions were devoted to the thermodynamic studies of nuclear fuels and fission and fusion reactor materials under adverse conditions. These papers and ensuing discussions provided a better understanding of the chemical behaviour of fuels and materials under these

Nuclear and activation characteristics of materials in 14.1-MeV and 2.5-MeV neutron field

International Nuclear Information System (INIS)

Seki, Yasushi; Takeyasu, Yuuichi.

1988-11-01

The nuclear and activation characteristics of various materials and elements of interest in terms of fusion reactor design are calculated and the results are graphically shown. The elements and materials are placed in a simple geometry modelling a blanket and shield of a fusion reactor. The neutrons with 14.1-MeV and 2.5-MeV energy are generated from the region represented as D-T and D-D plasma, respectively. The following activation characteristics after neutron irradiation are shown for each material and element; 1. Time evolution of induced activity, 2. Time evolution of decay heat, 3. Delayed gamma-ray dose distribution, 4. Decay heat distribution. In addition to the above activation characteristics, nuclear characteristics during the neutron irradiation, e.g. neutron energy spectra, neutron and gamma-ray flux distribution, nuclear heating distributions, and neutron and gamma-ray dose rate are also shown. (author)

Applications of Research Reactors Towards Research on Materials for Nuclear Fusion Technology. Proceedings of a Technical Meeting

International Nuclear Information System (INIS)

2013-11-01

Controlled nuclear fusion is widely considered to represent a nearly unlimited source of energy. Recent progress in the quest for fusion energy includes the design and current construction of the International Thermonuclear Experimental Reactor (ITER), for which a licence has recently been obtained as a first of its kind fusion nuclear installation. ITER is designed to demonstrate the scientific and technological feasibility of fusion energy production in excess of 500 MW for several consecutive minutes. ITER, however, will not be able to address all the nuclear fusion technology issues associated with the design, construction and operation of a commercial fusion power plant. The demonstration of an adequate tritium or fuel breeding ratio, as well as the development, characterization and testing of structural and functional materials in an integrated nuclear fusion environment, are examples of issues for which ITER is unable to deliver complete answers. To fill this knowledge gap, several facilities are being discussed, such as the International Fusion Materials Irradiation Facility and, eventually, a fusion demonstration power plant (DEMO). However, for these facilities, a vast body of preliminary research remains to be performed, for instance, concerning the preselection and testing of suitable materials able to withstand the high temperature and pressure, and intense radiation environment of a fusion reactor. Given their capacity for material testing in terms of available intense neutron fluxes, dedicated irradiation facilities and post-irradiation examination laboratories, high flux research reactors or material test reactors (MTRs) will play an indispensable role in the development of fusion technology. Moreover, research reactors have already achieved an esteemed legacy in the understanding of material properties and behaviour, and the knowledge gained from experiments in fission materials in certain cases can be applied to fusion systems, particularly those

Advances in nuclear fuel cycle materials and concepts. Vol. 1

Energy Technology Data Exchange (ETDEWEB)

El-Sayed, A A [Materials Division, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

This presentation gives an overview of the new trends in the materials used in various steps of the nuclear fuel cycle. This will cover fuels for various types of reactors (PWRs, HTRs, ... etc.) cladding materials, control rod materials, reactor structural materials, as well as materials used in the back end of the fuel cycle. Problems associated with corrosion of fuel cladding materials as well as those in control rod materials (B{sub 4} C swelling...etc.), and approaches for combating these influences are reviewed. For the case of reactor pressure vessel materials issues related to the influences of alloy composition, design approaches including the use of more forged parts and minimizing, as for as possible, longitudinal welds especially in the central region, are discussed. Furthermore the application of techniques for recovery of pre-irradiation mechanical properties of PVS components is also covered. New candidate materials for the construction of high level waste containers including modified types of stainless steel (high Ni and high MO), nickel-base alloys and titanium alloys are also detailed. Finally, nuclear fuel cycle concepts involving plutonium and actinides recycling shall be reviewed. 28 figs., 6 tabs.

Advances in nuclear fuel cycle materials and concepts. Vol. 1

International Nuclear Information System (INIS)

El-Sayed, A.A.

1996-01-01

This presentation gives an overview of the new trends in the materials used in various steps of the nuclear fuel cycle. This will cover fuels for various types of reactors (PWRs, HTRs, ... etc.) cladding materials, control rod materials, reactor structural materials, as well as materials used in the back end of the fuel cycle. Problems associated with corrosion of fuel cladding materials as well as those in control rod materials (B 4 C swelling...etc.), and approaches for combating these influences are reviewed. For the case of reactor pressure vessel materials issues related to the influences of alloy composition, design approaches including the use of more forged parts and minimizing, as for as possible, longitudinal welds especially in the central region, are discussed. Furthermore the application of techniques for recovery of pre-irradiation mechanical properties of PVS components is also covered. New candidate materials for the construction of high level waste containers including modified types of stainless steel (high Ni and high MO), nickel-base alloys and titanium alloys are also detailed. Finally, nuclear fuel cycle concepts involving plutonium and actinides recycling shall be reviewed. 28 figs., 6 tabs

Nuclear irradiation parameters of beryllium under fusion, fission and IFMIF irradiation conditions

International Nuclear Information System (INIS)

Fischer, U.; Chen, Y.; Leichtle, D.; Simakov, S.; Moeslang, A.; Vladimirov, P.

2004-01-01

A computational analysis is presented of the nuclear irradiation parameters for Beryllium under irradiation in typical neutron environments of fission and fusion reactors, and of the presently designed intense fusion neutron source IFMIF. The analysis shows that dpa and Tritium production rates at fusion relevant levels can be achieved with existing high flux fission reactors while the achievable Helium production is too low. The resulting He-Tritium and He/dpa ratios do not meet typical fusion irradiation conditions. Irradiation simulations in the medium flux test modules of the IFMIF neutron source facility were shown to be more suitable to match fusion typical irradiation conditions. To achieve sufficiently high production rates it is suggested to remove the creep-fatigue testing machine together with the W spectra shifter plate and move the tritium release module upstream towards the high flux test module. (author)

The regulations concerning refining business of nuclear source material and nuclear fuel materials

International Nuclear Information System (INIS)

1979-01-01

The regulations are provided for under the law for the regulations of nuclear source materials, nuclear fuel materials and reactors and provisions concerning refining business in the enforcement order for the law. The basic concepts and terms are defined, such as: exposure dose, accumulative dose; controlled area; inspected surrounding area and employee. Refining facilities listed in the application for designation shall be classified into clushing and leaching, thickning, refining facilities, storage facilities of nuclear source materials and nuclear fuel materials, disposal facilities of contaminated substances and building for refining, etc. Business program attached to the application shall include expected time of beginning of refining, estimated production amount of nuclear source materials or nuclear fuel materials for the first three years and funds necessary for construction, etc. Records shall be made and kept for particular periods on delivery and storage of nuclear source materials and nuclear fuel materials, control of radiation, maintenance and accidents of refining facilities. Safety securing, application of internationally regulated substances and measures in dangerous situations are stipulated respectively. Exposure dose of employees and other specified matters shall be reported by the refiner yearly to the Director General of Science and Technology Agency and the Minister of International Trade and Industry. (Okada, K.)

Post irradiation examinations on HTTR materials

International Nuclear Information System (INIS)

Sakai, Haruyuki; Ohmi, Masao; Eto, Motokuni; Watanabe, Katsutoshi

1995-01-01

The HTTR (High Temperature engineering Test Reactor) is being constructed at Oarai Research Establishment of the Japan Atomic Energy Research Institute. In order to develop necessary materials for the HTTR, after irradiations in the JMTR, PIEs are being carried out on these materials in the JMTRHL (JMTR Hot Laboratory). Impact test, tensile test, fatigue test, creep test, metallography and so on were performed for irradiated 2 1/4Cr 1Mo steel as the pressure vessel material and Alloy 800H as the cladding material of the control rod. A fatigue testing machine and four creep testing machines newly designed were fabricated and installed in the steel cells in order to evaluate the integrity of the HTTR materials. The development process and PIE results obtained with these machines are given in this paper

Gamma irradiation-induced modifications of resins found in nuclear waste embedding processes; Modifications induites par irradiation gamma dans les differentes resines rencontrees dans le traitement des dechets nucleaires

Energy Technology Data Exchange (ETDEWEB)

Allali, H [Faculte des Science et Techniques, Settat (Morocco); Debre, O; Lambert, M; Nsouli, B; Thomas, J P [Inst. de Physique Nucleaire, Lyon-1 Univ., 69 - Villeurbanne (France); Collaboration: IPN-Lyon, Laboratoire d` Etude des Materiaux Plastiques et des Biomateriaux, URA 507, UCBL, Agence Nationale pour la Gestion des Dechets Radiactifs

1999-12-31

The various resins involved in nuclear waste disposal are subject to gamma irradiation-induced modifications. From Time-of-Flight Mass Spectrometry of Secondary ions emitted under High Energy Projectiles bombardment (HSF-SIMS) of such materials the following results are obtained: the embedding epoxy resin DGEBA-DDM does not exhibit significant bulk changes in chemical structure, whatever the dose rate and irradiation medium (air or water), at least up to 2 MGy. However, oxidation processes are well observed at the very surface. Under the same conditions Ion exchange resins to be embedded are subjected to scissions of their functional sites, leading to fixed ion release. Dehydration under irradiation is observed pointing out the crucial role of water in ion transport outside of the material. (authors)

Current investigations of packaging materials used for food irradiation

International Nuclear Information System (INIS)

Fiszer, W.

1996-01-01

The article reviews current investigations of packaging materials applied for food irradiation. The increasing role of various synthetic materials is described. Author reviews radiation-induced damages in these materials. The article includes the list of materials accepted for food packaging and subsequent irradiation with different doses

Nuclear material control systems for nuclear power plants

International Nuclear Information System (INIS)

1975-06-01

Paragraph 70.51(c) of 10 CFR Part 70 requires each licensee who is authorized to possess at any one time special nuclear material in a quantity exceeding one effective kilogram to establish, maintain, and follow written material control and accounting procedures that are sufficient to enable the licensee to account for the special nuclear material in his possession under license. While other paragraphs and sections of Part 70 provide specific requirements for nuclear material control systems for fuel cycle plants, such detailed requirements are not included for nuclear power reactors. This guide identifies elements acceptable to the NRC staff for a nuclear material control system for nuclear power reactors. (U.S.)

Report Summarizing the Effort Required to Initiate Welding of Irradiated Materials within the Welding Cubicle

Energy Technology Data Exchange (ETDEWEB)

Frederick, Greg [Electric Power Research Institute (EPRI), Palo Alto, CA (United States); Sutton, Benjamin J. [Electric Power Research Institute (EPRI), Palo Alto, CA (United States); Tatman, Jonathan K. [Electric Power Research Institute (EPRI), Palo Alto, CA (United States); Vance, Mark Christopher [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Allen W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clark, Scarlett R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feng, Zhili [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chen, Jian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Tang, Wei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hu, Xunxiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gibson, Brian T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-06-01

The advanced welding facility within a hot cell at the Radiochemical Engineering Development Center of Oak Ridge National Laboratory (ORNL), which has been jointly funded by the U.S. Department of Energy (DOE), Office of Nuclear Energy, Light Water Reactor Sustainability Program and the Electric Power Research Institute, Long Term Operations Program and the Welding and Repair Technology Center, is in the final phase of development. Research and development activities in this facility will involve direct testing of advanced welding technologies on irradiated materials in order to address the primary technical challenge of helium induced cracking that can arise when conventional fusion welding techniques are utilized on neutron irradiated stainless steels and nickel-base alloys. This report details the effort that has been required since the beginning of fiscal year 2017 to initiate welding research and development activities on irradiated materials within the hot cell cubicle, which houses welding sub-systems that include laser beam welding (LBW) and friction stir welding (FSW) and provides material containment within the hot cell.

Mechanical Tests Plan after Neutron Irradiation for SMART SG Tube Materials in a Hot Cell

International Nuclear Information System (INIS)

Ahn, Sang Bok; Baik, Seung Jai; Kim, Do Sik; Yoo, Byung Ok; Jung, Yang Hong; Song, Woong Sub; Choo, Kee Nam; Park, Jin Seok; Lee, Yong Sun; Ryu, Woo Seog

2010-01-01

An advanced integral PWR, SMART (System- Integrated Modular Advanced ReacTor) is being developed in KAERI. It has compact size and a relatively small power rating compared to a conventional reactor. The main components such as the steam generators, main circulation pumps are located in the reactor vessel. Therefore they are damaged from neutron irradiations generated from nuclear fuel fissions during operation. The SMART SG tubes which are 17 mm in a diameter and 2.5 mm in a thickness will be made of Alloy 690. To ensure the operation safety the post irradiation examinations is necessary to evaluate the deterioration levels of various original properties. Specially the amount of mechanical properties change should be reflected and revised to design data. For that tensile, fracture, hardness test are planned and under preparations. In this paper the detailed plans are reviewed. Three kinds of materials having different heat treatment procedures are prepared to fabricate specimens. The capsules installed the specimens are going to be irradiated in HANARO. Finally the tests for them will be performed in IMEF, Irradiated Materials Examination Facility at KAERI

The regulations concerning refining business of nuclear source material and nuclear fuel materials

International Nuclear Information System (INIS)

1981-01-01

This rule is established under the provisions concerning refining business in the law concerning the regulation of nuclear raw materials, nuclear fuel materials and nuclear reactors and the ordinance for the execution of this law, and to enforce them. Basic terms are defined, such as: exposure radiation dose, cumulative dose, control area, surrounding monitoring area and worker. The application for the designation for refining business under the law shall be classified into the facilities for crushing and leaching-filtration, thikening, and refining, the storage facilities for nuclear raw materials and nuclear fuel materials, and the disposal facilities for radioactive wastes, etc. To the application, shall be attached business plans, the explanations concerning the technical abilities of applicants and the prevention of hazards by nuclear raw materials and nuclear fuel materials regarding refining facilities, etc. Records shall be made on the accept, delivery and stock of each kind of nuclear raw materials and nuclear fuel materials, radiation control, the maintenance of and accidents in refining facilities, and kept for specified periods, respectively. Security regulations shall be enacted for each works or enterprise on the functions and organizations of persons engaged in the control of refining facilities, the operation of the apparatuses which must be controlled for the prevention of accidents, and the establishment of control area and surrounding monitoring area, etc. The report on the usage of internationally regulated goods and the measures taken at the time of danger are defined particularly. (Okada, K.)

Application of the nuclear microprobe to the study of organic and inorganic composition of teeth irradiated by a laser beam

International Nuclear Information System (INIS)

Sommer, F.; Engelmann, Ch.; Couble, Ml.; Magloire, H.; Bonnin, P.

1986-01-01

The nuclear microprobe uses both direct observation of nuclear reactions induced by deuterons and X ray emission induced by protons or deuterons. Thanks to these techniques, concentration profiles of the main elements (C, N, P, Ca...) contained in different parts of healthy teeth (enamel, dentine and cementum) are drawn in control zones and laser irradiated zones. The results obtained show that important perturbations appear during the irradiation by the laser beam; we observe successively, depleted zones in carbon and nitrogen which contain calcium and phosphorus and hypomineralized zones which contain organic material. 10 refs [fr

X-irradiation-induced nuclear lesions in cultured mammaliam cells: an ultrastructural analysis

International Nuclear Information System (INIS)

Barham, S.S.; Walters, R.A.

1978-01-01

Electron-dense chromatin aggregates, hereafter referred to as lesions, have been characterized morphologically within interphase nuclei of Chinese hamster cells (line CHO) after a single acute exposure to 400, 800, 1200, or 2000 rad of x irradiation. At all doses studied, lesions were observed only after termination of radiation-induced division delay. Cell profiles were scored by electron microscopy for the presence or absence of nuclear lesions at various times after irradiation. The mitotic fraction from each irradiated population was also scored for each sample by light and electron microscopy. From these data and from simultaneous cell-density counts for each sample, it is apparent that postirradiation cell division is a prerequisite to formation of interphase nuclear lesions. Irradiated cell populations blocked in mitosis by Colcemid beyond the normal period of postirradiation division-delay failed to display nuclear lesions until after Colcemid was removed and cell division was completed. Enzyme digestions of isolated nuclei from irradiated cells with DNase I, RNase A, and Pronase suggest that the nuclear lesions are comprised primarily of chromatin. Nucleolar lesions, as well as various aberrant morphological forms of nucleoli, were also observed in cell populations after the onset of postirradiation cell division during the first 72 hr following exposure to irradiation. Delayed radiation-induced ultrastructural alterations of the nucleus included the formation of cytoplasmic invaginations into the nuclear space and inclusions of membranes within nuclei

Inhibition of EGFR nuclear shuttling decreases irradiation resistance in HeLa cells.

