Capsule Development and Utilization for Material Irradiation Tests

Energy Technology Data Exchange (ETDEWEB)

Kim, Bong Goo; Kang, Y. H.; Cho, M. S. (and others)

2007-06-15

The essential technology for an irradiation test of materials and nuclear fuel has been successively developed and utilized to meet the user's requirements in Phase I(July 21, 1997 to March 31, 2000). It enables irradiation tests to be performed for a non-fissile material under a temperature control(300{+-}10 .deg. C) in a He gas environment, and most of the irradiation tests for the internal and external users are able to be conducted effectively. The basic technology was established to irradiate a nuclear fuel, and a creep capsule was also developed to measure the creep property of a material during an irradiation test in HANARO in Phase II(April 1, 2000 to March 31, 2003). The development of a specific purpose capsule, essential technology for a re-irradiation of a nuclear fuel, advanced technology for an irradiation of materials and a nuclear fuel were performed in Phase III(April 1, 2003 to February 28, 2007). Therefore, the technology for an irradiation test was established to support the irradiation of materials and a nuclear fuel which is required for the National Nuclear R and D Programs. In addition, an improvement of the existing capsule design and fabrication technology, and the development of an instrumented capsule for a nuclear fuel and a specific purpose will be able to satisfy the user's requirements. In order to support the irradiation test of materials and a nuclear fuel for developing the next generation nuclear system, it is also necessary to continuously improve the design and fabrication technology of the existing capsule and the irradiation technology.

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.)

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)

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.)

Irradiation can for the activation of materials in nuclear reactors

International Nuclear Information System (INIS)

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

1985-01-01

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

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.

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...

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)

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

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)

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.

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

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

Effect of 60Co γ-irradiation on saccharification of uncooked sweet potato material

International Nuclear Information System (INIS)

Hu Tingchun; Xiong Xingyao; Yi Jinqiong; Wang Keqin; Su Xiaojun; Zou Jianfeng

2010-01-01

Using the starch and powder of sweet potato of Xiangshu 86 and Xiangshu 541 as materials, the effect of 60 Co γ-irradiation on the structure of starch particle and the efficiency of saccharification were studied. The result showed that some reticulate flaws appeared in the surface of irradiated starch particles, and the reticulate flaws were increased with the increase of irradiation dose. The content of reducing sugar and total soluble sugar in both starch and the powder were obviously increased along with the increase of irradiation dose ranged from 50 to 1200 kGy. The saccharification efficiency of Xiangshu 86 and Xiangshu 541 was obviously difference at the dose lower than 500 kGy, and then the efficiency showed the similar trends at higher dose irradiation, the saccharification rate reached the highest value after the treatment of 1200 kGy irradiation. (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

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.

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

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)

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.

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

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.

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.

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

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

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

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

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 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)

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

Similarity between the effects of carbon-ion irradiation and X-irradiation on the development of rat brain

International Nuclear Information System (INIS)

Inouye, Minoru; Hayasaka, Shizu; Murata, Yoshiharu; Takahashi, Sentaro; Kubota, Yoshihisa

2000-01-01

The effects of carbon-ion irradiation and X-irradiation on the development of rat brain were compared. Twenty pregnant rats were injected with bromodeoxyuridine (BrdU) at 9 pm on day 18 pregnancy and divided into five groups. Three hours after injection (day 19.0) one group was exposed to 290 MeV/u carbon-ion radiation by a single dose of 1.5 Gy. Other groups were exposed to X-radiation by 1.5, 2.0 or 2.5 Gy, or sham-treated, respectively. Fetuses were removed from one dam in each group 8 h after exposure and examined histologically. Extensive cell death was observed in the brain mantle from the irradiated groups. The cell death after 1.5 Gy carbon-ion irradiation was remarkably more extensive than that after 1.5 Gy X-irradiation, but comparable to that after 2.0 Gy or 2.5 Gy X-irradiation. The remaining rats were allowed to give birth and the offspring were sacrificed at 6 weeks of age. All of the irradiated offspring manifested microcephaly. The size of the brain mantle exposed to 1.5 Gy carbon-ion radiation was significantly smaller than that exposed to 1.5 Gy X-radiation and larger than that exposed to 2.5 Gy X-radiation. A histological examination of the cerebral cortex revealed that cortical layers II-IV were malformed. The defect by 1.5 Gy carbon-ion irradiation was more severe than that by the same dose of X-irradiation. Although the BrdU-incorporated neurons were greatly reduced in number in all irradiated groups, these cells reached the superficial area of the cortex. These findings indicated that the effects of both carbon-ion irradiation and X-irradiation on the development of rat brain are similar in character, and the effect of 1.5 Gy carbon-ion irradiation compares to that of 2.0-2.5 Gy X-irradiation. (author)

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

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

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.)

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

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

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

Development status of irradiation devices and instrumentation for material and nuclear fuel irradiation tests in HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, Bong Goo; Sohn, Jae Min; Choo, Kee Nam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-04-15

The High flux Advanced Neutron Application ReactOr (HANARO), an open-tank-in-pool type reactor, is one of the multi-purpose research reactors in the world. Since the commencement of HANARO's operations in 1995, a significant number of experimental facilities have been developed and installed at HANARO, and continued efforts to develop more facilities are in progress. Owing to the stable operation of the reactor and its frequent utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The irradiation testing equipment for nuclear fuels and materials at HANARO can be classified into capsules and the Fuel Test Loop (FTL). Capsules for irradiation tests of nuclear fuels in HANARO have been developed for use under the dry conditions of the coolant and materials at HANARO and are now successfully utilized to perform irradiation tests. The FTL can be used to conduct irradiation testing of a nuclear fuel under the operating conditions of commercial nuclear power plants. During irradiation tests conducted using these capsules in HANARO, instruments such as the thermocouple, Linear Variable Differential Transformer (LVDT), small heater, Fluence Monitor (F/M) and Self-Powered Neutron Detector (SPND) are used to measure various characteristics of the nuclear fuel and irradiated material. This paper describes not only the status of HANARO and the status and perspective of irradiation devices and instrumentation for carrying out nuclear fuel and material tests in HANARO but also some results from instrumentation during irradiation tests

Study of PDMS conformation in PDMS-based hybrid materials prepared by gamma irradiation

International Nuclear Information System (INIS)

Lancastre, J.J.H.; Fernandes, N.; Margaça, F.M.A.; Miranda Salvado, I.M.; Ferreira, L.M.; Falcão, A.N.; Casimiro, M.H.