Science.gov (United States)

Wei, Hong; Zhu, Zijie; Lu, Longtao

2017-01-01

Cervical cancer is a leading cause of mortality in women worldwide. The resistance to irradiation at the advanced stage is the main reason for the poor prognosis and high mortality. This work aims to elucidate the molecular mechanism underlying the radio-resistance. In this study, we determined the pEGFR-T654 and pDNA-PK-T2609 expression level changes in irradiated HeLa cells treated with T654 peptide, a nuclear localization signal (NLS) inhibitor, to inhibit EGFR nuclear transport. Cell viability, cell cycle and migratory capacity were analyzed. Xenograft animal model was used to evaluate the effect of EGFR nuclear transport inhibition on the tumor growth in vivo. The enhanced translocation of nuclear EGFR in the irradiated HeLa cells correlated with the increasing level of pEGFR-T654 and pDNA-PK-T2609. Inhibition of EGFR nuclear translocation by NLS peptide inhibitor attenuated DNA damage repair in the irradiated HeLa cells, decreased cell viability and promoted cell death through arrest at G0 phase. NLS peptide inhibitor impaired the migratory capacity of irradiated HeLa cells, and negatively affected tumorigenesis in xenograft mice. This work puts forward a potential molecular mechanism of the irradiation resistance in cervical cancer cells, providing a promising direction towards an efficient therapy of cervical cancer.

10 CFR 74.41 - Nuclear material control and accounting for special nuclear material of moderate strategic...

Science.gov (United States)

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Nuclear material control and accounting for special nuclear material of moderate strategic significance. 74.41 Section 74.41 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL NUCLEAR MATERIAL Special Nuclear Material...

Early works on the nuclear microprobe for microelectronics irradiation tests at the CEICI (Sevilla, Spain)

International Nuclear Information System (INIS)

Palomo, F.R.; Morilla, Y.; Mogollon, J.M.; Garcia-Lopez, J.; Labrador, J.A.; Aguirre, M.A.

2011-01-01

Particle radiation effects are a fundamental problem in the use of numerous electronic devices for space applications, which is aggravated with the technology shrinking towards smaller and smaller scales. The suitability of low-energy accelerators for irradiation testing is being considered nowadays. Moreover, the possibility to use a nuclear microprobe, with a lateral resolution of a few microns, allows us to evaluate the behavior under ion irradiation of specific elements in an electronic device. The CEICI is the new CEnter for Integrated Circuits Irradiation tests, created into the facilities at the Centro Nacional de Aceleradores (CNA) in Sevilla-Spain. We have verified that our 3 MV Tandem accelerator, typically used for ion beam characterization of materials, is also a valuable tool to perform irradiation experiments in the low LET (Linear Energy Transfer) region.

Gamma irradiator

International Nuclear Information System (INIS)

Simonet, G.

1986-09-01

Fiability of devices set around reactors depends on material resistance under irradiation noticeably joints, insulators, which belongs to composition of technical, safety or physical incasurement devices. The irradiated fuel elements, during their desactivation in a pool, are an interesting gamma irradiation device to simulate damages created in a nuclear environment. The existing facility at Osiris allows to generate an homogeneous rate dose in an important volume. The control of the element distances to irradiation box allows to control this dose rate [fr

Study of fission products (Cs, Ba, Mo, Ru) behaviour in irradiated and simulated nuclear fuels during severe accidents using X-ray absorption Spectroscopy, SIMS and EPMA

International Nuclear Information System (INIS)

Geiger, Ernesto

2016-01-01

The identification of Fission Products (FP) release mechanism from irradiated nuclear fuels during a severe accident is of main importance for the development of codes for the estimation of the source-term (nature and quantity of radionuclides released into the environment). among the many FP Ba, Cs, Mo and Ru present a particular interest, since they may interact with each other or other elements and thus affect their release. In the framework of this thesis, two work axes have been set up in order to identify, firstly, the chemical phases initially present before the accident and, secondly, their evolution during the accident itself. The experimental approach consisted in reproducing nuclear severe accidents conditions at laboratory scale using both irradiated fuels and model materials (natural UO_2 doped with 12 FP). The advantage of these latter is the possibility of using characterization methods such as X-ray absorption Spectroscopy which are not available for irradiated fuels. Three irradiated fuel samples have been studied, representative to an initial state (before the accident), to an intermediate stage (1773 K) and to an advanced stage (2873 K) of a nuclear severe accident. Regarding to model materials, many accident sequences have been carried out, from 573 to 1973 K. Experimental results have allowed to establish a new release mechanism, considering both reducing and oxidizing conditions during an accident. These results have also demonstrated the importance of model materials as a complement to irradiated nuclear fuels in the study of nuclear severe accidents. (author) [fr

The sea transport of irradiated nuclear fuel

International Nuclear Information System (INIS)

Miller, M.L.

1995-01-01

The paper describes the development of a transport system dedicated to the sea transport of irradiated nuclear fuel. It reviews the background to why shipments were required and the establishment of a specialist shipping company, Pacific Nuclear Transport Limited. A description of the ships, flasks and other equipment utilized is provided, together with details of key procedures implemented to ensure safety and customer satisfaction

Multiscale modeling of radiation effects in nuclear reactor structural materials

Energy Technology Data Exchange (ETDEWEB)

Kwon, Junhyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Most problems in irradiated materials originate from the atomic collision of high-energy particles and lattice atoms. This collision leads to displacement cascades through the energy transfer reaction and causes various types of defects such as vacancies, interstitials, and clusters. The behavior of the point defects created in the displacement cascades is important because these defects play a major role in a microstructural evolution and further affect the changes in material properties. Rapid advances have been made in the computational capabilities for a realistic simulation of complex physical phenomena, such as irradiation and aging effects. At the same time, progress has been made in understanding the effect of radiation in metals, especially iron-based alloys. In this work, we present some of our ongoing work in this area, which illustrates a multiscale modeling for evaluating a microstructural evolution and mechanical property changes during irradiation. Multiscale modeling approaches are briefly presented here in the following order: nuclear interaction, atomic-level interaction, atomistic modeling, microstructural evolution modeling and mechanical property modeling. This is one of many possible methods for classifying techniques. The effort in developing physical multiscale models applied to radiation damage has been focused on a single crystal or single-grain materials.

Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

Energy Technology Data Exchange (ETDEWEB)

Phillpot, Simon; Tulenko, James

2011-09-08

The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

International Nuclear Information System (INIS)

Phillpot, Simon; Tulenko, James

2011-01-01

The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

Nuclear fuel and/or fertile material element suitable for non-destructive determination of burn-up

International Nuclear Information System (INIS)

Muench, E.

1976-01-01

The invention refers to a nuclear fuel and/or fertile material element suitable for non-destructive burn-up analysis, where an isotope or a mixture of isotopes capable of being activated is provided for measuring the intensity of radiation emitted from radioactive nuclides, especially the intensity of gamma rays. The half-life of radioactive decay of the isotope or the mixture mentioned above after being activated is sufficiently large compared with the irradiation of the fuel and/or fertile material element in the nuclear reactor. (orig.) [de

The use of isotopic correlation technique for determination of sup(241)Am and sup(243)Am concentration in nuclear irradiated fuels

International Nuclear Information System (INIS)

Souza Sarkis, J.E. de.

1990-01-01

In the last years the isotopic correlation technique is emerging as a powerful tool for the determination of concentration and isotopic composition of heavy nuclides in the nuclear fuel cycle. Accordingly, this technique has gained significant importance for the safeguard of the nuclear materials as well as for the accounting and build up of actinides elements in the irradiated nuclear fuels. In this work 42 isotopic correlations between the nuclides sup(241)Am and sup(243)Am and post irradiation isotopic data of 7 samples from fuel element BE-124 and 1 sample from fuel element BE-120 from the Obrigheim pressurized water nuclear power reactor, Federal Republic of Germany, were proposed. These isotopic correlations allowed to estimate the isotopic concentrations of sup(241)Am and sup(243)Am with an average deviation, relative to the experimental data obtained from isotopic dilution mass spectrometry technique, of 10%. These results are more precise than those found using the computer code ORIGEN 2 demonstrating the great potential of this technique for the determination of isotopic concentration and build up of those nuclides in irradiated nuclear fuels. The analytical and other experimental aspects of the post irradiation isotopic analysis of nuclear fuels are also discussed. (author)

General purpose nuclear irradiation chamber

International Nuclear Information System (INIS)

Nurul Fadzlin Hasbullah; Nuurul Iffah Che Omar; Nahrul Khair Alang Md Rashid; Jaafar Abdullah

2013-01-01

Nuclear technology has found a great need for use in medicine, industry, and research. Smoke detectors in our homes, medical treatments and new varieties of plants by irradiating its seeds are just a few examples of the benefits of nuclear technology. Portable neutron source such as Californium-252, available at Industrial Technology Division (BTI/ PAT), Malaysian Nuclear Agency, has a 2.645 year half-life. However, 252 Cf is known to emit gamma radiation from the source. Thus, this chamber aims to provide a proper gamma shielding for samples to distinguish the use of mixed neutron with gamma-rays or pure neutron radiation. The chamber is compatible to be used with other portable neutron sources such as 241 Am-Be as well as the reactor TRIGA PUSPATI for higher neutron dose. This chamber was designed through a collaborative effort of Kulliyyah Engineering, IIUM with the Industrial Technology Division (BTI) team, Malaysian Nuclear Agency. (Author)

Research status on radiation damage in nuclear materials and recommendations for IAEA activities. Technical report

International Nuclear Information System (INIS)

Caro, Alfredo; Caro, Magdalena

2002-03-01

This report addresses the synergy between the continuous progress of parallel computing and the spectacular advances in the theoretical framework that describes materials. Together, they contributed to significantly advance our comprehension of materials properties like mechanical behavior. It also highlights its impact on nuclear technology, as it provides physical insight into the complex processes responsible for the degradation of structural materials under neutron irradiation

Corrosion of copper-based materials in irradiated moist air systems

International Nuclear Information System (INIS)

Reed, D.T.; Van Konynenburg, R.A.

1991-06-01

The atmospheric corrosion of oxygen-free copper (CDA-102), 70/30 copper-nickel (CDA-715), and 7% aluminum bronze (CDA-613) in an irradiated moist air environment was investigated. Experiments were performed in both dry and 40% RH (at sign 90 degree C) air at temperatures of 90 and 150 degree C. Initial corrosion rates were determined based on a combination of weight gain and weight loss measurements. Corrosion products observed were identified. These experiments support efforts by the Yucca Mountain Project (YMP) to evaluate possible metallic barrier materials for nuclear waste containers. 8 refs., 1 fig., 2 tabs

UV-irradiation effects on polyester nuclear track detector

International Nuclear Information System (INIS)

Agarwal, Chhavi; Kalsi, P.C.

2010-01-01

The effects of UV irradiation (λ=254 nm) on polyester nuclear track detector have been investigated employing bulk-etch technique, UV-visible spectrophotometry and infra-red spectrometry (FTIR). The activation energy values for bulk-etching were found to decrease with the UV-irradiation time indicating the scission of the polymer. Not much shift in the absorption edge due to UV irradiation was seen in the UV-visible spectra. FTIR studies also indicate the scission of the chemical bonds, thereby further validating the bulk-etch rate results.

Applying RFID technology in nuclear materials management

International Nuclear Information System (INIS)

Tsai, H.; Chen, K.; Liu, Y.; Norair, J.P.; Bellamy, S.; Shuler, J.

2008-01-01

The Packaging Certification Program (PCP) of US Department of Energy (DOE) Environmental Management (EM), Office of Safety Management and Operations (EM-60), has developed a radio frequency identification (RFID) system for the management of nuclear materials. Argonne National Laboratory, a PCP supporting laboratory, and Savi Technology, a Lockheed Martin Company, are collaborating in the development of the RFID system, a process that involves hardware modification (form factor, seal sensor and batteries), software development and irradiation experiments. Savannah River National Laboratory and Argonne will soon field test the active RFID system on Model 9975 drums, which are used for storage and transportation of fissile and radioactive materials. Potential benefits of the RFID system are enhanced safety and security, reduced need for manned surveillance, real time access of status and history data, and overall cost effectiveness

Reactor Materials Research

Energy Technology Data Exchange (ETDEWEB)

Van Walle, E

2002-04-01

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

Reactor Materials Research

International Nuclear Information System (INIS)

Van Walle, E.

2002-01-01

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

Irradiation behavior of graphite shielding materials for FBR

International Nuclear Information System (INIS)

Maruyama, Tadashi; Kaito, Takeji; Onose, Shoji; Shibahara, Itaru

1994-01-01

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 597degC. Postirradiation examination was carried out on dimensional change, elastic modulus, and the thermal conductivity. The result of measurement of dimensional change indicated 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 to two to three times of unirradiated values. A large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependency on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, but the change in specific heat was negligibly small. 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. (author)

Neutron Flux Characterization of Irradiation Holes for Irradiation Test at HANARO

Directory of Open Access Journals (Sweden)

Yang Seong Woo

2016-01-01

Full Text Available The High flux Advanced Neutron Application ReactOr (HANARO is a unique research reactor in the Republic of Korea, and has been used for irradiation testing since 1998. To conduct irradiation tests for nuclear materials, the irradiation holes of CT and OR5 have been used due to a high fast-neutron flux. Because the neutron flux must be accurately calculated to evaluate the neutron fluence of irradiated material, it was conducted using MCNP. The neutron flux was measured using fluence monitor wires to verify the calculated result. Some evaluations have been conducted, however, more than 20% errors have frequently occurred at the OR irradiation hole, while a good agreement between the calculated and measured data was shown at the CT irradiation hole.

Report on the Workshop on Accelerated Nuclear Energy Materials Development

Energy Technology Data Exchange (ETDEWEB)

King, Wayne E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Allen, Todd [Univ. of Wisconsin, Madison, WI (United States); Arsenlis, Tom [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bench, Graham [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bulatov, Vasily [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fluss, Michael [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Klein, Richard [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McMahon, Donn [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Middleton, Carolin [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Morley, Maureen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Pasamehmetoglu, Kemal [Idaho National Lab. (INL), Idaho Falls, ID (United States); Turchi, Patrice [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

2010-05-11

This document reports on the Office of Nuclear Energy’s (NE’s) Workshop on Accelerated Nuclear Energy Materials Development held May 11, 2010, in Washington, DC. The purpose of the workshop was twofold: (1) to provide feedback on an initiative to use uncertainty quantification (UQ) to integrate theory, simulation, and modeling with accelerated experimentation to predict the behavior of materials and fuels in an irradiation environment and thereby accelerate the lengthy materials design and qualification process; and (2) to provide feedback on and refinement to five topical areas to develop predictive models for fuels and cladding and new radiation-tolerant materials. The goal of the workshop was to gather technical feedback with respect to the Office of Nuclear Energy’s research and development while also identifying and highlighting crosscutting capability and applicability of the initiative to other federal offices, including the Department of Energy’s (DOE’s) National Nuclear Security Administration (NNSA), Nuclear Regulatory Commission (NRC), DOE Office of Basic Energy Sciences (BES), DOE Office of Fusion Energy Sciences (FES), and Naval Reactors. The goals of the initiative are twofold: (1) develop time- and length-scale transcending models that predict material properties using UQ to effectively integrate theory, simulation, and modeling with accelerated experiments; and (2) design and develop new radiation-tolerant materials using the knowledge gained and methodologies created to shorten the development and qualification time and reduce cost. The initiative is crosscutting and has synergy with industry and other federal offices including Naval Reactors, NRC, FES, BES, and the Office of Advanced Scientific Computing Research (ASCR). It is distinguished by its use of uncertainty quantification to effectively integrate theory, simulation, and modeling with high-dose experimental capabilities. The initiative aims to bring the methodology that is being

Study of boron carbide evolution under neutron irradiation

International Nuclear Information System (INIS)

Simeone, D.