2012-01-01

Polydimethylsiloxane-silicate based hybrid materials have recognized properties (high flexibility, low elastic modulus or high mechanical strength) for which there are a large number of applications in development, such as for the bioapplications field. The hybrids addressed in the present study were prepared by gamma irradiation of a mixture of polydimethylsiloxane (PDMS) with tetraethylorthosilicate (TEOS) and zirconium propoxide (PrZr) without addition of any solvent or other product. The materials are homogeneous, transparent, monolithic and flexible. The structure dependence on the PrZr content is addressed. A combination of X-ray diffraction (XRD) and Infrared Spectroscopy (IR) was used. The results reveal that the polymer in the hybrids prepared with PrZr, in a content≤5 wt%, shows a structure similar to that in the irradiated pure polymer sample. In these samples the presence of ordered polymer regions is clearly found. For samples prepared with higher content of Zr almost no ordered polymer regions are observed. The addition of PrZr plays an important role on polymer conformation in these hybrid materials. - Highlights: ► PDMS-based hybrid materials were prepared by γ-irradiation. ► FTIR, ATR/FT-IR and XRD techniques were used to characterize the materials. ► Changes in FTIR bands reflect growth of crosslinking network. ► Above certain Zr concentration regions of Zr-silicate oxide are formed. ► Zr content determines conformation of the polymer chain network.

Testing of irradiated and annealed 15H2MFA materials

International Nuclear Information System (INIS)

Gillemot, F.; Uri, G.

1994-01-01

A set of surveillance samples made from 15H2MFA material has been studied in the laboratory of AEKI. Miniature notched tensile specimens were cut from some remnants of irradiated and broke surveillance charpy remnants. The Absorbed Specific Fracture Energy (ASFE) was measured on the specimens. A cutting machine and testing technique were elaborated for the measurements. The second part of the Charpy remnants was annealed at 460 deg. C and 490 deg. C for 6-8 hours. The specimens were tested similarity and the results were compared. (author). 5 refs, 9 figs

On the Thermal Conductivity Change of Matrix Graphite Materials after Neutron Irradiation

Energy Technology Data Exchange (ETDEWEB)

Lee, Young-Woo; Yeo, Seunghwan; Kim, Eung-Seon; Sah, Injin; Park, Daegyu; Kim, Youngjun; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this work, the variations of the thermal conductivity of the A3-3 matrix graphite after neutron irradiation is discussed as well as of the IG-110 graphite for comparison. Neutron irradiation of the graphite specimens was carried out as a part of the first irradiation test of KAERI's coated particle fuel specimens by use of Hanaro research reactor. This work can be summarized as follows: 1) In the evaluation of the specific heat of the graphite materials, various literature data were used and the variations of the specific heat data of all the graphite specimens are observed well agreed, irrespectively of the difference in specimens (graphite and matrix graphite and irradiated and un-irradiated). 2) This implies that it should be reasonable that for both structural graphite and fuel matrix graphite, and even for the neuron-irradiated graphite, any of these specific heat data set be used in the calculation of the thermal conductivity. 3) For the irradiated A3-3 matrix graphite specimens, the thermal conductivity decreased on both directions. On the radial direction, the tendency of variation upon temperature is similar to that of unirradiated specimen, i.e., decreasing as the temperature increases. 4) In the German irradiation experiments with A3-27 matrix graphite specimens, the thermal conductivity of the un-irradiated specimen shows a decrease and that of irradiated specimen is nearly constant as the temperature increases. 5) The thermal conductivity of the irradiated IG-110 was considerably decreased compared with that of un-irradiated specimens The difference of the thermal conductivity of un-irradiated and irradiated IG-110 graphite specimens is much larger than that of un-irradiated and irradiated A3-3 matrix graphite specimens.

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)

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

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)

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)

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)

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

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)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Fusion Materials Irradiation Test Facility: experimental capabilities and test matrix

International Nuclear Information System (INIS)

Opperman, E.K.

1982-01-01

This report describes the experimental capabilities of the Fusion Materials Irradiation Test Facility (FMIT) and reference material specimen test matrices. The description of the experimental capabilities and the test matrices has been updated to match the current single test cell facility ad assessed experimenter needs. Sufficient detail has been provided so that the user can plan irradiation experiments and conceptual hardware. The types of experiments, irradiation environment and support services that will be available in FMIT are discussed

Irradiation data analysis and thermal analysis of the 02M-02K capsule for material irradiation test

International Nuclear Information System (INIS)

Choi, Myoung Hwan; Choo, K. N.; Kang, Y. H.; Kim, B. G.; Cho, M. S.; Sohn, J. M.; Shin, Y. T.; Park, S. J.; Kim, Y. J.

2004-11-01

In order to evaluate the fracture toughness of RPV materials, the material irradiation test using the instrumented capsule (02M-02K) were carried out in the HANARO in August 2003. Based on the user's requirements the thermal design analysis of the capsule 02M-02K was performed, and the specimens were suitably arranged in each step of the capsule main body. In this report, both the temperature data of specimens measured during irradiation test and the calculated data from the thermal analysis are compared and evaluated. Also, the temperature profile in each step with the HANARO reactor power and helium pressure is reviewed and evaluated. The effects of the gap size such as theoretically calculated from thermal expansion during irradiation test and measured one in the manufacturing of the capsule on the specimen temperature were reviewed. The thermal analysis was performed by using a Finite Element (FE) analysis program, ANSYS. Two-dimensional model for the 1/4 section of the capsule is generated, and the γ-heating rate of the materials used in the capsule at the control rod position of 430 mm is used as input data. The thermal analysis using a 3-dimensional model, which is quite similar to the actual shape of the capsule, is also conducted to obtain the temperature distribution in the axial direction. The analysis results show that the temperature difference between the top and bottom positions of a specimen is found to be smaller than 13.2 .deg. C. The maximum measured and calculated temperature in the step 3 of the capsule is 256 .deg. C and 264 .deg. C, respectively. The measured temperature data are obtained at the reactor power of 24 MW, the heater power of 0 W and the helium pressure of 760 torr. Generally, the temperature data obtained by the FE analysis are slightly lower than those of the measured except the step 1 of the capsule. However, the temperature difference between the measured and the calculated shows a good agreement within 9 percent. It is

Stored energy in fusion magnet materials irradiated at low temperatures

International Nuclear Information System (INIS)

Chaplin, R.L.; Kerchner, H.R.; Klabunde, C.E.; Coltman, R.R.