1999-01-01

Owing to its high neutron efficiency, boron carbide (B 4 C) is used as a neutron absorber in control rods of nuclear plants. Its behaviour under irradiation has been extensively studied for many years. It now seems clear that brittleness of the material induced by the 10 B(n,α) 7 Li capture reaction is due to penny shaped helium bubbles associated to a high strain field around them. However, no model explains the behaviour of the material under neutron irradiation. In order to build such a model, this work uses different techniques: nuclear microprobe X-ray diffraction profile analysis and Raman and Nuclear Magnetic Resonance Spectroscopy to present an evolution model of B 4 C under neutron irradiation. The use of nuclear reactions produced by a nuclear microprobe such as the 7 Li(p,p'γ) 7 Li reaction, allows to measure lithium profile in B 4 C pellets irradiated either in Pressurised Water Reactors or in Fast Breeder Reactors. Examining such profiles enables us to describe the migration of lithium atoms out of B 4 C materials under neutron irradiation. The analysis of X-ray diffraction profiles of irradiated B 4 C samples allows us to quantify the concentrations of helium bubbles as well as the strain fields around such bubbles.Furthermore Raman spectroscopy studies of different B 4 C samples lead us to propose that under neutron irradiation. the CBC linear chain disappears. Such a vanishing of this CBC chain. validated by NMR analysis, may explain the penny shaped of helium bubbles inside irradiated B 4 C. (author)

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.

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

Experimental evaluation of the neutrons flux of a irradiator with AmBe sources and its possibility of use in materials analysis

International Nuclear Information System (INIS)

Lima, Ruy Barros de

2003-01-01

This work had as a target to determine the irradiator thermal and over cadmium (epithermal and fast) neutrons flux , of the Nuclear Experimental Laboratory of the Nuclear Energy Center (CNEN) - IPEN, and the possibility of its use for Neutron Activation Analysis (NAA) by the absolute method. The neutrons flux quantification was performed indirectly by the gold naked and cadmium-covered foils activation technique. The neutrons flux was determined for two situations: with polyethylene block 5.0 cm thick and without the polyethylene block. The quantification of the elements present in the irradiated samples was obtained after the experimental determination of the incident neutrons flux in the irradiation position of the sample. Flux values along the irradiator axis were determined. Some materials were analyzed, presenting good agreement with reference values. (author)

Auditing nuclear materials statements

International Nuclear Information System (INIS)

Anon.

1973-01-01

A standard that may be used as a guide for persons making independent examinations of nuclear materials statements or reports regarding inventory quantities on hand, receipts, production, shipment, losses, etc. is presented. The objective of the examination of nuclear materials statements by the independent auditor is the expression of an opinion on the fairness with which the statements present the nuclear materials position of a nuclear materials facility and the movement of such inventory materials for the period under review. The opinion is based upon an examination made in accordance with auditing criteria, including an evaluation of internal control, a test of recorded transactions, and a review of measured discards and materials unaccounted for (MUF). The standard draws heavily upon financial auditing standards and procedures published by the American Institute of Certified Public Accountants

The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

Energy Technology Data Exchange (ETDEWEB)

Garrison, L. M., E-mail: garrisonlm@ornl.gov; Egle, B. J. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, Tennessee 37831 (United States); Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706 (United States); Zenobia, S. J.; Kulcinski, G. L.; Santarius, J. F. [Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706 (United States)

2016-08-15

The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA–500 μA; the typical current used is 72 μA, which is an average flux of 9 × 10{sup 14} ions/(cm{sup 2} s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

Protection and control of nuclear materials

International Nuclear Information System (INIS)

Jalouneix, J.; Winter, D.

2007-01-01

In the framework of the French regulation on nuclear materials possession, the first liability is the one of operators who have to know at any time the quantity, quality and localization of any nuclear material in their possession. This requires an organization of the follow up and of the inventory of these materials together with an efficient protection against theft or sabotage. The French organization foresees a control of the implementation of this regulation at nuclear facilities and during the transport of nuclear materials by the minister of industry with the sustain of the institute of radiation protection and nuclear safety (IRSN). This article presents this organization: 1 - protection against malevolence; 2 - national protection and control of nuclear materials: goals, administrative organization, legal and regulatory content (authorization, control, sanctions), nuclear materials protection inside facilities (physical protection, follow up and inventory, security studies), protection of nuclear material transports (physical protection, follow up), control of nuclear materials (inspection at facilities, control of nuclear material measurements, inspection of nuclear materials during transport); 3 - international commitments of France: non-proliferation treaty, EURATOM regulation, international convention on the physical protection of nuclear materials, enforcement in France. (J.S.)

Application of ion beams in materials science of radioactive waste forms: focus on the performance of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Garrido, Frederico [Centre de Spectrometrie Nucleaire et de Spectrometrie de Masse, CNRS-IN2P3-Universite Paris-Sud, Batiments 104-108, 91405 Orsay Campus (France)]. E-mail: garrido@csnsm.in2p3.fr; Nowicki, Lech [Andrzej Soltan Institute for Nuclear Studies, Hoza 69, 00-681 Warsaw (Poland); Thome, Lionel [Centre de Spectrometrie Nucleaire et de Spectrometrie de Masse, CNRS-IN2P3-Universite Paris-Sud, Ba-hat timents 104-108, 91405 Orsay Campus (France)

2005-10-15

Ion beam techniques provide unique tools for the qualification of radioactive waste forms. They address three major issues: (i) the simulation by ion irradiation of the stability of a matrix submitted to radiative environment; (ii) the doping of a material with stable or radioactive elements which simulate the species to be confined; (iii) the characterisation of a material via nuclear microanalysis techniques. Among various classes of nuclear matrices the spent nuclear fuel is widely considered as a potential candidate for the stabilisation of radioactive wastes in scenarios of long term interim storage or final geological disposal. Illustrative examples revealing the potentialities of the use of ion beams either as a pure characterisation tool - to investigate the chemical stability of the UO{sub 2} matrix under an oxygen potential - or in a combined way (e.g. irradiation/characterisation, doping/characterisation) - to explore the radiation stability and the behaviour of foreign species - are presented. Transformations (stoichiometry, depth and structure of growing hyperstoichiometric U{sub 4}O{sub 9}/U{sub 3}O{sub 7} oxides) occurring during low-temperature air oxidation of uranium dioxide single crystals are reported. Swift heavy ion irradiation of UO{sub 2} single crystals leads to a peculiar single crystal-polycrystal transformation (i.e. polygonisation of the fluorite-type structure of the material). Irradiation of UO{sub 2} at low energy shows that the damage production is directly linked to the energy deposited in nuclear elastic collisions. The lattice location of helium atoms (generated in large amount during the storage period) in interstitial octahedral positions is discussed.

The nuclear materials contraband

International Nuclear Information System (INIS)

Williams, P.; Woessner, P.

1996-01-01

Several seizures of nuclear materials carried by contraband have been achieved. Some countries or criminal organizations could manufacture atomic bombs and use them. This alarming situation is described into details. Only 40% of drugs are seized by the American police and probably less in western Europe. The nuclear materials market is smaller than the drugs'one but the customs has also less experience to intercept the uranium dispatch for instance more especially as the peddlers are well organized. A severe control of the international transports would certainly allow to seize a large part of nuclear contraband materials but some dangerous isotopes as uranium 235 or plutonium 239 are little radioactive and which prevents their detection by the Geiger-Mueller counters. In France, some regulations allow to control the materials used to manufacture the nuclear weapons, and diminish thus the risk of a nuclear materials contraband. (O.L.). 4 refs., 2 figs

Countermeasure technologies against materials deterioration of nuclear power plant components

International Nuclear Information System (INIS)

2004-09-01

This report was tentative safety standard on countermeasure technologies against materials deterioration of nuclear power plant components issued in 2004 on the base of the testing data obtained until March 2004, which was to be applied for technical evaluation for lifetime management of aged plants and preventive maintenance or repair of neutron irradiated components such as core shrouds and jet pumps. In order to prevent stress corrosion cracks (SCCs) of austenitic stainless steel welds of reactor components, thermal surface modification using laser beams was used on neutron irradiated materials with laser cladding or surface melting process methods by limiting heat input according to amount of accumulated helium so as to prevent crack initiation caused by helium bubble growth and coalescence. Laser cladding method of laser welding using molten sleeve set inside pipe surface to prevent SCCs of nickel-chromium-iron alloy welds, alloy 690 cladding method using tungsten inert gas (TIG) welding to prevent SCCs of nickel-chromium-iron alloy welds for dissimilar joints of pipes, and laser surface solid solution heat treatment method of laser irradiation on surfaces to prevent SCCs of austenitic stainless steel welds were also included as repair technologies. (T. Tanaka)

Microstructural examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

Srivastava, D.; Tewari, R.; Dey, G.K.; Sah, D.N.; Banerjee, S.

2005-01-01

Irradiation induced microstructural changes in Zr alloys strongly influence the creep, growth and mechanical properties of pressure tube material. Since dimensional changes and mechanical property degradation can limit the life of pressure tube, it is essential to study and develop an understanding of the microstructure produced by neutron irradiation, by examining samples taken from the irradiated components. In the present work, an effort has been made to examine, microstructure of the Zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water Reactor (PHWR). The present work is a first step towards a comprehensive program of characterization of microstructure of reactor materials after irradiation to different fluence levels in power reactors. In this study, samples from a Zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit 1, for a period for 6.77 effective full power years (EFPYs), have been prepared and examined. The samples selected from the tube are expected to have a cumulative radiation damage of about 3 dpa. Samples prepared from the off cuts of RAPS-1 pressure tubes were also studied for examining the unirradiated microstructure of the material. The samples were examined in a 200kV JEOL 2000 FX microscope. This paper presents the distinct features observed in irradiated sample and a comprehensive comparison of the microstructures of the unirradiated and irradiated material. The effect of annealing on the annihilation of the defects generated during irradiation has been also studied. The bright field micrographs revealed that microstructure of the irradiated samples was different in many respects from the microstructure of the unirradiated samples. The presence of defect structure in the form of loops etc could be seen in the irradiated sample. These loops were mostly c-type loops lying in the basal plane. The dissolution and redistribution of the precipitates were

Standard Reference Development of nuclear material for Tensile and Hardness Test Properties

International Nuclear Information System (INIS)

Choo, Y. S.; Kim, D. S.; Yoo, B. O.; Ahn, S. B.; Baik, S. J.; Chun, Y. B.; Kim, K. H.; Hong, K. P.; Ryu, W. S.

2007-12-01

Standard reference is a official approved data such a coefficient of physics, approved material properties, and etc., which should be analyzed and evaluated by scientific method to acquire official approval for accuracy and credibility of measured data and information. So it could be used broadly and continuously by various fields of nation and society. It is classified to effective standard reference, verified standard reference, and certified standard reference. There are sixteen fields in designated standard references such a physical chemistry field, material field, metal field, and the others. The standard reference of neutron irradiated nuclear structural material is classified to metal field. This report summarized the whole processes about data collection, data production, data evaluation and the suggestion of details evaluation technical standard for tensile and hardness properties, which were achieved by carry out the project 'nuclear material standard reference development' as a result

Standard Reference Development of nuclear material for Tensile and Hardness Test Properties

Energy Technology Data Exchange (ETDEWEB)

Choo, Y. S.; Kim, D. S.; Yoo, B. O.; Ahn, S. B.; Baik, S. J.; Chun, Y. B.; Kim, K. H.; Hong, K. P.; Ryu, W. S

2007-12-15

Standard reference is a official approved data such a coefficient of physics, approved material properties, and etc., which should be analyzed and evaluated by scientific method to acquire official approval for accuracy and credibility of measured data and information. So it could be used broadly and continuously by various fields of nation and society. It is classified to effective standard reference, verified standard reference, and certified standard reference. There are sixteen fields in designated standard references such a physical chemistry field, material field, metal field, and the others. The standard reference of neutron irradiated nuclear structural material is classified to metal field. This report summarized the whole processes about data collection, data production, data evaluation and the suggestion of details evaluation technical standard for tensile and hardness properties, which were achieved by carry out the project 'nuclear material standard reference development' as a result.

Lattice strain in irradiated materials unveils a prevalent defect evolution mechanism

Science.gov (United States)

Debelle, Aurélien; Crocombette, Jean-Paul; Boulle, Alexandre; Chartier, Alain; Jourdan, Thomas; Pellegrino, Stéphanie; Bachiller-Perea, Diana; Carpentier, Denise; Channagiri, Jayanth; Nguyen, Tien-Hien; Garrido, Frédérico; Thomé, Lionel

2018-01-01

Modification of materials using ion beams has become a widespread route to improve or design materials for advanced applications, from ion doping for microelectronic devices to emulation of nuclear reactor environments. Yet, despite decades of studies, major issues regarding ion/solid interactions are not solved, one of them being the lattice-strain development process in irradiated crystals. In this work, we address this question using a consistent approach that combines x-ray diffraction (XRD) measurements with both molecular dynamics (MD) and rate equation cluster dynamics (RECD) simulations. We investigate four distinct materials that differ notably in terms of crystalline structure and nature of the atomic bonding. We demonstrate that these materials exhibit a common behavior with respect to the strain development process. In fact, a strain build-up followed by a strain relaxation is observed in the four investigated cases. The strain variation is unambiguously ascribed to a change in the defect configuration, as revealed by MD simulations. Strain development is due to the clustering of interstitial defects into dislocation loops, while the strain release is associated with the disappearance of these loops through their integration into a network of dislocation lines. RECD calculations of strain depth profiles, which are in agreement with experimental data, indicate that the driving force for the change in the defect nature is the defect clustering process. This study paves the way for quantitative predictions of the microstructure changes in irradiated materials.

Characterization of Radiation-Induced Clustering using Atom Probe Tomography in Nuclear Structural Materials

Energy Technology Data Exchange (ETDEWEB)

Lee, Gyeong Geun; Lim, Sang Yeob; Chang, Kun Ok; Ha, Jin Hyung; Kwon, Jun Hyun [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The degradations include the change in mechanical properties, which are related to the microstructure evolution caused by irradiation. The most widely used tool for the imaging irradiated microstructure is transmission electron microscopy (TEM). The composition of irradiation defects can be analyzed using X-ray spectroscopy (EDS) equipped in the TEM. However, composition characterization of the nano-sized irradiation defects in the matrix is limited due to the beam broadening of TEM and the overlapping of the probed volume during EDS analysis. Recently, Atom probe tomography (APT) has been introduced to the characterization of irradiation defects. APT provides sub-nano scale position of atoms and the chemical composition of a selected volume. SS316 irradiated with Fe ions at above 300 .deg. C caused significant clustering and segregation of Si and Ni at defect sinks. The neutron irradiated low alloy steel showed similar clustering of Ni and Si. The approach of using APT was demonstrated to be well suited for discovering the structure of irradiation defects and performing quantitative analysis in nuclear materials irradiated at high temperature.

Control of Nuclear Materials and Special Equipment (Nuclear Safety Regulations)

International Nuclear Information System (INIS)

Cizmek, A.; Prah, M.; Medakovic, S.; Ilijas, B.

2008-01-01

Based on Nuclear Safety Act (OG 173/03) the State Office for Nuclear Safety (SONS) in 2008 adopted beside Ordinance on performing nuclear activities (OG 74/06) and Ordinance on special conditions for individual activities to be performed by expert organizations which perform activities in the area of nuclear safety (OG 74/06) the new Ordinance on the control of nuclear material and special equipment (OG 15/08). Ordinance on the control of nuclear material and special equipment lays down the list of nuclear materials and special equipment as well as of nuclear activities covered by the system of control of production of special equipment and non-nuclear material, the procedure for notifying the intention to and filing the application for a license to carry out nuclear activities, and the format and contents of the forms for doing so. This Ordinance also lays down the manner in which nuclear material records have to be kept, the procedure for notifying the State administration organization (regulatory body) responsible for nuclear safety by the nuclear material user, and the keeping of registers of nuclear activities, nuclear material and special equipment by the State administration organization (regulatory body) responsible for nuclear safety, as well as the form and content of official nuclear safety inspector identification card and badge.(author)

Irradiation damage behavior of low alloy steel wrought and weld materials

International Nuclear Information System (INIS)

Stofanak, R.J.; Poskie, T.J.; Li, Y.Y.; Wire, G.L.

1993-01-01

A study was undertaken to evaluate the irradiation damage response of several different types of low alloy steel: vintage type ASTM A302 Grade B (A302B) plates and welds containing different Ni and Cu concentrations, 3.5% Ni steels similar to ASTM A508 Class 4, welds containing about 1% Ni (similar to type 105S), and 3.5% Ni steels with ''superclean'' composition. All materials were irradiated at several different irradiation damage levels ranging from 0.0003 to 0.06 dpa at 232C (450F). Complete Charpy V-notch impact energy transition temperature curves were generated for all materials before and after irradiation to determine transition temperature at 4IJ (30 ft-lb) or 47J (35 ft-lb) and the upper shelf energy. Irradiation damage behavior was measured by shift in Charpy 41J or 47J transition temperature (ΔTT4 41J or ΔTT 47J ) and lowering of upper shelf Charpy energy at a given irradiation damage level. It was found that chemical composition greatly influenced irradiation damage behavior; highest irradiation damage (greatest ΔTT) was found in an A302B type weld containing 1.28% Ni and 0.20% Cu while the least damage was found in 3.5% Ni, 0.05% Cu, superclean wrought materials. Combination of Ni and Cu was found to affect irradiation damage behavior at higher irradiation damage levels in the A302B welds where the 1.28% Ni, 0.20% Cu weld showed more damage than a 0.60% Ni, 0.31% Cu weld. For the 3.5% Ni steels, fabrication influenced irradiation behavior in that a silicon (Si) killed material showed greater irradiation damage than a low silicon material. In general, the 3.5% Ni materials with low copper showed less irradiation damage than the A302B materials

Materials for cold neutron sources: Cryogenic and irradiation effects

International Nuclear Information System (INIS)

Alexander, D.J.