1989-08-01

During the power cycle of a fusion reactor, the radiation reaching the superconducting magnet system will produce an accumulation of immobile defects in the magnet materials. During a subsequent warm-up cycle of the magnet system, the defects will become mobile and interact to produce new defect configurations as well as some mutual defect annihilations which generate heat-the release of stored energy. This report presents a brief qualitative discussion of the mechanisms for the production and release of stored energy in irradiated materials, a theoretical analysis of the thermal response of irradiated materials, theoretical analysis of the thermal response of irradiated materials during warm-up, and a discussion of the possible impact of stored energy release on fusion magnet operation 20 refs

Experimental data base for assessment of irradiation induced ageing effects in pre-irradiated RPV materials of German PWR

Energy Technology Data Exchange (ETDEWEB)

Hein, H.; Gundermann, A.; Keim, E.; Schnabel, H. [AREVA NP GmbH (Germany); Ganswind, J. [VGB PowerTech e.V (Germany)

2011-07-01

The 5 year research program CARISMA which ended in 2008 has produced a data base to characterize the fracture toughness of pre-irradiated original RPV (Reactor Pressure Vessel) materials being representative for all four German PWR construction lines of former Siemens/KWU company. For this purpose tensile, Charpy-V impact, crack initiation and crack arrest tests have been performed for three base materials and four weld metals irradiated to neutron fluences beyond the designed EoL range. RPV steels with optimized chemical composition and with high copper as well as high nickel content were examined in this study. The RTNDT concept and the Master Curve approach were applied for the assessment of the generated data in order to compare both approaches. A further objective was to clarify in which extent crack arrest curves can be generated for irradiated materials and how crack arrest can be integrated into the Master Curve approach. By the ongoing follow-up project CARINA the experimental data base will be extended by additional representative materials irradiated under different conditions and with respect to the accumulated neutron fluences and specific impact parameters such as neutron flux and manufacturing effects. The irradiation data cover also the long term irradiation behavior of the RPV steels concerned. Moreover, most of the irradiated materials were and will be used for microstructural examinations to get a deeper insight in the irradiation embrittlement mechanisms and their causal relationship to the material property changes. By evaluation of the data base the applicability of the Master Curve approach for both crack initiation and arrest was confirmed to a large extent. Moreover, within both research programs progress was made in the development of crack arrest test techniques and in specific issues of RPV integrity assessment. (authors)

An investigation of high-temperature irradiation test program of new ceramic materials

International Nuclear Information System (INIS)

Ishino, Shiori; Terai, Takayuki; Oku, Tatsuo

1999-08-01

The Japan Atomic Energy Research Institute entrusted the Atomic Energy Society of Japan with an investigation into the trend of irradiation processing/damage research on new ceramic materials. The present report describes the result of the investigation, which was aimed at effective execution of irradiation programs using the High Temperature Engineering Test Reactor (HTTR) by examining preferential research subjects and their concrete research methods. Objects of the investigation were currently on-going preliminary tests of functional materials (high-temperature oxide superconductor and high-temperature semiconductor) and structural materials (carbon/carbon and SiC/SiC composite materials), together with newly proposed subjects of, e.g., radiation effects on ceramics-coated materials and super-plastic ceramic materials as well as microscopic computer simulation of deformation and fracture of ceramics. These works have revealed 1) the background of each research subject, 2) its objective and significance from viewpoints of science and engineering, 3) research methodology in stages from preliminary tests to real HTTR irradiation, and 4) concrete HTTR-irradiation methods which include main specifications of test specimens, irradiation facilities and post-irradiation examination facilities and apparatuses. The present efforts have constructed the important fundamentals in the new ceramic materials field for further planning and execution of the innovative basic research on high-temperature engineering. (author)

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

International Nuclear Information System (INIS)

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

1997-10-01

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

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

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

Thermal analysis of the APT materials irradiation samples

International Nuclear Information System (INIS)

Maloy, S.A.; Willcutt, G.J.; James, M.R.; Teague, J.; Diebe, D.A.; Sommer, W.F.; Ferguson, P.D.

1998-01-01

The accelerator production of tritium (APT) project proposes to use a 1.7 GeV, 100 mA proton beam to produce neutrons from an Inconel 718 clad tungsten target. The neutrons are multiplied and moderated in a lead/water blanket before being captured in He 3 to form tritium. In this process, the materials in the target and blanket region are exposed to a wide range of different fluxes comprised of protons and neutrons with energies into the GeV range. To investigate the effect of irradiation on the mechanical properties of candidate APT materials (Inconel 718, 316L stainless steel, Al 6061-T6, Mod 9Cr-1Mo, 304L stainless steel and Al5052-0), the APT Engineering Design and Development group fielded an extensive materials irradiation using the LANSCE (Los Alamos Neutron Science Center) accelerator, which operates at an energy of 800 MeV and a current of 1 mA. The test set-up was designed to place mechanical test specimens in locations in and near the proton beam where the environment of proton and neutron fluxes and temperatures are prototypic to those expected in the APT target/blanket (50--170 C). After irradiating for about 3,600 hours, the maximum achieved proton fluence was 4--5 x 10 21 p/cm 2 for the materials in the center of the beam. To obtain relevant data on the change in the mechanical properties with fluence, it is essential to know the temperature at which the materials were irradiated. This paper explains the method of determining the specimen temperature and reports some specific examples

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.

Corrosion of electron-irradiated Zr-2.5Nb and Zircaloy-2

International Nuclear Information System (INIS)

Woo, O.-T.; McDougall, G.M.; Hutcheon, R.M.; Urbanic, V.F.; Griffiths, M.; Coleman, C.E.

2000-01-01

We used 10-MeV electrons to rapidly produce radiation damage in zirconium alloys, investigated whether electrons produced the same microstructural changes as neutrons, then performed post-irradiation corrosion tests to determine whether electron-irradiated materials displayed similar corrosion behavior to neutron-irradiated materials. Two irradiations were completed using 10-MeV electrons with the beam normal to thin disks of material of 4 diameter slightly larger than the beam. The beam distribution. and disk cooling were designed to produce radial temperature and dose distributions having maxima at the disk center. A high-temperature irradiation was performed on annealed Zr-2.5Nb disks, achieving a central dose of 1.3 dpa and at a central temperature of ∼450 deg C. After irradiation, the samples contained needle-like β-Nb precipitates in the α-Zr matrix similar to those produced by neutrons. A low-temperature irradiation was performed on half-moon disks of Zr-2.5Nb and Zircaloy-2 pressure tube materials at 310 deg C central temperature and 1.3-dpa central dose. Dislocation loops were observed, again similar to those produced in neutron-irradiated materials. Some of the high-temperature electron-irradiated disks were exposed to 300 deg C moist air (saturated with D 2 O), and in separate tests, high- and low-temperature irradiated disks were corroded in 300 deg C D 2 0 (11.0 pD at room temperature) in an autoclave. Measurements of oxide thickness by Fourier Transform Infrared Reflectance (FTIR) spectroscopy showed that electron irradiation reduced the corrosion rate of Zr-2.5Nb compared with that of unirradiated material, as observed for neutron irradiation. For exposures to moist air and to D 2 O, the theoretical deuterium uptakes for the electron-irradiated materials were, respectively, about 4 times and 1.5 to 2 times those for the unirradiated materials. This is also in good agreement with results for neutron-irradiated pressure tube materials. Thus, 10-Me

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

Production behavior of irradiation defects in solid breeder materials

Energy Technology Data Exchange (ETDEWEB)

Moriyama, Hirotake; Moritani, Kimikazu [Kyoto Univ. (Japan)

1998-03-01

The irradiation effects in solid breeder materials are important for the performance assessment of fusion reactor blanket systems. For a clearer understanding of such effects, we have studied the production behavior of irradiation defects in some lithium ceramics by an in-situ luminescence measurement technique under ion beam irradiation. The luminescence spectra were measured at different temperatures, and the temperature-transient behaviors of luminescence intensity were also measured. The production mechanisms of irradiation defects were discussed on the basis of the observations. (author)

Comparison of swelling for structural materials on neutron and ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Loomis, B.A.

1986-03-01

The swelling of V-base alloys, Type 316 stainless steel, Fe-25Ni-15Cr alloys, ferritic steels, Cu, Ni, Nb-1% Zr, and Mo on neutron irradiation is compared with the swelling for these materials on ion irradiation. The results of this comparison show that utilization of the ion-irradiation technique provides for a discriminative assessment of the potential for swelling of candidate materials for fusion reactors.

Repair-welding technology of irradiated materials - WIM project

International Nuclear Information System (INIS)

Nakata, K.; Oishi, M.

1998-01-01

A new project on the development of repair-welding technology for core internals and reactor (pressure) vessel, consigned by the Ministry of International Trade and Industry (MITI), has been started from October 1997. The objective of the project is classified into three points as follows: (1) to develop repair-welding techniques for neutron irradiated materials, (2) to prove the availability of the techniques for core internals and reactor (pressure) vessel, and (3) to recommend the updated repair-welding for the Technical Rules and Standards. Total planning, neutron irradiation, preparation of welding equipment are now in progress. The materials are austenitic stainless steels and a low alloy steel. Neutron irradiation is performed using test reactors. In order to suppress the helium aggregation along grain boundaries, low heat input welding techniques, such as laser, low heat input TIG and friction weldings, will be applied. (author)

Design and fabrication of irradiation testing capsule for research reactor materials

International Nuclear Information System (INIS)

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

2012-01-01

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

Design and fabrication of irradiation testing capsule for research reactor materials

Energy Technology Data Exchange (ETDEWEB)

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

2012-10-15

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

Irradiation creep experiments on fusion reactor candidate structural materials

International Nuclear Information System (INIS)

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

1991-01-01

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

Growth kinetics of dislocation loops in irradiated ceramic materials

International Nuclear Information System (INIS)

Ryazanov, A.I.; Kinoshita, C.

2002-01-01

Ceramic materials are expected to be applied in the future fusion reactor as radio frequency (RF) windows, toroidal insulating breaks and diagnostic probes. The radiation resistance of ceramic materials, degradation of the electrical properties and radiation induced conductivity of these materials under neutron irradiation are determined by the kinetics of the accumulation of point defects in the matrix and point defect cluster formation (dislocation loops, voids, etc.). Under irradiation, due to the ionization process, excitation of electronic subsystem and covalent type of interaction between atoms the point defects in ceramic materials are characterized by the charge state (e.g. an F + center, an oxygen vacancy with a single trapped electron) and the effective charge. For the investigation of radiation resistance of ceramic materials for future fusion applications it is very important to understand the physical mechanisms of formation and growth of dislocation loops and voids under irradiation taking into account in this system the effective charge of point defects. In the present paper the physical mechanisms of dislocation loop growth in ceramic material are investigated. For this aim a theoretical model is suggested for the description of the kinetics of point defect accumulation in the matrix taking into account the charge state of the point defects and the effect of an electric field on diffusion migration process of charged point defects. A self-consistent system of kinetic equations describing the generation of electrical fields near dislocation loops and diffusion migration of charged point defects in elastic and electrical fields is formulated. The solution of the kinetic equations allows to find the growth rate of dislocation loops in ceramic materials under irradiation taking into account the charge state of the point defects and the effect of electric and elastic stress fields near dislocation loop on the diffusion processes

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.)

Corrosion of electron-irradiated Zr-2.5Nb and Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Woo, O.-T.; McDougall, G.M.; Hutcheon, R.M.; Urbanic, V.F.; Griffiths, M.; Coleman, C.E

2000-07-01

We used 10-MeV electrons to rapidly produce radiation damage in zirconium alloys, investigated whether electrons produced the same microstructural changes as neutrons, then performed post-irradiation corrosion tests to determine whether electron-irradiated materials displayed similar corrosion behavior to neutron-irradiated materials. Two irradiations were completed using 10-MeV electrons with the beam normal to thin disks of material of 4 diameter slightly larger than the beam. The beam distribution. and disk cooling were designed to produce radial temperature and dose distributions having maxima at the disk center. A high-temperature irradiation was performed on annealed Zr-2.5Nb disks, achieving a central dose of 1.3 dpa and at a central temperature of {approx}450 deg C. After irradiation, the samples contained needle-like {beta}-Nb precipitates in the {alpha}-Zr matrix similar to those produced by neutrons. A low-temperature irradiation was performed on half-moon disks of Zr-2.5Nb and Zircaloy-2 pressure tube materials at 310 deg C central temperature and 1.3-dpa central dose. Dislocation loops were observed, again similar to those produced in neutron-irradiated materials. Some of the high-temperature electron-irradiated disks were exposed to 300 deg C moist air (saturated with D{sub 2}O), and in separate tests, high- and low-temperature irradiated disks were corroded in 300 deg C D{sub 2}0 (11.0 pD at room temperature) in an autoclave. Measurements of oxide thickness by Fourier Transform Infrared Reflectance (FTIR) spectroscopy showed that electron irradiation reduced the corrosion rate of Zr-2.5Nb compared with that of unirradiated material, as observed for neutron irradiation. For exposures to moist air and to D{sub 2}O, the theoretical deuterium uptakes for the electron-irradiated materials were, respectively, about 4 times and 1.5 to 2 times those for the unirradiated materials. This is also in good agreement with results for neutron-irradiated pressure