1990-01-01

Materials for the construction of cold neutron sources must satisfy a range of demands. The cryogenic temperature and irradiation create a severe environment. Candidate materials are identified and existing cold sources are briefly surveyed to determine which materials may be used. Aluminum- and magnesium-based alloys are the preferred materials. Existing data for the effects of cryogenic temperature and near-ambient irradiation on the mechanical properties of these alloys are briefly reviewed, and the very limited information on the effects of cryogenic irradiation are outlined. Generating mechanical property data under cold source operating conditions is a daunting prospect. It is clear that the cold source material will be degraded by neutron irradiation, and so the cold source must be designed as a brittle vessel. The continued effective operation of many different cold sources at a number of reactors makes it clear that this can be accomplished. 46 refs., 8 figs., 2 tab

Report of 6th research meeting on basic process of fuel cycle for nuclear fusion reactors, Yayoi Research Group; 3rd expert committee on research of nuclear fusion fuel material correlation basis

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-03-01

In this report, the lecture materials of Yayoi Research Group, 6th research meeting on basic process of fuel cycle for nuclear fusion reactors which was held at the University of Tokyo on March 25, 1996, are collected. This workshop was held also as 3rd expert committee on research of nuclear fusion fuel material correlation basis of Atomic Energy Society of Japan. This workshop has the character of the preparatory meeting for the session on `Interface effect in nuclear fusion energy system` of the international workshop `Interface effect in quantum energy system`, and 6 lectures and one comment were given. The topics were deuterium transport in Mo under deuterium ion implantation, the change of the stratum structure of graphite by hydrogen ion irradiation, the tritium behavior in opposing materials, the basic studies of the irradiation effects of solid breeding materials, the research on the behavior of hydroxyl group on the surface of solid breeding materials, the sweep gas effect on the surface of solid breeding materials, and the dynamic behavior of ion-implanted deuterium in proton-conductive oxides. (K.I.)

A spallation-based irradiation test facility for fusion and future fission materials

CERN Document Server

Samec, K; Kadi, Y; Luis, R; Romanets, Y; Behzad, M; Aleksan, R; Bousson, S

2014-01-01

The EU’s FP7 TIARA program for developing accelerator-based facilities has recently demonstrated the unique capabilities of a compact and powerful spallation source for irradiating advanced nuclear materials. The spectrum and intensity of the neutron flux produced in the proposed facility fulfils the requirements of the DEMO fusion reactor for ITER, ADS reactors and also Gen III / IV reactors. Test conditions can be modulated, covering temperature from 400 to 550°C, liquid metal corrosion, cyclical or static stress up to 500 MPa and neutron/proton irradiation damage of up to 25 DPA per annum. The entire “TMIF” facility fits inside a cube 2 metres on a side, and is dimensioned for an accelerator beam power of 100 kW, thus reducing costs and offering great versatility and flexibility.

Illicit diversion of nuclear materials

International Nuclear Information System (INIS)

Bett, F.L.

1975-08-01

This paper discusses the means of preventing illegal use of nuclear material by terrorists or other sub-national groups and by governments. With respect to sub-national groups, it concludes that the preventive measures of national safeguards systems, when taken together with the practical difficulties of using nuclear material, would make the diversion and illegal use of nuclear material unattractive in comparison with other avenues open to these groups to attain their ends. It notes that there are only certain areas in the nuclear fuel cycle, e.g. production of some types of nuclear fuel embodying highly enriched uranium and shipment of strategically significant nuclear material, which contain material potentially useful to these groups. It also discusses the difficult practical problems, e.g. coping with radiation, which would face the groups in making use of the materials for terrorist purposes. Concerning illegal use by Governments, the paper describes the role of international safeguards, as applied by the International Atomic Energy Agency, and the real deterrent effect of these safeguards which is achieved through the requirements to maintain comprehensive operating records of the use of nuclear material and by regular inspections to verify these records. The paper makes the point that Australia would not consider supplying nuclear material unless it were subject to international safeguards. (author)

Method for monitoring irradiated nuclear fuel using cerenkov radiation

International Nuclear Information System (INIS)

Caldwell, J.T.; Dowdy, E.J.; Nicholson, N.

1983-01-01

A method is provided for monitoring irradiated nuclear fuel inventories located in a water-filled storage pond wherein the intensity of the cerenkov radiation emitted from the water in the vicinity of the nuclear fuel is measured. This intensity is then compared with the expected intensity for nuclear fuel having a corresponding degree of irradiation exposure and time period after removal from a reactor core. Where the nuclear fuel inventory is located in an assembly having fuel pins or rods with intervening voids, the cerenkov light intensity measurement is taken at selected bright spots corresponding to the water-filled interstices of the assembly in the water storage, the waterfilled interstices acting as cerenkov light channels so as to reduce cross-talk. On-line digital analysis of an analog video signal is possible, or video tapes may be used for later measurement using a video editor and an electrometer. Direct measurement of the cerenkov radiation intensity also is possible using spot photometers pointed at the assembly

Study of the variations of the microhardness and electrical resistivity in nuclear materials

International Nuclear Information System (INIS)

Lucki, G.

1982-01-01

Data about the effect of neutron irradiation on the microhardness of stainless steel of type AISI 321 with 0.05; 0.10; 0.20; 0.50 and 1.00 wt % of Nb additions are presented with the purpose of contributing to the technology of fabrication and characterization of materials intended to perform in nuclear environments. The samples have been irradiated with fast neutrons (E > 1 MeV) inside IPEN's reactor core in argon atmosphere and temperature range of 200 - 600 0 C (isothermal annealings diring 7 hours) with fluences of 10 17 n/cm 2 . Identical isothermal annealings have been performed without irradiation. Present work also reports to Vickers microhardness of ZrNb (97.5 - 2.5 wt. %) alloy and zircalloys 2 and 4 before irradiation. (Author) [pt

The regulations concerning refining business of nuclear source material and nuclear fuel materials

International Nuclear Information System (INIS)

1987-01-01

Regulations specified here cover application for designation of undertakings of refining (spallation and eaching filtration facilities, thickening facilities, refining facilities, nuclear material substances or nuclear fuel substances storage facilities, waste disposal facilities, etc.), application for permission for alteration (business management plan, procurement plan, fund raising plan, etc.), application for approval of merger (procedure, conditions, reason and date of merger, etc.), submission of report on alteration (location, structure, arrangements processes and construction plan for refining facilities, etc.), revocation of designation, rules for records, rules for safety (personnel, organization, safety training for employees, handling of important apparatus and tools, monitoring and removal of comtaminants, management of radioactivity measuring devices, inspection and testing, acceptance, transport and storage of nuclear material and fuel, etc.), measures for emergency, submission of report on abolition of an undertaking, submission of report on disorganization, measures required in the wake of revocation of designation, submission of information report (exposure to radioactive rays, stolen or missing nuclear material or nuclear fuel, unusual leak of nuclear fuel or material contaminated with nuclear fuel), etc. (Nogami, K.)

Standard Guide for Packaging Materials for Foods to Be Irradiated

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 This guide provides a format to assist producers and users of food packaging materials in selecting materials that have the desirable characteristics for their intended use and comply with applicable standards or government authorizations. It outlines parameters that should be considered when selecting food-contact packaging materials intended for use during irradiation of prepackaged foods and it examines the criteria for fitness for their use. 1.2 This guide identifies known regulations and regulatory frameworks worldwide pertaining to packaging materials for holding foods during irradiation; but it does not address all regulatory issues associated with the selection and use of packaging materials for foods to be irradiated. It is the responsibility of the user of this guide to determine the pertinent regulatory issues in each country where foods are to be irradiated and where irradiated foods are distributed. 1.3 This guide does not address all of the food safety issues associated with the synergisti...

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.)

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

Study on the Standard Establishment for the Integrity Assessment of Nuclear Fuel Cladding Materials

Energy Technology Data Exchange (ETDEWEB)

Kang, S-S; Kim, S-H; Jung, Y-K; Yang, C-Y; Kim, I-G; Choi, Y-H; Kim, H-J; Kim, M-W; Rho, B-H [KINS, Daejeon (Korea, Republic of)

2008-02-15

Fuel cladding material plays important role as a primary structure under the high temperature, high pressure and neutron environment of nuclear power plant. According to this environment, cladding material can be experienced several type aging phenomena including the neutron irradiation embrittlement. On the other hand, although the early nuclear power plant was designed to fitting into the 40MWd/KgU burn-up, the currently power plant intends to go to the high burn-up range. In this case, the safety criteria which was established at low burn-up needs to conform the applicability at the high burn-up. In this study, the safety criteria of fuel cladding material was reviewed to assess the cladding material integrity, and the material characteristics of cladding were reviewed. The current LOCA criterial was also reviewed, and the basic study for re-establishment of LOCA criteria was performed. The time concept safety criteria was also discussed to prevent the breakaway oxidation. Through the this study, safety issues will be produced and be helpful for integrity insurance of nuclear fuel cladding material. This report is the final report.

Gamma irradiation technology for composite materials

International Nuclear Information System (INIS)

Romero, Guillermo R; Gonzalez, Maria E.

2003-01-01

A composite of sugar cane bagasse and low-density polyethylene was prepared. Gamma -radiation of Cobalt-60 (Co 60 ) and reactive additives were used, to make compatible the lignocellulosic fibers with the polymeric matrix. Gamma-radiation was applied in different stages with different purposes: a) Irradiation of cellulosic fibers treated or not with reactive additive, in presence of air, to produce macro radicals increasing their reactivity during extrusion with polyethylene. A homogeneous and fusible material resulted that can be used as raw material in thermoforming processes with cost in between that of its constitutive elements; b) Irradiation of final products, to produce the cross-linking of polymeric chains. The fibers remain trapped in the cross-linked matrix. A homogeneous and infusible material with high mechanical properties was obtained. (author)

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

Global nuclear material control model

International Nuclear Information System (INIS)

Dreicer, J.S.; Rutherford, D.A.

1996-01-01

The nuclear danger can be reduced by a system for global management, protection, control, and accounting as part of a disposition program for special nuclear materials. The development of an international fissile material management and control regime requires conceptual research supported by an analytical and modeling tool that treats the nuclear fuel cycle as a complete system. Such a tool must represent the fundamental data, information, and capabilities of the fuel cycle including an assessment of the global distribution of military and civilian fissile material inventories, a representation of the proliferation pertinent physical processes, and a framework supportive of national or international perspective. They have developed a prototype global nuclear material management and control systems analysis capability, the Global Nuclear Material Control (GNMC) model. The GNMC model establishes the framework for evaluating the global production, disposition, and safeguards and security requirements for fissile nuclear material

Radiation-thermal effects change of physico-mechanical properties in reactor materials irradiated with neutrons and energetic charged particles

International Nuclear Information System (INIS)

Hofman, A.

1999-01-01

In the first part of the report (chapter 1) the earlier results of the important scientific and technological investigations which were performed in the seventies years in Poland have been presented. They concerned the fabrication, corrosion, mechanical properties of materials for research and power reactors. Being of the general survey character, the chapter includes own, original results of research of thermal irradiation effects on microstructure evolution phase transformations and mechanical properties of reactor materials. The kinetics of isothermal transformation β→α in U-Cr 0.4% wt. alloy has been studied. Factors affecting stress-corrosion cracking of zirconium in iodine vapour have been investigated. The rings and loops for irradiation specimens and Hot Laboratory for postirradiation examination of construction materials is described. In the second part (chapters 2, 3, 4, 5) performed the investigations and simulations of radiation damage in metals by heavy ion beams (E > 1 MeV/a.m.n.) were described scientific base and technical problems of the method of irradiation of heavy ions and of the examination of irradiated samples is presented. It is followed by a summary of the results of simulation and reactor experiments on different materials. Radiation hardening of a number metals (Al, Zr, Cu, Ni, U) irradiated by heavy ion and neutrons, mechanical properties and microstructural evolution in ion and neutron irradiated austenitic stainless steel is described. The last chapter is a description of practical aspects of the presented studies in nuclear science and technology. (author)

The construction of irradiated material examination facility

International Nuclear Information System (INIS)

Ro, Seung Gy; Lee, Key Soon; Herr, Young Hoi

1990-03-01

A detail design of the examination process, the hot cell facility and the annexed facility of the irradiated material examination facility (IMEF) which will be utilized to examine and evaluate physical and mechanical properties of neutron-irradiated materials, has been performed. Also a start-up work of the underground structure construction has been launched out. The project management and tasks required for the license application were duly carried out. The resultant detail design data will be used for the next step. (author)

Absolute nuclear material assay

Science.gov (United States)

Prasad, Manoj K [Pleasanton, CA; Snyderman, Neal J [Berkeley, CA; Rowland, Mark S [Alamo, CA

2010-07-13

A method of absolute nuclear material assay of an unknown source comprising counting neutrons from the unknown source and providing an absolute nuclear material assay utilizing a model to optimally compare to the measured count distributions. In one embodiment, the step of providing an absolute nuclear material assay comprises utilizing a random sampling of analytically computed fission chain distributions to generate a continuous time-evolving sequence of event-counts by spreading the fission chain distribution in time.

Assessment of the gas dynamic trap mirror facility as intense neutron source for fusion material test irradiations

International Nuclear Information System (INIS)

Fischer, U.; Moeslang, A.; Ivanov, A.A.

2000-01-01

The gas dynamic trap (GDT) mirror machine has been proposed by the Budker Institute of nuclear physics, Novosibirsk, as a volumetric neutron source for fusion material test irradiations. On the basis of the GDT plasma confinement concept, 14 MeV neutrons are generated at high production rates in the two end sections of the axially symmetrical central mirror cell, serving as suitable irradiation test regions. In this paper, we present an assessment of the GDT as intense neutron source for fusion material test irradiations. This includes comparisons to irradiation conditions in fusion reactor systems (ITER, Demo) and the International Fusion Material Irradiation Facility (IFMIF), as well as a conceptual design for a helium-cooled tubular test assembly elaborated for the largest of the two test zones taking proper account of neutronics, thermal-hydraulic and mechanical aspects. This tubular test assembly incorporates ten rigs of about 200 cm length used for inserting instrumented test capsules with miniaturized specimens taking advantage of the 'small specimen test technology'. The proposed design allows individual temperatures in each of the rigs, and active heating systems inside the capsules ensures specimen temperature stability even during beam-off periods. The major concern is about the maximum achievable dpa accumulation of less than 15 dpa per full power year on the basis of the present design parameters of the GDT neutron source. A design upgrading is proposed to allow for higher neutron wall loadings in the material test regions

Contribution to the study of the effects of α-irradiation in nuclear glasses

International Nuclear Information System (INIS)

Abbas, A.

2001-01-01

The main topic of this work is to characterise the effects of α-disintegration in nuclear waste glasses. Experimental and numerical approaches have been considered. The structure of the French nuclear waste glass (R7T7) has been simulated using four- and six-oxides simplified glasses which contain the main elements of the R7T7 glass: SiO 2 , B 2 O 3 , Na 2 O, ZrO 2 , Al 2 O 3 and CaO. Four- and six-oxides glasses have been irradiated with 1 MeV-He + (ionisation) and 2.1 MeV-Kr 3+ (ionisation and atomic collisions) ions in order to reproduce the effects of the α-particle and of the recoil nucleus emitted during α-disintegration of actinides, and also to differentiate electronic and ballistic effects. Irradiated glasses have been characterised using several techniques, which have been adapted to the peculiarities of our samples (isolated material, small irradiated depth). The results point out the salient role of sodium in the observed modifications: depth concentration profiles obtained with RBS show an accumulation of sodium at the irradiated surface. We found a apparent acceleration of sodium release in leaching experiments which confirm that point. Modifications observed in Raman spectra of irradiated glasses show an increase of the polymerisation (increase of Q 3 /Q 2 ratio) due to sodium migration. In simplified glasses we have found that the modifications of mechanical properties by external irradiations reproduce the modifications observed in actinide doped nuclear glass (decrease of hardness and increase of fracture toughness). At the same time, we performed Molecular Dynamics simulations of a six-oxides glass. We have shown that the surface modifies the glass structure down to a depth of 10 Angstrom: modification of depth concentration profiles, decrease of the atomic coordination number (A1, B and Si). During cascades, we found that atomic displacements are easier near the surface. This behaviour is also observed when the glass is submitted to an

Methods for nuclear material control used in the basic production of a typical radiochemical plant

International Nuclear Information System (INIS)

Kositsyn, V.F.; Mukhortov, N.F.; Korovin, Yu.I.; Rudenko, V.S.; Petrov, A.M.