Effects of gamma-rays irradiation on tracking resistance of organic insulating materials

Energy Technology Data Exchange (ETDEWEB)

Du, Boxue; Suzuki, Akio; Kobayashi, Shigeo [Tokyo Univ. of Agriculture and Technology, Koganei (Japan). Faculty of Technology

1996-04-01

This paper describes the influence of gamma-rays irradiation on tracking failure of organic insulating materials by use of the IEC Publ.112 method. Tracking resistance of organic insulating materials under wet polluted condition has been studied by many investigators with a test method of the IEC Publ.112. The investigations on irradiation effects on tracking resistance should be enhanced due to the increasing usage of organic insulating materials in the radiation environments. The tracking resistance seems to be affected by gamma-irradiation, but the knowledge on the influence of gamma-irradiation is quite a few and systematic studies are needed. In this paper, modified polyphenylene oxide, polybutylene naphthalate, modified polycarbonate and polybutylene terephthalate which were irradiated in air until 1x10{sup 7}R and 1x10{sup 8}R with dose rate of 10{sup 6}R/hr using {sup 60}Co gamma-source have been employed. The total dose effects on the number of drops to tracking failure, contact angle and charges of scintillation have been studied. As the total doses are increased, the number of drops to tracking failure decreases with polybutylene terephthalate. On the other hand, the number of drops to tracking failure increases with polybutylene naphthalate and modified polycarbonate when the total doses are increased. The effects of gamma-rays irradiation on tracking failure are due to radiation-induced degradation or cross-linking of organic insulating materials. When the organic insulating materials are degraded by gamma-irradiation, the tracking resistance decreases, but for cross-linking type materials, the tracking resistance increases. (author)

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

International Nuclear Information System (INIS)

Rennich, M.J.

1995-12-01

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

Diffusion and solubility of oxygen in γ-ray irradiated polymer insulation materials

International Nuclear Information System (INIS)

Seguchi, Tadao; Yamamoto, Yasuaki.

1986-03-01

The effects of 60 Co γ-rays irradiation on diffusion and solubility of oxygen in polymer materials for electric cable insulation materials were investigated. The polymers were polyethylene, ethylene-propylene rubber, chlorinated polyethylene, chlorosulphonated polyethylene, and chloroprene rubber. They were pure grade and several types of formulation grade. The sheets of these polymers were irradiated up to 5 - 200 Mrad under vacuum or in oxygen under pressure of 3 - 15 atm at room temperature or at 70 deg C. By a method of gas desorption, the diffusion coefficient (D) and solubility coefficient (S) of oxygen or argon in polymer materials were determined at various temperatures of 10 - 80 deg C. The D and S decreased with increase of dose, and the decrease by irradiation with oxidation was more remarkable than that by irradiation without oxidation. However, the decreases of D and S by irradiation were reduced by the formulation of polymers. The additives in formulated polymers would reduce the reactions of crosslinking or oxidation by γ-ray irradiation. The activation energy of D was scarcely changed by irradiations with and without oxidation. (author)

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)

Irradiation of aluminium alloy materials with electron beam

International Nuclear Information System (INIS)

Konno, Osamu; Masumoto, Kazuyoshi

1982-01-01

It is a theme with a room for discussion to employ the stainless steel composed of longer half-life materials for the vacuum system of accelerators, from the viewpoint of radiation exposure. Therefore, it is desirable to use aluminium of shorter half-life in place of stainless steel. As a result of investigation on the above theme in the 1.2 GeV electron linac project in Tohoku University, it has been concluded that aluminium alloy vacuum chambers can reduce exposure dose by about one or two figures as compared with stainless steel ones. Of course, aluminium alloy contains trace amounts of Mg, Si, Ti, Cr, Mn, Fe, Zn, Cu and others. Therefore, four kinds of aluminium alloy considered to be usable have been examined for induced radioactivity by electron beam irradiation. Stainless steel SUS 304 has been also irradiated for comparison. Radiation energy has been 30 MeV and 200 MeV. When stainless steel and aluminium alloy were compared, aluminium alloy was very effective for reducing surface dose in low energy irradiation. In 200 MeV irradiation, the dose ratio of aluminium alloy to stainless steel became 1/30 to 1/100 after one week, though the dose difference between these two materials became smaller in 100 days or more after irradiation. If practical inspection and repair are implemented during the period from a few days to one week after shutdown, the aluminium alloy is preferable for exposure dose reduction even in high energy irradiation. (Wakatsuki, Y.)

Neutron-Irradiated Samples as Test Materials for MPEX

International Nuclear Information System (INIS)

Ellis, Ronald James; Rapp, Juergen

2015-01-01

Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility

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

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

International Nuclear Information System (INIS)

Wang Zhiguang

2007-01-01

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

Radiation-Induced Fluidity and Glass-Liquid Transition in Irradiated Amorphous Materials

International Nuclear Information System (INIS)

Ojovan, M.I.

2009-01-01

This paper describes the fluidity behaviour of continuously irradiated glasses using the Congruent Bond Lattice model in which broken bonds 'configurons' facilitate the flow. Irradiation breaks the bonds creating configurons which at high concentrations provide the transition of material from the glassy to liquid state. An explicit equation of viscosity has been derived which gives results in agreement with experimental data. This equation provides correct viscosity data for non-irradiated materials and shows a significant increase of fluidity in radiation fields. It demonstrates a decrease of activation energy of flow for irradiated glasses. A simple equation for glass-transition temperature was also obtained which shows that irradiated glasses have lower glass transition temperatures and are readily transformed from glassy to liquid state e.g. fluidized in strong radiation fields. (authors)

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.)