1999-01-01

Techniques for destructive and non-destructive assay of the component and isotopic composition of nuclear materials are described, namely gravimetric, titrimetric, coulometric, mass spectrometry, as well as those based on registration of neutron and γ radiations. Their metrologic characteristics are described. The techniques described are suggested to be used for nuclear material (NM) control and accounting purposes at the model radiochemical plant for processing irradiated fuel subassemblies from power reactors. The measurement control program is also described. This program is intended for the measurement quality assurance in the framework of NM control and accountancy system [ru

Irradiation damage

International Nuclear Information System (INIS)

Howe, L.M.

2000-01-01

There is considerable interest in irradiation effects in intermetallic compounds from both the applied and fundamental aspects. Initially, this interest was associated mainly with nuclear reactor programs but it now extends to the fields of ion-beam modification of metals, behaviour of amorphous materials, ion-beam processing of electronic materials, and ion-beam simulations of various kinds. The field of irradiation damage in intermetallic compounds is rapidly expanding, and no attempt will be made in this chapter to cover all of the various aspects. Instead, attention will be focused on some specific areas and, hopefully, through these, some insight will be given into the physical processes involved, the present state of our knowledge, and the challenge of obtaining more comprehensive understanding in the future. The specific areas that will be covered are: point defects in intermetallic compounds; irradiation-enhanced ordering and irradiation-induced disordering of ordered alloys; irradiation-induced amorphization

Irradiation damage

Energy Technology Data Exchange (ETDEWEB)

Howe, L.M

2000-07-01

There is considerable interest in irradiation effects in intermetallic compounds from both the applied and fundamental aspects. Initially, this interest was associated mainly with nuclear reactor programs but it now extends to the fields of ion-beam modification of metals, behaviour of amorphous materials, ion-beam processing of electronic materials, and ion-beam simulations of various kinds. The field of irradiation damage in intermetallic compounds is rapidly expanding, and no attempt will be made in this chapter to cover all of the various aspects. Instead, attention will be focused on some specific areas and, hopefully, through these, some insight will be given into the physical processes involved, the present state of our knowledge, and the challenge of obtaining more comprehensive understanding in the future. The specific areas that will be covered are: point defects in intermetallic compounds; irradiation-enhanced ordering and irradiation-induced disordering of ordered alloys; irradiation-induced amorphization.

Determination of internationally controlled materials according to provisions of the law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1984-01-01

The internationally controlled materials determined according to the law for nuclear source materials, etc. are the following: nuclear source materials, nuclear fuel materials, moderating materials, facilities including reactors, etc. sold, transferred, etc. to Japan according to the agreements for peaceful uses of atomic energy between Japan, and the United States, the United Kingdom, Canada, Australia and France by the respective governments and those organs under them; nuclear fuel materials resulting from usage of the above sold and transferred materials, facilities; nuclear fuel materials sold to Japan according to agreements set by the International Atomic Energy Agency; nuclear fuel materials involved with the safeguards in nuclear weapons non-proliferation treaty with IAEA. (Mori, K.)

Cladding tube materials for advanced nuclear facilities with closed fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Bartosova, I. [Slovenska technicka univerzita v Bratislave, Fakulta elektrotechniky a informatiky, Ustav jadroveho a fyzikalneho inzinierstva, 81219 Bratislava (Slovakia)

2013-04-16

The paper is aimed on perspective materials for fuel cladding in advanced nuclear reactors. Samples of Eurofer and ODS Eurofer were studied by various techniques such as Positron Annihilation Lifetime Spectroscopy, Vickers Hardness and Coincidence Doppler Broadening. After studying the samples by these methods, we implanted them by Helium atoms to simulate irradiation damage. Samples were then remeasured by Positron Annihilation Lifetime Spectroscopy to determine the affect of implantation on its behavior. (authors)

Disk-bend ductility tests for irradiated materials

International Nuclear Information System (INIS)

Klueh, R.L.; Braski, D.N.

1984-01-01

We modified the HEDL disk-bend test machine and are using it to qualitatively screen alloys that are susceptible to embrittlement caused by irradiation. Tests designed to understand the disk-bend test in relation to a uniaxial test are discussed. Selected results of tests of neutron-irradiated material are also presented

Post Irradiation Capabilities at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Schulthess, J.L.

2011-08-01

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) oversees the research, development, and demonstration activities that ensure nuclear energy remains a viable energy option for the United States. Fuel and material development through fabrication, irradiation, and characterization play a significant role in accomplishing the research needed to support nuclear energy. All fuel and material development requires the understanding of irradiation effects on the fuel performance and relies on irradiation experiments ranging from tests aimed at targeted scientific questions to integral effects under representative and prototypic conditions. The DOE recently emphasized a solution-driven, goal-oriented, science-based approach to nuclear energy development. Nuclear power systems and materials were initially developed during the latter half of the 20th century and greatly facilitated by the United States ability and willingness to conduct large-scale experiments. Fifty-two research and test reactors with associated facilities for performing fabrication and pre and post irradiation examinations were constructed at what is now Idaho National Laboratory (INL), another 14 at Oak Ridge National Laboratory (ORNL), and a few more at other national laboratory sites. Building on the scientific advances of the last several decades, our understanding of fundamental nuclear science, improvements in computational platforms, and other tools now enable technological advancements with less reliance on large-scale experimentation.

Post Irradiation Capabilities at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Schulthess, J.L.; Robert D. Mariani; Rory Kennedy; Doug Toomer

2011-08-01

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) oversees the research, development, and demonstration activities that ensure nuclear energy remains a viable energy option for the United States. Fuel and material development through fabrication, irradiation, and characterization play a significant role in accomplishing the research needed to support nuclear energy. All fuel and material development requires the understanding of irradiation effects on the fuel performance and relies on irradiation experiments ranging from tests aimed at targeted scientific questions to integral effects under representative and prototypic conditions. The DOE recently emphasized a solution-driven, goal-oriented, science-based approach to nuclear energy development. Nuclear power systems and materials were initially developed during the latter half of the 20th century and greatly facilitated by the United States’ ability and willingness to conduct large-scale experiments. Fifty-two research and test reactors with associated facilities for performing fabrication and pre and post irradiation examinations were constructed at what is now Idaho National Laboratory (INL), another 14 at Oak Ridge National Laboratory (ORNL), and a few more at other national laboratory sites. Building on the scientific advances of the last several decades, our understanding of fundamental nuclear science, improvements in computational platforms, and other tools now enable technological advancements with less reliance on large-scale experimentation.

A spallation-based irradiation test facility for fusion and future fission materials

International Nuclear Information System (INIS)

Samec, K.; Fusco, Y.; Kadi, Y.; Luis, R.; Romanets, Y.; Behzad, M.; Aleksan, R.; Bousson, S.

2014-01-01

The EU's FP7 TIARA program for developing accelerator-based facilities has recently demonstrated the unique capabilities of a compact and powerful spallation source for irradiating advanced nuclear materials. The spectrum and intensity of the neutron flux produced in the proposed facility fulfils the requirements of the proposed DEMO fusion reactor, ADS reactors and also Gen III / IV reactors. Test conditions can be modulated, covering temperature from 400 to 550 deg. C, liquid metal corrosion, cyclical or static stress up to 500 MPa and neutron/proton irradiation damage of up to 25 DPA per annum over a volume occupying one litre. The entire 'TMIF' facility fits inside a cube 2 metres on a side, and is dimensioned for an accelerator beam power of 100 kW, thus reducing costs and offering great versatility and flexibility. (authors)

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

The nuclear pore density in rat liver cells upon regeneration and total body X-ray irradiation

International Nuclear Information System (INIS)

Kuz'mina, S.N.; Troitskaya, L.P.; Mirkhamidova, P.A.; Bul'dyaeva, T.V.; Zbarskij, I.B.; Grigor'ev, V.B.; Akademiya Meditsinskikh Nauk SSSR, Moscow. Inst. Virusologii)

1979-01-01

The nuclear pore density has been investigated in rat liver cells in the course of regeneration and X-ray irradiation. It has been found that the number of pore complexes (PC) per nuclear shell (NS) unit area in the liver cells is not constant. In an hour following whole-body irradiation of rats with a regenerating liver at the 1200 R dose the number of PC per 1 μm 2 of the nuclear shell area decreases by 5, 8 times as compared with the PC density in the regenerating liver cells of the irradiated rats, the PC degradation and structural rupture being observed. It has been established by means of the freezing-etching method which enables PC surfaces observation as for cytoplasma as well as for nucleoplasma that the PC peripheral granulas and the central granula consist of subparticles being approximately of the same size. The central granula forms a channel through which the material containing RNA passes from the nucleus to the cytoplasma. On the basis of the fact that the treatement by Triton X-100, disarranging the integrity of the NS membranous structure, preserves PC in relation to the fibrous layer as well as on the basis of the unequal nuclear pore state observed on the platinum-carbon replicas from nuclei splits it is supposed that PC can be formed in the nucleus and then in the course of repening ''built in'' PS

Professional Nuclear Materials Management

International Nuclear Information System (INIS)

Forcella, A.A.; O'Leary, W.J.

1966-01-01

This paper describes the scope of nuclear materials management for a typical power reactor in the United States of America. Since this power reactor is financed by private capital, one of the principal obligations of the reactor operator is to ensure that the investment is protected and will furnish an adequate financial return. Because of the high intrinsic value of nuclear materials, appropriate security and accountability must be continually exercised to minimize losses beyond security and accountability for the nuclear materials. Intelligent forethought and planning must be employed to ensure that additional capital is not lost as avoidable additional costs or loss of revenue in a number of areas. The nuclear materials manager must therefore provide in advance against the following contingencies and maintain constant control or liaison against deviations from planning during (a) pre-reactor acquisition of fuel and fuel elements, (b) in-reactor utilization of the fuel elements, and (c) post-reactor recovery of fuel values. During pre-reactor planning and operations, it is important that the fuel element be designed for economy in manufacture, handling, shipping, and replaceability. The time schedule for manufacturing operations must minimize losses of revenue from unproductive dead storage of high cost materials. For in-reactor operations, the maximum achievable burn-up of the fissionable material must be obtained by means of appropriate fuel rearrangement schemes. Concurrently the unproductive down-time of the reactor for fuel rearrangement, inspections, and the like must be minimized. In the post-reactor period, when the fuel has reached a predetermined depletion of fissionable material, the nuclear materials manager must provide for the most economical reprocessing and recovery of fissionable values and by-products. Nuclear materials management is consequently an essential factor in achieving competitive fuel cycle and unit energy costs with power reactors

FMIT - the fusion materials irradiation test facility

International Nuclear Information System (INIS)

Liska, D.J.

1980-01-01

A joint effort by the Hanford Engineering Development Laboratory (HEDL) and Los Alamos Scientific Laboratory (LASL) has produced a preliminary design for a Fusion Materials Irradiation Test Facility (FMIT) that uses a high-power linear accelerator to fire a deuteron beam into a high-speed jet of molten lithium. The result is a continuous energy spectrum of neutrons with a 14-MeV average energy which can irradiate material samples to projected end-of-life levels in about 3 years, with a total accumulated fluence of 10 21 to 10 22 n/cm 2

Pressure vessel steels: influence of chemical composition on irradiation sensitivity

International Nuclear Information System (INIS)

Ghoniem, M.M.; Hammad, F.H.

1998-01-01

Neutron irradiation of the steels used in the construction of the nuclear reactor pressure vessels can lead to the embrittlement of these materials, increasing the ductile-to-brittle transition temperature and decreasing the fracture energy, which can limit the plant life. The knowledge of irradiation embrittlement and the means for minimizing such degradation is therefore important in the field of assuring the safety of the nuclear power plants. Irradiation embrittlement is quite a complex process. It involves many variables. The most important of these are irradiation temperature, neutron fluence (neutron dose), neutron flux (neutron dose rate), and chemical composition of the irradiated material. This paper is concerned with the effect of chemical composition, the role of residual and alloying elements in the irradiation embrittlement of nuclear reactor pressure vessel steels in light water reactors. It presents a critical review for the published work in this field through the last 25 years

Definition of Nuclear Material in Aspects of Nuclear Nonproliferation and Security

International Nuclear Information System (INIS)

Jeon, Ji Hye; Lee, Chan Suh

2014-01-01

Nuclear safety accidents directly affect human health but nuclear security incidents indirectly influence human, which demonstrates the reason why security receives less attention. However, it is acknowledged that nuclear terrorism is indeed one of the most dreadful threat humanity faces. As part of strengthening nuclear security as well as nonproliferation to response to the threat, we need a better understanding of the nuclear material which needs to be safe under the objective of nuclear security. In reality, practitioners implement safeguards and physical protection in compliance with the regulation text in domestic legislation. Thus, it is important to specify nuclear material clearly in law for effective implementation. Therefore, the definition of terminology related to nuclear material is explored herein, within the highest-level legislation on the safeguards and physical protection. First the definition in Korean legislation is analyzed. Then, so as to suggest some improvements, other international efforts are examined and some case studies are conducted on other states which have similar level of nuclear technology and industry to Korea. Finally, a draft of definition on nuclear material in perspective of nuclear nonproliferation and security is suggested based on the analysis below. The recommendation showed the draft nuclear material definition in nuclear control. The text will facilitate the understanding of nuclear material in the context of nuclear nonproliferation and security. It might provide appropriate provision for future legislation related to nuclear nonproliferation and security. For effective safeguards and physical protection measures, nuclear material should be presented with in a consistent manner as shown in the case of United Kingdom. It will be much more helpful if further material engineering studies on each nuclear material are produced. Multi-dimensional approach is required for the studies on the degree of efforts to divert

Definition of Nuclear Material in Aspects of Nuclear Nonproliferation and Security

Energy Technology Data Exchange (ETDEWEB)

Jeon, Ji Hye; Lee, Chan Suh [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

2014-10-15

Nuclear safety accidents directly affect human health but nuclear security incidents indirectly influence human, which demonstrates the reason why security receives less attention. However, it is acknowledged that nuclear terrorism is indeed one of the most dreadful threat humanity faces. As part of strengthening nuclear security as well as nonproliferation to response to the threat, we need a better understanding of the nuclear material which needs to be safe under the objective of nuclear security. In reality, practitioners implement safeguards and physical protection in compliance with the regulation text in domestic legislation. Thus, it is important to specify nuclear material clearly in law for effective implementation. Therefore, the definition of terminology related to nuclear material is explored herein, within the highest-level legislation on the safeguards and physical protection. First the definition in Korean legislation is analyzed. Then, so as to suggest some improvements, other international efforts are examined and some case studies are conducted on other states which have similar level of nuclear technology and industry to Korea. Finally, a draft of definition on nuclear material in perspective of nuclear nonproliferation and security is suggested based on the analysis below. The recommendation showed the draft nuclear material definition in nuclear control. The text will facilitate the understanding of nuclear material in the context of nuclear nonproliferation and security. It might provide appropriate provision for future legislation related to nuclear nonproliferation and security. For effective safeguards and physical protection measures, nuclear material should be presented with in a consistent manner as shown in the case of United Kingdom. It will be much more helpful if further material engineering studies on each nuclear material are produced. Multi-dimensional approach is required for the studies on the degree of efforts to divert

Nuclear materials management storage study

International Nuclear Information System (INIS)

Becker, G.W. Jr.

1994-02-01

The Office of Weapons and Materials Planning (DP-27) requested the Planning Support Group (PSG) at the Savannah River Site to help coordinate a Departmental complex-wide nuclear materials storage study. This study will support the development of management strategies and plans until Defense Programs' Complex 21 is operational by DOE organizations that have direct interest/concerns about or responsibilities for nuclear material storage. They include the Materials Planning Division (DP-273) of DP-27, the Office of the Deputy Assistant Secretary for Facilities (DP-60), the Office of Weapons Complex Reconfiguration (DP-40), and other program areas, including Environmental Restoration and Waste Management (EM). To facilitate data collection, a questionnaire was developed and issued to nuclear materials custodian sites soliciting information on nuclear materials characteristics, storage plans, issues, etc. Sites were asked to functionally group materials identified in DOE Order 5660.1A (Management of Nuclear Materials) based on common physical and chemical characteristics and common material management strategies and to relate these groupings to Nuclear Materials Management Safeguards and Security (NMMSS) records. A database was constructed using 843 storage records from 70 responding sites. The database and an initial report summarizing storage issues were issued to participating Field Offices and DP-27 for comment. This report presents the background for the Storage Study and an initial, unclassified summary of storage issues and concerns identified by the sites

Nuclear material operations manual

International Nuclear Information System (INIS)

Tyler, R.P.

1981-02-01

This manual provides a concise and comprehensive documentation of the operating procedures currently practiced at Sandia National Laboratories with regard to the management, control, and accountability of nuclear materials. The manual is divided into chapters which are devoted to the separate functions performed in nuclear material operations-management, control, accountability, and safeguards, and the final two chapters comprise a document which is also issued separately to provide a summary of the information and operating procedures relevant to custodians and users of radioactive and nuclear materials. The manual also contains samples of the forms utilized in carrying out nuclear material activities. To enhance the clarity of presentation, operating procedures are presented in the form of playscripts in which the responsible organizations and necessary actions are clearly delineated in a chronological fashion from the initiation of a transaction to its completion

Nuclear material operations manuals

International Nuclear Information System (INIS)

Tyler, R.P.

1979-06-01

This manual is intended to provide a concise and comprehensive documentation of the operating procedures currently practiced at Sandia Laboratories with regard to the management, control, and accountability of radioactive and nuclear materials. The manual is divided into chapters which are devoted to the separate functions performed in nuclear material operations-management, control, accountability, and safeguards, and the final two chapters comprise a document which is also issued separately to provide a summary of the information and operating procedures relevant to custodians and users of radioactive and nuclear materials. The manual also contains samples of the forms utilized in carrying out nuclear material activities. To enhance the clarity of presentation, operating procedures are presented in the form of playscripts in which the responsible organizations and necessary actions are clearly delineated in a chronological fashion from the initiation of a transaction to its completion

Selection of nuclear reactor coolant materials

International Nuclear Information System (INIS)

Shi Lisheng; Wang Bairong

2012-01-01

Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

Anti-Inflammatory Effect of ETAS®50 by Inhibiting Nuclear Factor-κB p65 Nuclear Import in Ultraviolet-B-Irradiated Normal Human Dermal Fibroblasts

Directory of Open Access Journals (Sweden)

Ken Shirato

2018-01-01

Full Text Available Ultraviolet (UV irradiation induces proinflammatory responses in skin cells, including dermal fibroblasts, accelerating premature skin aging (photoaging. ETAS 50, a standardized extract from the Asparagus officinalis stem, is a novel and unique functional food that suppresses proinflammatory responses of hydrogen peroxide-stimulated skin fibroblasts and interleukin- (IL- 1β-stimulated hepatocytes. To elucidate its antiphotoaging potencies, we examined whether ETAS 50 treatment after UV-B irradiation attenuates proinflammatory responses of normal human dermal fibroblasts (NHDFs. UV-B-irradiated NHDFs showed reduced levels of the cytosolic inhibitor of nuclear factor-κB α (IκBα protein and increased levels of nuclear p65 protein. The nuclear factor-κB nuclear translocation inhibitor JSH-23 abolished UV-B irradiation-induced IL-1β mRNA expression, indicating that p65 regulates transcriptional induction. ETAS 50 also markedly suppressed UV-B irradiation-induced increases in IL-1β mRNA levels. Immunofluorescence analysis revealed that ETAS 50 retained p65 in the cytosol after UV-B irradiation. Western blotting also showed that ETAS 50 suppressed the UV-B irradiation-induced increases in nuclear p65 protein. Moreover, ETAS 50 clearly suppressed UV-B irradiation-induced distribution of importin-α protein levels in the nucleus without recovering cytosolic IκBα protein levels. These results suggest that ETAS 50 exerts anti-inflammatory effects on UV-B-irradiated NHDFs by suppressing the nuclear import machinery of p65. Therefore, ETAS 50 may prevent photoaging by suppressing UV irradiation-induced proinflammatory responses of dermal fibroblasts.

Application of nuclear irradiation to traditional chinese medicine

International Nuclear Information System (INIS)

Liang Jianping; Li Xuehu; Lu Xihong; Tao Lei; Wang Shuyang

2010-01-01

The application of nuclear irradiation in the field of traditional Chinese medicine has received much attention. In this paper we reviewed the application of nuclear radiation on the cultivation, breeding and disinfection of traditional Chinese medicine, and pointed out that the combination of radiation-induced mutagenesis and biological technology would promise broad prospects for increasing the cellular mutation rate and speeding up the genetic improvement of traditional Chinese medicine. (authors)

Dry storage of irradiated nuclear fuels and vitrified wastes

International Nuclear Information System (INIS)

Deacon, D.

1982-01-01

A review is given of the work of GEC Energy Systems Ltd. over the years in the dry storage of irradiated fuel. The dry-storage module (designated as Cell 4) for irradiated magnox fuel recently constructed at Wylfa nuclear power station is described. Development work on the long-term dry storage of irradiated oxide fuels is reported. Four different methods of storage are compared. These are the pond, vault, cask and caisson stores. It is concluded that there are important advantages with the passive air-cooled ESL dry stove. (U.K.)

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

Nuclear material accounting handbook

International Nuclear Information System (INIS)

2008-01-01

The handbook documents existing best practices and methods used to account for nuclear material and to prepare the required nuclear material accounting reports for submission to the IAEA. It provides a description of the processes and steps necessary for the establishment, implementation and maintenance of nuclear material accounting and control at the material balance area, facility and State levels, and defines the relevant terms. This handbook serves the needs of State personnel at various levels, including State authorities, facility operators and participants in training programmes. It can assist in developing and maintaining accounting systems which will support a State's ability to account for its nuclear material such that the IAEA can verify State declarations, and at the same time support the State's ability to ensure its nuclear security. In addition, the handbook is useful for IAEA staff, who is closely involved with nuclear material accounting. The handbook includes the steps and procedures a State needs to set up and maintain to provide assurance that it can account for its nuclear material and submit the prescribed nuclear material accounting reports defined in Section 1 and described in Sections 3 and 4 in terms of the relevant agreement(s), thereby enabling the IAEA to discharge its verification function as defined in Section 1 and described in Sections 3 and 4. The contents of the handbook are based on the model safeguards agreement and, where applicable, there will also be reference to the model additional protocol. As a State using The handbook consists of five sections. In Section 1, definitions or descriptions of terms used are provided in relation to where the IAEA applies safeguards or, for that matter, accounting for and control of nuclear material in a State. The IAEA's approach in applying safeguards in a State is also defined and briefly described, with special emphasis on verification. In Section 2, the obligations of the State

The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1977-01-01

Concerning refining, fabrication and reprocessing operations of such materials as well as the installation and operation of reactors, necessary regulations are carried out. Namely, in case of establishing the business of refining, fabricating and reprocessing nuclear materials as well as installing nuclear reactors, applications for the permission of the Prime Minister and the Minister of International Trade and Industry should be filed. Change of such operations should be permitted after filing applications. These permissions are retractable. As regards the reactors installed aboard foreign ships, it must be reported to enter Japanese waters and the permission by the Prime Minister must be obtained. In case of nuclear fuel fabricators, a chief technician of nuclear fuel materials (qualified) must be appointed per each fabricator. In case of installing nuclear reactors, the design and methods of construction should be permitted by the Prime Minister. The standard for such permission is specified, and a chief engineer for operating reactors (qualified) must be appointed. Successors inherit the positions of ones who have operated nuclear material refining, fabrication and reprocessing businesses or operated nuclear reactors. (Rikitake, Y.)

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

Joint research centre fusion materials irradiations in HFR: Present status and prospectives

International Nuclear Information System (INIS)

Casini, G.; Fenici, P.

1989-01-01

First a review is made of the Joint Research Centre experimental activity at HFR-Petten in the frame of the Fusion Technology and Safety Programme. The materials under investigation are: Cr-Ni Austenitic steels (316-L type) and Cr-Mn Austenitic steels (AMCR and FI type) as structural materials and Pb-17Li eutetic as tritium breeding material. The experiments on structural materials comprise: Sample irradiations with post-irradiation tensile tests (FRUST) Sample irradiations under constant load and post-irradiation strain measurement (TRIESTE) On-line creep tests (CRISP). The experiments on Pb-17Li breeder material regard sample irradiations to investigate tritium production and recovery as well as tritium permeation through blanket structures (LIBRETTO Experiment). Both irradiations on structural and breeding materials will be pursued up to the end of the current JRC-Multiannual Programme (1988-1991) and even further. In the last part of the paper expected developments of the testing programme at HFR are discussed. New areas of research should involve materials for divertor applications (NET/ITER) and advanced low activation composite materials for Commercial Power Reactors

Research on technology utilizing data freeway for base nuclear power materials

International Nuclear Information System (INIS)

Fujita, Mitsutane; Kurihara, Yutaka; Noda, Tetsuji; Shiraishi, Haruki; Kitajima, Masahiro; Nagakawa, Josei; Yamamoto, Norikazu

1997-01-01

In order to carry out the selection of the nuclear power materials which are used in radiation, from high temperature to very low temperature, and in corrosive environment, and the development of the materials effectively, the construction of huge material data base is indispensable. The development of the distributed type material data base called 'freeway' is advanced jointly by National Research Institute for Metals, Japan Atomic Energy Research Institute, Power Reactor and Nuclear Fuel Development Corporation and Japan Science and Technology Corporation. It has been aimed at that the results obtained in each research institute are made into a data base by that institute, and those data bases can be utilized mutually through network. In fiscal year 1996, the transfer to the system, by which the function showing the contents of system data and the function of data retrieval can be utilized from internet, was begun jointly. The present state of the data freeway, the operation environment of World Wide Web, and the trial making of the computation program for forecasting the change of the chemical composition of materials by neutron irradiation are reported. (K.I.)

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)

Nuclear material management: challenges and prospects

International Nuclear Information System (INIS)

Rieu, J.; Besnainou, J.; Leboucher, I.; Chiguer, M.; Capus, G.; Greneche, D.; Durret, L.F.; Carbonnier, J.L.; Delpech, M.; Loaec, Ch.; Devezeaux de Lavergne, J.G.; Granger, S.; Devid, S.; Bidaud, A.; Jalouneix, J.; Toubon, H.; Pochon, E.; Bariteau, J.P.; Bernard, P.; Krellmann, J.; Sicard, B.

2008-01-01

The articles in this dossier were derived from the papers of the yearly S.F.E.N. convention, which took place in Paris, 12-13 March 2008. They deal with the new challenges and prospects in the field of nuclear material management, throughout the nuclear whole fuel cycle, namely: the institutional frame of nuclear materials management, the recycling, the uranium market, the enrichment market, the different scenarios for the management of civil nuclear materials, the technical possibilities of spent fuels utilization, the option of thorium, the convention on the physical protection of nuclear materials and installations, the characterisation of nuclear materials by nondestructive nuclear measurements, the proliferation from civil installations, the use of plutonium ( from military origin) and the international agreements. (N.C.)

Nuclear material operations manual

International Nuclear Information System (INIS)

Tyler, R.P.; Gassman, L.D.

1978-04-01

This manual is intended to provide a concise and comprehensive documentation of the operating procedures currently practiced at Sandia Laboratories with regard to the management, control, and accountability of radioactive and nuclear materials. The manual is divided into chapters which are devoted to the separate functions performed in nuclear material operations--management, control, accountability, and safeguards, and the final two chapters comprise a document which is also issued separately to provide a summary of the information and operating procedures relevant to custodians and users of radioactive and nuclear materials. The manual also contains samples of the forms utilized in carrying out nuclear material activities. To enhance the clarity of presentation, operating procedures are presented in the form of ''play-scripts'' in which the responsible organizations and necessary actions are clearly delineated in a chronological fashion from the initiation of a transaction to its completion

Nuclear measurements and reference materials

International Nuclear Information System (INIS)

1988-01-01

This report summarizes the progress of the JRC programs on nuclear data, nuclear metrology, nuclear reference materials and non-nuclear reference materials. Budget restrictions and personnel difficulties were encountered during 1987. Fission properties of 235 U as a function of neutron energy and of the resonances can be successfully described on the basis of a three exit channel fission model. Double differential neutron emission cross-sections were accomplished on 7 Li and were started for the tritium production cross-section of 9 Be. Reference materials of uranium minerals and ores were prepared. Special nuclear targets were prepared. A batch of 250 g of Pu0 2 was characterized in view of certification as reference material for the elemental assay of plutonium

Nuclear materials management procedures

International Nuclear Information System (INIS)

Veevers, K.; Silver, J.M.; Quealy, K.J.; Steege, E. van der.

1987-10-01

This manual describes the procedures for the management of nuclear materials and associated materials at the Lucas Heights Research Laboratories. The procedures are designed to comply with Australia's nuclear non-proliferation obligations to the International Atomic Energy Agency (IAEA), bilateral agreements with other countries and ANSTO's responsibilities under the Nuclear Non-Proliferation (Safeguards) Act, 1987. The manual replaces those issued by the Australian Atomic Energy Commission in 1959, 1960 and 1969

Study on the standard establishment for the integrity assessment of nuclear fuel cladding Materials

Energy Technology Data Exchange (ETDEWEB)

Kang, S. S.; Kim, S. H.; Jung, Y. K.; Yang, C. Y.; Kim, I. G.; Choi, Y. H.; Kim, H. J.; Kim, M. W.; Rho, B. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2007-02-15

Fuel cladding material plays important role as a primary structure under the high temperature, high pressure and neutron environment of nuclear power plant. According to this environment, cladding material can be experienced several type aging phenomena including the neutron irradiation embrittlement. On the other hand, although the early nuclear power plant was designed to fitting into the 40MWd/KgU burn-up, the currently power plant intends to go to the high burn-up range. In this case, the safety criteria which was established at low burn-up needs to conform the applicability at the high burn-up. In this study, the safety criteria of fuel cladding material was reviewed to assess the cladding material integrity, and the material characteristics of cladding were reviewed. The current LOCA criterial was also reviewed, and the basic study for re-establishment of LOCA criteria was performed. The time concept safety criteria was also discussed to prevent the breakaway oxidation. Through the this study, safety issues will be produced and be helpful for integrity insurance of nuclear fuel cladding material. This report is 2nd term report.

Hot Cell Post-Irradiation Examination and Poolside Inspection of Nuclear Fuel. Proceedings of the IAEA-HOTLAB Technical Meeting

International Nuclear Information System (INIS)

2013-04-01

The growing operational requirements for nuclear fuel, such as longer fuel cycles, higher burnups and wider use of transient regimes, require more robust fuel designs and more radiation resistant materials. Development of such advanced fuels is only possible with testing and analysis of their performance and application of adequate post-irradiation examination (PIE) methods and techniques. In addition, operational feedback data from poolside and PIE facilities are absolutely necessary for verification of fuel modelling codes and analysis of fuel failure mechanisms. For these reasons, the International Atomic Energy Agency (IAEA) has supported the international exchange of knowledge and sharing of best practices in the application of modern destructive and non-destructive methods of investigation of highly radioactive materials through a series of technical meetings (TMs), the last of which was held in 2006 in Buenos Aires. Since 1963, similar meetings, initially at the European level, have been organized by the Hot Laboratories and Remote Handling Working Group (HOTLAB), a partner in the development of the IAEA's Post Irradiation Examination Facilities Database (PIEDB), part of the IAEA's Integrated Nuclear Fuel Cycle Information System. With this successful partnership in mind, in 2010 the IAEA Technical Working Group on Fuel Performance and Technology recommended that a joint IAEA-HOTLAB TM be held on 'Hot Cell Post-Irradiation Examination and Pool-Side Inspection of Nuclear Fuel', covering questions relevant to the IAEA sub-programmes on 'Nuclear Power Reactor Fuel Engineering' and 'Management of Spent Fuel from Nuclear Power Reactors'. The TM was held on 23-27 May 2011, in Smolenice, Slovakia, with the participation of a large number of interested organizations and comprehensive coverage of major PIE and poolside inspection issues relating to both operation and storage of fuel for nuclear power reactors. The proceedings, summaries and conclusions of that joint

Material input of nuclear fuel

International Nuclear Information System (INIS)

Rissanen, S.; Tarjanne, R.