Induction of materials for mutation breeding of strawberry (Fragaria × Ananassa) by gamma irradiation

International Nuclear Information System (INIS)

Le Ngoc Trieu; Nguyen Tuong Mien; Le Tien Thanh; Huynh Thi Trung; Pham Van Nhi; Vu Thi Trac

2015-01-01

From collected New Zealand strawberry runners, micropropagation was executed to establish 500 shoot clusters for investigation effect of Gamma ray irradiation doses on survival rate. LD_5_0 at 52 Gy was recorded 45 days after re-injection and used as base for choosing 5 irradiation doses of 20, 40, 60, 80, 100 Gy for creation potentially existent mutant materials. 30 shoot clusters were irradiated at each chosen dose. Irradiated material was propagated by in vitro techniques to achieve 300 plantlets/chosen dose. There was no recorded alteration in survival rate and other morphological characteristics of irradiated materials compared to the control in nursery period. These materials were transplanted to plastic greenhouse to screen the mutant. (author)

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

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

International Nuclear Information System (INIS)

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

1993-02-01

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

Effects of material property changes on irradiation assisted stress corrosion cracking

Energy Technology Data Exchange (ETDEWEB)

Nakano, Morihito; Fukuya, Koji; Fujii, Katsuhiko [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2002-09-01

Irradiation assisted stress corrosion cracking (IASCC) susceptibility and radiation-induced material changes in microstructure and microchemistry under pressurized water reactor (PWR) environment were examined on irradiated stainless steels (SSs), post-irradiation annealed SSs and post-irradiation deformed SS. The yield stress and grain boundary segregation were considerably high in SSs highly irradiated to 1-8 x 10{sup 26}n/m{sup 2} (E > 0.1 MeV) in PWR at 290-320degC, resulting in a high IASCC susceptibility. Following post-irradiation annealing of highly irradiated SSs, IASCC susceptibility showed significant recovery from 89% (as-irradiated) to 8% (550degC) of %IGSCC, while the hardness recovered from Hv375 (400degC) to Hv315 (550degC). Apparent recovery of segregation at grain boundaries was not observed. The SSs irradiated to 5.3 x 10{sup 24}n/m{sup 2} (E>1MeV) in the Japan Materials Testing Reactor (JMTR) at < 400degC, which had grain boundary segregation and low hardness, showed no IASCC susceptibility. Due to post-irradiation deforming for JMTR irradiated SS, the hardness increased but IASCC did not occur. These results suggested that the hardening would be a key factor for IASCC initiation under PWR hydrogenated water and that a yield stress threshold for IASCC initiation under slow strain rate tensile (SSRT) testing would the about 600MPa. (author)

Effects of material property changes on irradiation assisted stress corrosion cracking

International Nuclear Information System (INIS)

Nakano, Morihito; Fukuya, Koji; Fujii, Katsuhiko

2002-01-01

Irradiation assisted stress corrosion cracking (IASCC) susceptibility and radiation-induced material changes in microstructure and microchemistry under pressurized water reactor (PWR) environment were examined on irradiated stainless steels (SSs), post-irradiation annealed SSs and post-irradiation deformed SS. The yield stress and grain boundary segregation were considerably high in SSs highly irradiated to 1-8 x 10 26 n/m 2 (E > 0.1 MeV) in PWR at 290-320degC, resulting in a high IASCC susceptibility. Following post-irradiation annealing of highly irradiated SSs, IASCC susceptibility showed significant recovery from 89% (as-irradiated) to 8% (550degC) of %IGSCC, while the hardness recovered from Hv375 (400degC) to Hv315 (550degC). Apparent recovery of segregation at grain boundaries was not observed. The SSs irradiated to 5.3 x 10 24 n/m 2 (E>1MeV) in the Japan Materials Testing Reactor (JMTR) at < 400degC, which had grain boundary segregation and low hardness, showed no IASCC susceptibility. Due to post-irradiation deforming for JMTR irradiated SS, the hardness increased but IASCC did not occur. These results suggested that the hardening would be a key factor for IASCC initiation under PWR hydrogenated water and that a yield stress threshold for IASCC initiation under slow strain rate tensile (SSRT) testing would the about 600MPa. (author)

Development of Multiscale Materials Modeling Techniques and Coarse- Graining Strategies for Predicting Materials Degradation in Extreme Irradiation Environments

Energy Technology Data Exchange (ETDEWEB)

Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States)

2016-01-12

Exposure of metallic structural materials to irradiation environments results in significant microstructural evolution, property changes and performance degradation, which limits the extended operation of current generation light water reactors and restricts the design of advanced fission and fusion reactors [1-8]. This effect of irradiation on materials microstructure and properties is a classic example of an inherently multiscale phenomenon, as schematically illustrated in Figure 1a. Pertinent processes range from the atomic nucleus to structural component length scales, spanning more than 15 orders of magnitude. Time scales bridge more than 22 orders of magnitude, with the shortest being less than a femtosecond [1,8]. Further, the mix of radiation-induced features formed and the corresponding property degradation depend on a wide range of material and irradiation variables. This emphasizes the importance of closely integrating models with high-resolution experimental characterization of the evolving radiation- damaged microstructure, including measurements performed in-situ during irradiation. In this article, we review some recent successes through the use of closely coordinated modeling and experimental studies of the defect cluster evolution in irradiated body-centered cubic materials, followed by a discussion of outstanding challenges still to be addressed, which are necessary for the development of comprehensive models of radiation effects in structural materials.

Preparation of silica-based hybrid materials by gamma irradiation

International Nuclear Information System (INIS)

Gomes, S.R.; Margaca, F.M.A.; Miranda Salvado, I.M.; Ferreira, L.M.; Falcao, A.N.

2006-01-01

Gamma-ray irradiation is well known to promote the crosslinking of polymer chains. The method is now used by the authors to prepare hybrid materials from a mixture of polymer and metallic alkoxides of silicium and zirconium that are usually obtained via the sol-gel process. Macroscopically homogeneous and transparent hybrid materials have been obtained by γ-irradiation of polydimethylsiloxane (PDMS), tetraethylorthosilicate (TEOS) and zirconium propoxide (PrZr). The influence of several parameters has been studied. The dose rate was found to have no significant impact in the prepared material. The polymer molecular weight was also observed not to play any special role. It was found that all irradiated samples consist of a polymer gel matrix. In the case where both alkoxides are present there are inorganic oxide regions linked to the PDMS network. However when one of the alkoxides is absent there is no formation of inorganic oxide regions linked to the polymer matrix, there being only a few individual derived molecules of the other alkoxide linked to the polymer

Sampling by electro-erosion on irradiated materials

International Nuclear Information System (INIS)

Riviere, M.; Pizzanelli, J.P.