2001-01-01

The Material Input (MI) of nuclear fuel, expressed in terms of the total amount of natural material needed for manufacturing a product, is examined. The suitability of the MI method for assessing the environmental impacts of fuels is also discussed. Material input is expressed as a Material Input Coefficient (MIC), equalling to the total mass of natural material divided by the mass of the completed product. The material input coefficient is, however, only an intermediate result, which should not be used as such for the comparison of different fuels, because the energy contents of nuclear fuel is about 100 000-fold compared to the energy contents of fossil fuels. As a final result, the material input is expressed in proportion to the amount of generated electricity, which is called MIPS (Material Input Per Service unit). Material input is a simplified and commensurable indicator for the use of natural material, but because it does not take into account the harmfulness of materials or the way how the residual material is processed, it does not alone express the amount of environmental impacts. The examination of the mere amount does not differentiate between for example coal, natural gas or waste rock containing usually just sand. Natural gas is, however, substantially more harmful for the ecosystem than sand. Therefore, other methods should also be used to consider the environmental load of a product. The material input coefficient of nuclear fuel is calculated using data from different types of mines. The calculations are made among other things by using the data of an open pit mine (Key Lake, Canada), an underground mine (McArthur River, Canada) and a by-product mine (Olympic Dam, Australia). Furthermore, the coefficient is calculated for nuclear fuel corresponding to the nuclear fuel supply of Teollisuuden Voima (TVO) company in 2001. Because there is some uncertainty in the initial data, the inaccuracy of the final results can be even 20-50 per cent. The value

Monitoring for fuel sheath defects in three shipments of irradiated CANDU nuclear fuel

International Nuclear Information System (INIS)

Johnson, H.M.

1978-01-01

Analyses of radioactive gases within the Pegase shipping flask were performed at the outset and at the completion of three shipments of irradiated nuclear fuel from the Douglas Point Generating Station to Whiteshell Nuclear Research Establishment. No increases in the concentration of active gases, volatiles or particulates were observed. The activity of the WR-1 bay water rose only marginally due to the storage of the fuel. Other tests indicated that minimal surface contamination was present. These data established that defects in fuel element sheaths did not arise during the transport or the handling of this irradiated fuel. The observation has significance for the prospect of irradiated nuclear fuel transfer and handling in preparation for storage or disposal. (author)

Proceedings of the Tripartite Seminar on Nuclear Material Accounting and Control at Radiochemical Plants

International Nuclear Information System (INIS)

1999-01-01

The problems of creation and operation of nuclear materials (NM) control and accounting systems and their components at radiochemical plants were discussed in seminar during November 2-6 of 1998. There were 63 Russian and 25 foreign participants in seminar. The seminar programme includes following sessions and articles: the aspects of State NM control and accountancy; NM control and accounting in radiochemical plants and at separate stages of reprocessing of spent nuclear fuel and irradiated fuel elements of commercial reactors; NM control and accountancy in storage facilities of radiochemical plants; NM control and accounting computerization, material balance assessment, preparation of reports; qualitative and quantitative measurements in NM control and accounting at radiochemical plants destructive analysis techniques [ru

Study of boron carbide evolution under neutron irradiation; Contribution a l'etude de l'evolution du carbure de bore sous irradiation neutronique

Energy Technology Data Exchange (ETDEWEB)

Simeone, D. [CEA/Saclay, Dept. de Mecanique et de Technologie (DMT), 91 - Gif-sur-Yvette (France)]|[Universite Blaise Pascal, Clermont-Ferrand II, (CNRS), 63 - Aubiere (France)

1999-07-01

Owing to its high neutron efficiency, boron carbide (B{sub 4}C) is used as a neutron absorber in control rods of nuclear plants. Its behaviour under irradiation has been extensively studied for many years. It now seems clear that brittleness of the material induced by the {sup 10}B(n,{alpha}){sup 7}Li capture reaction is due to penny shaped helium bubbles associated to a high strain field around them. However, no model explains the behaviour of the material under neutron irradiation. In order to build such a model, this work uses different techniques: nuclear microprobe X-ray diffraction profile analysis and Raman and Nuclear Magnetic Resonance Spectroscopy to present an evolution model of B{sub 4}C under neutron irradiation. The use of nuclear reactions produced by a nuclear microprobe such as the {sup 7}Li(p,p'{gamma}){sup 7}Li reaction, allows to measure lithium profile in B{sub 4}C pellets irradiated either in Pressurised Water Reactors or in Fast Breeder Reactors. Examining such profiles enables us to describe the migration of lithium atoms out of B{sub 4}C materials under neutron irradiation. The analysis of X-ray diffraction profiles of irradiated B{sub 4}C samples allows us to quantify the concentrations of helium bubbles as well as the strain fields around such bubbles.Furthermore Raman spectroscopy studies of different B{sub 4}C samples lead us to propose that under neutron irradiation. the CBC linear chain disappears. Such a vanishing of this CBC chain. validated by NMR analysis, may explain the penny shaped of helium bubbles inside irradiated B{sub 4}C. (author)

Statistical methods for nuclear material management

International Nuclear Information System (INIS)

Bowen, W.M.; Bennett, C.A.

1988-12-01

This book is intended as a reference manual of statistical methodology for nuclear material management practitioners. It describes statistical methods currently or potentially important in nuclear material management, explains the choice of methods for specific applications, and provides examples of practical applications to nuclear material management problems. Together with the accompanying training manual, which contains fully worked out problems keyed to each chapter, this book can also be used as a textbook for courses in statistical methods for nuclear material management. It should provide increased understanding and guidance to help improve the application of statistical methods to nuclear material management problems

Statistical methods for nuclear material management

Energy Technology Data Exchange (ETDEWEB)

Bowen W.M.; Bennett, C.A. (eds.)

1988-12-01

This book is intended as a reference manual of statistical methodology for nuclear material management practitioners. It describes statistical methods currently or potentially important in nuclear material management, explains the choice of methods for specific applications, and provides examples of practical applications to nuclear material management problems. Together with the accompanying training manual, which contains fully worked out problems keyed to each chapter, this book can also be used as a textbook for courses in statistical methods for nuclear material management. It should provide increased understanding and guidance to help improve the application of statistical methods to nuclear material management problems.

Considerations for sampling nuclear materials for SNM accounting measurements. Special nuclear material accountability report

International Nuclear Information System (INIS)

Brouns, R.J.; Roberts, F.P.; Upson, U.L.

1978-05-01

This report presents principles and guidelines for sampling nuclear materials to measure chemical and isotopic content of the material. Development of sampling plans and procedures that maintain the random and systematic errors of sampling within acceptable limits for SNM(Special Nuclear Materials) accounting purposes are emphasized

Nuclear Materials Management. Proceedings of the Symposium on Nuclear Materials Management

Energy Technology Data Exchange (ETDEWEB)

NONE

1966-02-15

An increasing number of countries are using nuclear materials which, because of their high value and the potential hazards involved, require special methods of handling. To discuss these and to provide a forum at which different systems for achieving the necessary economy and safety could be compared, the International Atomic Energy Agency held a Symposium at Vienna on Nuclear Materials Management from 30 August to 3 September, 1965. It was attended by 115 participants from 19 Member States and two international organizations. Nuclear materials are already being used on an industrial scale and their high cost demands close and continuous control to ensure that they are delivered precisely on time and that they are used to the fullest possible extent before they are withdrawn from service. Routine industrial methods of material control and verification are widely used to ensure safe and economical operation and handling in nuclear power stations, in fuel-element fabrication and reprocessing plants, and in storage facilities. In addition special refinements are needed to take account of the value and the degree of purity required of nuclear materials. Quality as well as quantity has to be checked thoroughly and the utmost economy in processing is necessary. The radioactivity of the material poses special problems of handling and storage and creates a potential hazard to health. A further problem is that of criticality. These dangers and the means of averting them are well understood, as is evidenced by the outstandingly good safety record of the atomic energy industry. But besides accommodating all these special problems, day-to-day procedures must be simple enough to fit in with industrial conditions. Many of the 58 papers presented at the Symposium emphasized that records, checks, measurements and handling precautions, if suitably devised, provide the control vital to efficient operation, serve as checks against loss or waste of valuable materials and help meet the

Capabilities of the Institute of Nuclear Physics (Kazakhstan) for technical expertise of seized nuclear and other radioactive materials

International Nuclear Information System (INIS)

Lukashenko, S.; Chakrov, P.; Gorlachyov, I.; Knyazev, B.; Yakushev, E.

2002-01-01

Full text: Institute of Nuclear Physics of the National Nuclear Center of the Republic of Kazakhstan (INP NNC RK) widely uses the nuclear-physical and others analytical methods which were used during the last years to carry out technical expertise of the nuclear and radioactive materials as well. The spectrometric methods for determination radionuclide composition. INP NNC RK has modern spectrometric equipment for solving all types of analytical and radio analytical problems including: gamma spectrometers - planar, coaxial and well type, alpha spectrometers ('Canberra'), liquid scintillation counter 'TriCarb 3100', beta spectrometers. An original procedures with own software are developed for each spectrometric device. Mass-spectrometric methods. The thermion mass - spectrometry (TI-MS) with prism ionic optics are used for environment objects and nuclear materials analysis. Now the operations on determination of plutonium and uranium isotope composition of the environmental objects of former Semipalatinsk nuclear test site by usage of this method are under way. Scanning electron microscopy (SEM). At the INP, SEM techniques have been used traditionally in studies of irradiated metal materials (original surface, fracture surfaces, cross sections), but rather recently they were successfully applied for characterization of 'hot particles' from nuclear testing site, polymer materials, and also uranium fuel pellets. (The microscope used in AMRAY-1200B equipped with ANS X-ray analyzer). Determination methods of macro - and microelements composition. For determination of macro - and microelement composition the set of various methods are used, including: neutron - activation analysis, atomic - emission spectrometry with high - frequency inductively- coupled plasma, roentgen fluorescent analysis, traditional chemical methods: titrimetry, voltamperometry etc. For determination the most difficult elements - carbon and oxygen the nuclear reactions method is developed at the

Irradiation, annealing, and reirradiation research in the ORNL heavy-section steel irradiation program

International Nuclear Information System (INIS)

Nanstad, R.K.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.

1997-01-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results from work performed as part of the Heavy-Section Steel Irradiation (HSSI) Program managed by Oak Ridge National Laboratory (ORNL) for the U.S. Nuclear Regulatory Commission. The HSSI Program focuses on annealing and re-embrittlement response of materials which are representative of those in commercial RPVs and which are considered to be radiation-sensitive. Experimental studies include (1) the annealing of materials in the existing inventory of previously irradiated materials, (2) reirradiation of previously irradiated/annealed materials in a collaborative program with the University of California, Santa Barbara (UCSB), (3) irradiation/annealing/reirradiation of U.S. and Russian materials in a cooperative program with the Russian Research Center-Kurchatov Institute (RRC-KI), (4) the design and fabrication of an irradiation/anneal/reirradiation capsule and facility for operation at the University of Michigan Ford Reactor, (5) the investigation of potential for irradiation-and/or thermal-induced temper embrittlement in heat-affected zones (HAZs) of RPV steels due to phosphorous segregation at grain boundaries, and (6) investigation of the relationship between Charpy impact toughness and fracture toughness under all conditions of irradiation, annealing, and reirradiation

Determination of internationally controlled materials according to provisions of the law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1981-01-01

This rule is established under the provisions of the law concerning the regulation of nuclear raw materials, nuclear fuel materials and reactors, and the former notification No. 26, 1961, is hereby abolished. Internationally regulated goods under the law are as follows: nuclear raw materials, nuclear fuel materials and moderator materials transferred by sale or other means from the governments of the U.S., U.K., Canada, Australia and France or the persons under their jurisdictions according to the agreements concluded between the governments of Japan and these countries, respectively, the nuclear fuel materials recovered from these materials or produced by their usage, nuclear reactors, the facilities and heavy water transferred by sale or other means from these governments or the persons under their jurisdictions, the nuclear fuel materials produced by the usage of such reactors, facilities and heavy water, the nuclear fuel materials sold by the International Atomic Energy Agency under the contract between the Japanese government and the IAEA, the nuclear fuel materials recovered from these materials or produced by their usage, the heavy water produced by the facilities themselves transferred from the Canadian government, Canadian governmental enterprises or the persons under the jurisdiction of the Canadian government or produced by the usage of these facilities, etc. (Okada, K.)

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)

Effects of non-steady irradiation conditions on fusion materials performance

International Nuclear Information System (INIS)

Matsui, H.; Fukumoto, K.; Nagumo, T.; Nita, N.

2001-01-01

During startup of fusion reactors, materials are exposed to neutron irradiation under non-steady temperature condition. Since the temperature of irradiation has decisive effects on the microstructural evolution, the non-steady temperature will have important consequences in the performance of fusion reactor materials. In the present study, a series of vanadium based alloys have been irradiated with neutrons in a temperature cycling condition. It has been found from this study that cavity number density is much greater in temperature cycled specimens than in steady temperature irradiation. Keeping the upper temperature constant, cavity number density is greater for smaller difference between the upper and the lower temperature. It follows that relatively small temperature excursions may have rather significant effects on the fusion material performance in service. (author)

EDF, a utility and its own needs in the field of transport of nuclear materials

International Nuclear Information System (INIS)

Gouin, P.; Mignot, E.; Hoang, L.P.

1989-01-01

As one of the most important producers of nuclear electricity in the world, EDF is concerned by all the aspects of the transport of nuclear materials and more particularly by those related to the nuclear fuel cycle. EDF is not itself a specialist in this field and most of the transports along the nuclear fuel cycle is done for their own account by their usual partners such as COGEMA or TRANSNUCLEAIRE. Since the beginning of the French nuclear program, they have generally used for these transports casks that already exist on the market and which were well suited to their needs. Nevertheless, new and specific needs appeared during the progress of their nuclear program and have lead them to: study and build new casks or packages, use existing casks for new purposes, develop a device for the measurement of fuel assemblies burn up, develop a software to optimize the evacuation of irradiated fuel for reprocessing. The purpose of this paper is to describe these realization but as a preliminary, they will present briefly the importance of the transport of nuclear materials for EDF

Deformation behavior of irradiated Zr-2.5Nb pressure tube material

International Nuclear Information System (INIS)

Himbeault, D.D.; Chow, C.K.; Puls, M.P.

1994-01-01

A study of the deformation behavior of irradiated highly textured Zr-2.5Nb pressure tube material in the temperature range of 30 degree C to 300 degree C was undertaken to understand better the mechanism for the deterioration of the fracture toughness with neutron irradiation. Strain localization behavior, believed to be a main contributor to reduced toughness, was observed in irradiated transverse tensile specimens at temperature greater than 100 degree C. The strain localization behavior was found to occur by the cooperative twinning of the highly textured grains of the material, resulting in a local softening of the material, where the flow than localizes. It is believed that the effect of the irradiation is to favor twinning at the expense of slip in the early stages of deformation. This effect becomes more pronounced at higher temperature, thus leading to the high-temperature strain localization behavior of the material. A limited amount of dislocation channeling was also observed; however, it is not considered to have a major role in the strain localization behavior of the material. Contrary to previous reports on irradiated zirconium alloys, static strain aging is observed in the irradiated material in the temperature range of 150 degree C to 300 degree C

Design verification test of instrumented capsule (02F-11K) for nuclear fuel irradiation in HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, Bong Goo; Sohn, J. M.; Oh, J. M. [and others

2004-01-01

An instrumented capsule is being developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. The instrumented capsule includes three test fuel rods installed thermocouple to measure fuel centerline temperature and three SPNDs (Self-Powered Neutron Detector) to monitor the neutron flux. Its stability was verified by out-of-pile performance test, and its safety evaluation was also shown that the safety requirements were satisfied. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.8 full power days at 24 MWth). During irradiation, the centerline temperature of PWR UO{sub 2} fuel pellets fabricated by KEPCO Nuclear Fuel Company and the neutron flux were continuously measured and monitored. The test fuel rods were irradiated at less than 350 W/cm to 5.13 GWD/MTU with fuel centerline peak temperature below 1,375 .deg. C. The structural stability of the capsule was satisfied by the naked eye in service pool of HANARO. The capsule and test fuel rods were dismantled and test fuel rods were examined at the hot cell of IMEF (Irradiated Material Examination Facility)

Advanced research workshop: nuclear materials safety

International Nuclear Information System (INIS)

Jardine, L J; Moshkov, M M.

1999-01-01

The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on the storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds of

Perspectives for online analysis of raw material by pulsed neutron irradiation

Science.gov (United States)

Bach, Pierre; Le Tourneur, P.; Poumarede, B.

1997-02-01

On-line analysis by pulsed neutron irradiation is an example of an advanced technology application of nuclear techniques, concerning real problems in the cement, mineral and coal industries. The most significant of these nuclear techniques is their capability of continuous measurement without contact and without sampling, which can lead to improved control of processes and resultant large financial savings. Compared to Californium neutron sources, the use of electrical pulsed neutron generators allows to obtain a higher signal/noise ratio for a more sensitive measurement, and allows to overcome a number of safety problems concerning transportation, installation and maintenance. An experiment related to a possible new on-line raw material analyzer is described, using a pulsed neutron generator. The key factors contributing to an accurate measurement are related to a suitable generator, to a high count rate gamma ray spectroscopy electronics, and to computational tools. Calculation and results for the optimization of the neutron irradiation time diagram are reported. One of the operational characteristics of such an equipment is related to neutron flux available: it is possible to adjust it to the requested accuracy, i.e. for a high accuracy during a few hours/day and for a lower accuracy the rest of the time. This feature allows to operate the neutron tube during a longer time, and then to reduce the cost of analysis.

In situ transmission electron microscope studies of ion irradiation-induced and irradiation-enhanced phase changes

International Nuclear Information System (INIS)

Allen, C.W.

1992-01-01

Motivated at least initially by materials needs for nuclear reactor development, extensive irradiation effects studies employing transmission electron microscopes (TEM) have been performed for several decades, involving irradiation-induced and irradiation-enhanced microstructural changes, including phase transformations such as precipitation, dissolution, crystallization, amorphization, and order-disorder phenomena. From the introduction of commercial high voltage electron microscopes (HVEM) in the mid-1960s, studies of electron irradiation effects have constituted a major aspect of HVEM application in materials science. For irradiation effects studies two additional developments have had particularly significant impact; the development of TEM specimen holder sin which specimen temperature can be controlled in the range 10-2200 K and the interfacing of ion accelerators which allows in situ TEM studies of irradiation effects and the ion beam modification of materials within this broad temperature range. This paper treats several aspects of in situ studies of electron and ion beam-induced and enhanced phase changes and presents two case studies involving in situ experiments performed in an HVEM to illustrate the strategies of such an approach of the materials research of irradiation effects

Synthesis on spinel behaviour under irradiation

International Nuclear Information System (INIS)

Chauvin, N.; Dodane, C.; Noirot, J.; Konings, R.J.M.; Matzke, H.J.; Wiss, T.; Conrad, R.

2001-01-01

The spinel MgAl 2 O 4 is one of the materials able to be used in reactor for the transmutation of the minor actinides stemming from the back-end of the fuel cycle. It has been studied under irradiation since many years. Indeed, one of the first uses considered is to be employed as material for fusion reactors. Otherwise, it was shown that spinel presents nuclear and physico-chemical properties suitable for an utilization as nuclear inert matrix that loaded with an actinide phase constitutes a target devoted to the heterogeneous recycling in reactor. In order to improve the knowledge on spinel behaviour under irradiation, an assessment of the former studies must be done. The objective of this paper is to gather all the results of the spinel irradiations and to take out synthetic conclusion on the opportunity to use this material for the transmutation programme. (author)

Material property changes of stainless steels under PWR irradiation

International Nuclear Information System (INIS)

Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

2009-01-01

Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

An investigation of neutron irradiation test on superplastic zirconia-ceramic materials

International Nuclear Information System (INIS)

Shibata, Taiju; Ishihara, Masahiro; Baba, Shinichi; Hayashi, Kimio

2000-05-01

A neutron irradiation test on superplastic ceramic materials at high temperature has been proposed as an innovative basic research on high-temperature engineering using the High Temperature Engineering Test Reactor (HTTR). For the effective execution of the test, we reviewed the superplastic deformation mechanism of ceramic materials and discussed neutron irradiation effects on the superplastic deformation process of stabilized Tetragonal Zirconia Polycrystal (TZP), which is a representative superplastic ceramic material. As a result, we pointed out that the decrease in the activation energy for superplastic deformation is expected by the radiation-enhanced diffusion. We selected a fast neutron fluence of 5x10 20 n/cm 2 and an irradiation temperature of about 600degC as test conditions for the first irradiation test on TZP and decided to perform a preliminary irradiation test by the Japan Materials Testing Reactor (JMTR). Moreover, we estimated the radioactivity of irradiated TZP and indicated that it is in the order of 10 10 Bq/g (about 0.3 Ci/g) immediately after irradiation to a thermal neutron fluence of 3x10 20 n/cm 2 and that it decays to about 1/100 in a year. (author)

Evaluating the attractiveness of nuclear material for proliferation-resistance and nuclear security

International Nuclear Information System (INIS)

Choi, Jor-Shan; Ikegame, Kou; Kuno, Yusuke

2011-01-01

The attractiveness of nuclear material, defined as a function of the isotopic composition of the nuclear material in formulas expressing the material's intrinsic properties, is of considerably debate in recent developments of proliferation-resistance measures of a nuclear energy system. A reason for such debate arises from the fact that the concept of nuclear material attractiveness can be confusing because the desirability of a material for nuclear explosive use depends on many tangible and intangible factors including the intent and capability of the adversary. In addition, a material that is unattractive to an advanced nation (in the case of proliferation) may be very attractive to a terrorist (in the case of physical protection and nuclear security). Hence, the concept of 'Nuclear Material Attractiveness' for different nuclear materials must be considered in the context of safeguards and security. The development of a ranking scheme on the attractiveness of nuclear materials could be a useful concept to start-off the strategies for safeguards and security on a new footing (i.e., why and how nuclear material is attractive, and what are the quantifiable basis). Japan may benefit from such concept regarding the attractiveness of nuclear materials when recovering nuclear materials from the damaged cores in Fukushima because safety, security, and safeguards (3S) would be a prominent consideration for the recovery operation, and it would be the first time such operation is performed in a non-nuclear weapons state. (author)

Uncertainty estimation in nuclear material weighing

Energy Technology Data Exchange (ETDEWEB)

Thaure, Bernard [Institut de Radioprotection et de Surete Nucleaire, Fontenay aux Roses, (France)

2011-12-15

The assessment of nuclear material quantities located in nuclear plants requires knowledge of additions and subtractions of amounts of different types of materials. Most generally, the quantity of nuclear material held is deduced from 3 parameters: a mass (or a volume of product); a concentration of nuclear material in the product considered; and an isotopic composition. Global uncertainties associated with nuclear material quantities depend upon the confidence level of results obtained in the measurement of every different parameter. Uncertainties are generally estimated by considering five influencing parameters (ISHIKAWA's rule): the material itself; the measurement system; the applied method; the environmental conditions; and the operator. A good practice guide, to be used to deal with weighing errors and problems encountered, is presented in the paper.

Development of a Small Punch Test Technique for an Evaluation of the Mechanical Properties of Irradiated Materials in a Hot Cell

International Nuclear Information System (INIS)

Kim, Do-Sik; Ahn, Sang-Bok; Yoo, Byung-Ok; Choo, Yong-Sun; Hong, Kwon-Pyo

2006-01-01

Miniaturized specimens have been widely used to evaluate the mechanical properties of steels and plastics. Especially for a study on the irradiation effects in nuclear materials, the small specimen test techniques have attracted considerable attention. Therefore, it is essential that the test techniques be developed and verified to extract the mechanical properties information from small specimens. Among the test techniques using small specimens, the small punch (SP) test technique using small disc-sized specimen has been successfully used to estimate the tensile properties (yield strength and ultimate tensile strength), DBTT (ductile-brittle transition temperature), fracture toughness and creep properties of metals irradiated in a reactor or a proton accelerator. In this paper, the existing SP test techniques are reviewed and summarized. In addition, the information on the development of the SP test procedure is obtained to evaluate the radiation effects on the mechanical properties of nuclear materials in a hot cell

Evaluation by nuclear relaxation of hydrogen tucuman irradiated fruit

International Nuclear Information System (INIS)

Lima, Keila dos Santos Cople; Lima, Antonio Luis dos Santos; Araujo, Leandro Moreira; Tavares, Maria Ines Bruno

2011-01-01

The tucuman (Astrocarium vulgare Mart.) develops a yellow-orange fibrous pulp and a yellow-green peel as it gets mature. In Brazil it is considered a good source of carotenoids, which is the main precursor of vitamin A. Considering that the irradiation process is an alternative to avoid post-harvesting losses, without changing the nutritional value of food, this study had the objective of evaluating of different gamma irradiation doses (0.0; 0.5; 1.0; 2.0 kGy) in Tucuman samples, divided into control (non irradiated) and irradiated, by the low-field Magnetic Nuclear Resonance (NMR) analysis. The technique is used in quantitative determinations of nondestructively and non-invasive instrumentation employing low, with minimal sample, quickly, preserving its constitution and nature. (author)

The nuclear risk

International Nuclear Information System (INIS)

Choudens, Henri de

2013-06-01

After a presentation of some general notions related to radioactivity, to the action and effects of radiation on matter and on living creatures, to natural irradiation and irradiation resulting from human activities, and to standards and regulation in radiation protection, this publication presents issues related to the safety of nuclear installations: presentation of the different types of nuclear installations, safety principles. The next part proposes an overview of different types of criticality accidents by distinguishing those occurring in medical or industrial installations and radioactive sources, those related to military activities, and those occurring in nuclear reactors. Other addressed topics are: radioactive wastes and effluents (definitions, classification, origin and quantity, packaging and storage), transportation of radioactive materials (regulation, parcel types, radiation protection during transportation, transport accidents), regulation of basic nuclear installations and safety organisation (safety control during normal operation, crisis management, post-accidental management, public information), the protection of sensitive nuclear materials (material types and general principles, regulation, arrangement to protect materials in installations, transports, audits, international controls). Nuclear fusion is then presented (principle, confinement types, the ITER project), and an overview of military applications is finally proposed (ship propulsion, nuclear weapons, the non proliferation issue)

Determination of B and Li in nuclear materials by secondary-ion mass spectrometry

International Nuclear Information System (INIS)

Eby, R.E.; Christie, W.H.

1981-01-01

Secondary ion mass spectrometry (SIMS) was used to perform mass and isotopic analysis for B and Li in samples that are not readily amenable to more conventional mass spectrometric techniques (e.g., surface ionization, electron impact, etc.). In this paper three specific applications of SIMS analysis to nuclear materials are discussed: first, the quantitative determination of B and its isotopic composition in borosilicate glasses; second, the determination of the isotopic composition of B and Li in irradiated nuclear-grade aluminum oxide/boron carbide composite pellets, and, lastly, the quantitative and isotopic determination of B and Li in highly radioactive solutions of unknown composition

Nano-pulsed laser irradiation scanning system for phase-change materials

International Nuclear Information System (INIS)

Kim, Sookyung; Li Xuezhe; Lee, Sangbin; Kim, Kyung-Ho; Lee, Seung-Yop

2008-01-01

Recently, the demand of a laser irradiation tester is increasing for phase change random access memory (PRAM) as well as conventional optical storage media. In this study, a nano-pulsed laser irradiation system is developed to characterize the optical property and writing performance of phase-change materials, based on a commercially available digital versatile disk (DVD) optical pick-up. The precisely controlled focusing and scanning on the material's surface are implemented using the auto-focusing mechanism and a voice coil motor (VCM) of the commercial DVD pick-up. The laser irradiation system provides various writing and reading functions such as adjustable laser power, pulse duration, recording pattern (spot, line and area), and writing/reading repetition, phase transition, and in situ reflectivity measurement before/after irradiation. Measurements of power time effect (PTE) diagram and reflectivity map of Ge 2 Sb 2 Te 5 samples show that the proposed laser irradiation system provides the powerful scanning tool to quantify the optical characteristics of phase-change materials

The Change of the Seebeck Coefficient Due to Neutron Irradiation and Thermal Fatigue of Nuclear Reactor Pressure Vessel Steel and its Application to the Monitoring of Material Degradation

International Nuclear Information System (INIS)

Niffenegger, M.; Reichlin, K.; Kalkhof, D.

2002-05-01

The monitoring of material degradation, that might be caused by neutron irradiation and thermal fatigue, is an important topic in lifetime extension of nuclear power plants. We therefore investigated the application of the Seebeck effect for determining material degradation of common reactor pressure vessel steel. The Seebeck coefficient (SC) of several irradiated Charpy specimens made from Japanese JRQ-steel were measured. The specimens suffered a fluence from 0 up to 4.5 x 10 19 neutrons per cm 2 with energies higher than 1 MeV. The measured changes of the SC within this range were about 500 nV, increasing continuously in the range under investigation. Some indications of saturation appeared at fluencies larger than 4.55 x 10 19 neutrons per cm 2 . We obtained a linear dependency between the SC and the temperature shift ΔT 41 of the Charpy-Energy- Temperature curve which is widely used to characterize material embrittlement. Similar measurements were performed on specimens made from the widely used austenitic steel X6CrNiTi18-10 (according to DIN 1.4541) that were fatigued by applying a cyclic strain amplitude of 0.28%. For this kind of fatigue the observed change of SC was somewhat smaller than for the irradiated specimens. Further investigations were made to quantify the size of the gage volume in which the thermoelectric power is generated. It appeared that the information gathered from a Thermo Electric Power (TEP) measurement is very local. To overcome this problem we propose a novel TEP-method using a Thermoelectric Scanning Microscope (TSM). We finally conclude that the change of the SC has a potential for monitoring of material degradation due to neutron irradiation and thermal fatigue, but it has to be taken into account that several influencing parameters could contribute to the TEP in either an additional or extinguishing manner. A disadvantage of the method is the requirement of a clean surface without any oxide layer. A part of this disadvantage can

Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 3, Site team reports

Energy Technology Data Exchange (ETDEWEB)

1993-11-01

A self assessment was conducted of those Hanford facilities that are utilized to store Reactor Irradiated Nuclear Material, (RINM). The objective of the assessment is to identify the Hanford inventories of RINM and the ES & H concerns associated with such storage. The assessment was performed as proscribed by the Project Plan issued by the DOE Spent Fuel Working Group. The Project Plan is the plan of execution intended to complete the Secretary`s request for information relevant to the inventories and vulnerabilities of DOE storage of spent nuclear fuel. The Hanford RINM inventory, the facilities involved and the nature of the fuel stored are summarized. This table succinctly reveals the variety of the Hanford facilities involved, the variety of the types of RINM involved, and the wide range of the quantities of material involved in Hanford`s RINM storage circumstances. ES & H concerns are defined as those circumstances that have the potential, now or in the future, to lead to a criticality event, to a worker radiation exposure event, to an environmental release event, or to public announcements of such circumstances and the sensationalized reporting of the inherent risks.

Polymeric materials obtained by electron beam irradiation

International Nuclear Information System (INIS)

Dragusin, M.; Moraru, R.; Martin, D.; Radoiu, M.; Marghitu, S.; Oproiu, C.

1995-01-01

Research activities in the field of electron beam irradiation of monomer aqueous solution to produce polymeric materials used for waste waters treatment, agriculture and medicine are presented. The technologies and special features of these polymeric materials are also described. The influence of the chemical composition of the solution to ba irradiated, absorbed dose level and absorbed dose rate level are discussed. Two kinds of polyelectrolytes, PA and PV types and three kinds of hydrogels, pAAm, pAAmNa and pNaAc types, the production of which was first developed with IETI-10000 Co-60 source and then adapted to the linacs built in Accelerator Laboratory, are described. (author)

Studies on the irradiated solids

International Nuclear Information System (INIS)

Lesueur, D.

1988-01-01

The 1988 progress report of the Irradiated Solids laboratory (Polytechnic School, France), is presented. The Laboratory activities concern the investigations on disordered solids (chemical or structural disorder). The disorder itself, its effects on the material physical properties and the particle-matter interactions, are investigated. The research works are performed in the following fields: solid state physics, irradiation and stoechiometric defects, and nuclear materials. The scientific reviews, the congress communications and the thesis are listed [fr

The Physical Protection of Nuclear Material and Nuclear Facilities

International Nuclear Information System (INIS)

1999-08-01

Physical protection against the theft or unauthorized diversion of nuclear materials and against the sabotage of nuclear facilities by individuals or groups has long been a matter of national and international concern. Although responsibility for establishing and operating a comprehensive physical protection system for nuclear materials and facilities within a State rests entirely with the Government of that State, it is not a matter of indifference to other States whether and to what extent that responsibility is fulfilled. Physical protection has therefore become a matter of international concern and co-operation. The need for international co-operation becomes evident in situations where the effectiveness of physical protection in one State depends on the taking by other States also of adequate measures to deter or defeat hostile actions against nuclear facilities and nuclear materials, particularly when such materials are transported across national frontiers

The Physical Protection of Nuclear Material and Nuclear Facilities

International Nuclear Information System (INIS)

1999-06-01