1986-05-01

Sampling on irradiated materials, in particular for mechanical property study of steels in the FAST NEUTRON program needed the set in a hot cell of a machining device by electroerosion. This device allows sampling of tenacity, traction, resilience test pieces [fr

Irradiation effect of the insulating materials for fusion superconducting magnets at cryogenic temperature

Science.gov (United States)

Kobayashi, Koji; Akiyama, Yoko; Nishijima, Shigehiro

2017-09-01

In ITER, superconducting magnets should be used in such severe environment as high fluence of fast neutron, cryogenic temperature and large electromagnetic forces. Insulating material is one of the most sensitive component to radiation. So radiation resistance on mechanical properties at cryogenic temperature are required for insulating material. The purpose of this study is to evaluate irradiation effect of insulating material at cryogenic temperature by gamma-ray irradiation. Firstly, glass fiber reinforced plastic (GFRP) and hybrid composite were prepared. After irradiation at room temperature (RT) or liquid nitrogen temperature (LNT, 77 K), interlaminar shear strength (ILSS) and glass-transition temperature (Tg) measurement were conducted. It was shown that insulating materials irradiated at room temperature were much degraded than those at cryogenic temperature.

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)

HFR irradiation testing of fusion materials

International Nuclear Information System (INIS)

Conrad, R.; von der Hardt, P.; Loelgen, R.; Scheurer, H.; Zeisser, P.

1984-01-01

The present and future role of the High Flux Reactor Petten for fusion materials testing has been assessed. For practical purposes the Tokamak-based fusion reactor is chosen as a point of departure to identify material problems and materials data needs. The identification is largely based on the INTOR and NET design studies, the reported programme strategies of Japan, the U.S.A. and the European Communities for technical development of thermonuclear fusion reactors and on interviews with several experts. Existing and planned irradiation facilities, their capabilities and limitations concerning materials testing have been surveyed and discussed. It is concluded that fission reactors can supply important contributions for fusion materials testing. From the point of view of future availability of fission testing reactors and their performance it appears that the HFR is a useful tool for materials testing for a large variety of materials. Prospects and recommendations for future developments are given

In-situ high temperature irradiation setup for temperature dependent structural studies of materials under swift heavy ion irradiation

International Nuclear Information System (INIS)

Kulriya, P.K.; Kumari, Renu; Kumar, Rajesh; Grover, V.; Shukla, R.; Tyagi, A.K.; Avasthi, D.K.

2015-01-01

An in-situ high temperature (1000 K) setup is designed and installed in the materials science beam line of superconducting linear accelerator at the Inter-University Accelerator Centre (IUAC) for temperature dependent ion irradiation studies on the materials exposed with swift heavy ion (SHI) irradiation. The Gd 2 Ti 2 O 7 pyrochlore is irradiated using 120 MeV Au ion at 1000 K using the high temperature irradiation facility and characterized by ex-situ X-ray diffraction (XRD). Another set of Gd 2 Ti 2 O 7 samples are irradiated with the same ion beam parameter at 300 K and simultaneously characterized using in-situ XRD available in same beam line. The XRD studies along with the Raman spectroscopic investigations reveal that the structural modification induced by the ion irradiation is strongly dependent on the temperature of the sample. The Gd 2 Ti 2 O 7 is readily amorphized at an ion fluence 6 × 10 12 ions/cm 2 on irradiation at 300 K, whereas it is transformed to a radiation-resistant anion-deficient fluorite structure on high temperature irradiation, that amorphized at ion fluence higher than 1 × 10 13 ions/cm 2 . The temperature dependent ion irradiation studies showed that the ion fluence required to cause amorphization at 1000 K irradiation is significantly higher than that required at room temperature irradiation. In addition to testing the efficiency of the in-situ high temperature irradiation facility, the present study establishes that the radiation stability of the pyrochlore is enhanced at higher temperatures

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.)

Electron-beam-irradiation-induced crystallization of amorphous solid phase change materials

Science.gov (United States)

Zhou, Dong; Wu, Liangcai; Wen, Lin; Ma, Liya; Zhang, Xingyao; Li, Yudong; Guo, Qi; Song, Zhitang

2018-04-01

The electron-beam-irradiation-induced crystallization of phase change materials in a nano sized area was studied by in situ transmission electron microscopy and selected area electron diffraction. Amorphous phase change materials changed to a polycrystalline state after being irradiated with a 200 kV electron beam for a long time. The results indicate that the crystallization temperature strongly depends on the difference in the heteronuclear bond enthalpy of the phase change materials. The selected area electron diffraction patterns reveal that Ge2Sb2Te5 is a nucleation-dominated material, when Si2Sb2Te3 and Ti0.5Sb2Te3 are growth-dominated materials.

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

International Nuclear Information System (INIS)

Mann, F.M.

1988-08-01

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

Effects of CTR irradiation on the mechanical properties of structural materials

International Nuclear Information System (INIS)

Wiffen, F.W.

1976-11-01

Mechanical properties of CTR structural materials are important in determining the reliability and economics of fusion power. Furthermore, these properties are significantly affected by the high neutron flux experienced by components in the regions near the plasma of the fusion reactor. In general, irradiation hardens the material and leads to a reduction in ductility. An exception to this is in some complex engineering alloys where either hardening or softening can be observed depending on the alloy and the irradiation conditions. Regardless of this restriction, irradiation usually leads to a reduction in ductility. Available tensile data examined in this paper show that significant ductility reduction can be found for irradiation conditions typical of CTR operation. Consideration of these effects show that extensive work will be needed to fully establish the in-service properties of CTR structures. This information will be used by designers to develop conditions and design philosophies adapted to avoid the most deleterious conditions and minimize stresses on structures on reactor design. The information will also be used as input to alloy development programs with goals of producing materials more resistant to property degradation during irradiation. It is clear that a great deal of additional work will be required both to understand the effect of CTR irradiation on properties and to develop optimal alloys for this application

Chemical and physical change of packaging materials for food by γ-ray irradiation

International Nuclear Information System (INIS)

Kawamura, Yoko; Takeda, Yuiko; Yamada, Takashi

1998-01-01

Packaging materials for food made of polyethylene, polypropylene and polystyrene were irradiated with 60 Co γ-ray. Exposure was 10, 30 and 50 kGy at 5 kGy/h exposure rate. With irradiating, all packaging materials of polyethylene and polypropylene produced volatile substances, for example, aldehydes, ketones and alcohols, especially, large amount of acetic acid and acetone. These volatile compounds were not observed in the sample unirradiated and increased with increasing exposure. Accordingly, it is concluded that they were decomposition products depend on irradiation. Polypropylene products were much more easily decomposed than polyethylene one because much more kinds and amount of volatile products were formed. However, on polystyrene products, content of styrene and ethylbenzene, monomer of raw materials, were reduced by irradiation and small amount of volatile substances were formed. These results proved its resistance to irradiation. (S.Y.)

Nanostructured Solar Irradiation Control Materials for Solar Energy Conversion

Science.gov (United States)

Kang, Jinho; Marshall, I. A.; Torrico, M. N.; Taylor, C. R.; Ely, Jeffry; Henderson, Angel Z.; Kim, J.-W.; Sauti, G.; Gibbons, L. J.; Park, C.;

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.

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

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.

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

International Nuclear Information System (INIS)

Conlin, J.A.

1985-08-01

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

A Study on the Thermal Neutron Filter for the Irradiation of Electronic Materials at HANARO

Energy Technology Data Exchange (ETDEWEB)

Yang, Seong Woo; Kim, Sung Ryul; Park, Seung Jae; Shin, Yoon Taeg; Cho, Man Soon; Cho, Kee Nam [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The representative example is a technique of making the semiconductor with the transmutation using the pure Si. This NTD (Neutron Transmutation Doping) Si is used as a high-quality semiconductor because it has a uniform resistance. Likewise, the electronic materials are being investigated to improve the performance of material using the neutron irradiation method. The mechanism for reaction between the electronic materials and the neutrons depends on the energy of the neutron. Capturing reaction by thermal neutrons causes the transmutation and a lot of defects are made by fast neutrons. The study for the effect by such neutron energy is necessary to understand the performance improvement of the irradiated electronic materials. The thermal neutron filter was investigated to be used for the irradiation of electronic materials at HANARO. IP irradiation hole was selected and the irradiation device was designed. The analysis was conducted considering four candidate materials.

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.

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

International Nuclear Information System (INIS)

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

2013-01-01

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

SEM analysis for irradiated materials

International Nuclear Information System (INIS)

Liu Xiaosong; Yao Liang

2008-06-01

A radiation-proof Scanning Electron Microscope (SEM) system is introduced. It has been widely used in various areas. For analyzing radioactive samples, normal SEM system needs lots of alterations. Based on KYKY-2800B SEM, the sample room, belt line, operating table and aerator were updated. New radiation-proof SEM system has used to analytic surface contaminated samples and RPV materials samples. An elementary means of SEM analysis for radioactive samples was studied, and this examination supported some available references for further irradiated fuel researches. (authors)

Neutron irradiation effects on superconducting and stabilizing materials for fusion magnets

International Nuclear Information System (INIS)

Maurer, W.

1984-05-01

Available low-temperature neutron irradiation data for the superconductors NbTi and Nb 3 Sn and the stabilization materials Cu and Al are collected and maximum tolerable doses for these materials are defined. A neutron flux in a reactor of about 10 9 n/cm 2 s at the magnet position is expected. However, in fusion experiments the flux can be higher by an order of magnitude or more. The energy spectrum is similar to a fission reactor. A fluence of about 10 18 n/cm 2 results during the lifetime of a fusion magnet (about 20 full power years). At this fluence and energy spectrum no severe degradation of the superconducting properties of NbTi and Nb 3 Sn will occur. But the radiation-induced resistivity is for Cu about a twentieth of the room temperature resistivity and a tenth for Al. (orig.) [de

Neutron irradiation test of copper alloy/stainless steel joint materials

International Nuclear Information System (INIS)

Yamada, Hirokazu; Kawamura, Hiroshi

2006-01-01

As a study about the joint technology of copper alloy and stainless steel for utilization as cooling piping in International Thermonuclear Experimental Reactor (ITER), Al 2 O 3 -dispersed strengthened copper or CuCrZr was jointed to stainless steel by three kinds of joint methods (casting joint, brazing joint and friction welding method) for the evaluation of the neutron irradiation effect on joints. A neutron irradiation test was performed to three types of joints and each copper alloy. The average value of fast neutron fluence in this irradiation test was about 2 x 10 24 n/m 2 (E>1 MeV), and the irradiation temperature was about 130degC. As post-irradiation examinations, tensile tests, hardness tests and observation of fracture surface after the tensile tests were performed. All type joints changed to be brittle by the neutron irradiation effect like each copper alloy material, and no particular neutron irradiation effect due to the effect of joint process was observed. On the casting and friction welding, hardness of copper alloy near the joint boundary changed to be lower than that of each copper alloy by the effect of joint procedure. However, tensile strength of joints was almost the same as that of each copper alloy before/after neutron irradiation. On the other hand, tensile strength of joints by brazing changed to be much lower than CuAl-25 base material by the effect of joint process before/after neutron irradiation. Results in this study showed that the friction welding method and the casting would be able to apply to the joint method of piping in ITER. This report is based on the final report of the ITER Engineering Design Activities (EDA). (author)

Swedish studies on irradiation effect in structural materials

Energy Technology Data Exchange (ETDEWEB)

Grounes, M; Myers, H P

1962-12-15

A brief description of work in hand at AB Atomenergi concerning the effects of neutron irradiation on structural materials is given. Some recent data is listed for the following pressure vessel steels 2103/R3 as used in the Aagesta reactor, SIS 142103, NO345, Fortiweld and weld metal OK 54 P. Zircaloy-2 has been studied regarding the combined effects of neutron irradiation and hydrogen content on tensile properties. The difficulties associated with determination of neutron dose and the correlation of damage with dose and neutron energy spectrum are discussed.

Swedish studies on irradiation effect in structural materials

International Nuclear Information System (INIS)

Grounes, M.; Myers, H.P.

1962-12-01

A brief description of work in hand at AB Atomenergi concerning the effects of neutron irradiation on structural materials is given. Some recent data is listed for the following pressure vessel steels 2103/R3 as used in the Aagesta reactor, SIS 142103, NO345, Fortiweld and weld metal OK 54 P. Zircaloy-2 has been studied regarding the combined effects of neutron irradiation and hydrogen content on tensile properties. The difficulties associated with determination of neutron dose and the correlation of damage with dose and neutron energy spectrum are discussed

Studies on gamma irradiated rubber materials

Science.gov (United States)

Lungu, I. B.; Stelescu, M. D.; Cutrubinis, M.

2018-01-01

Due to the increase in use and production of polymer materials, there is a constant pressure of finding a solution to more environmental friendly composites. Beside the constant effort of recycling used materials, it seems more appropriate to manufacture and use biodegradable and renewable row materials. Natural polymers like starch, cellulose, lignin etc are ideal for preparing biodegradable composites. Some of the dynamic markets that use polymer materials are the food and pharmaceutical industries. Because of their desinfastation and sometimes sterility requirements, different treatment processes are applied, one of it being radiation treatment. The scope of this paper is to analyze the mechanical behaviour of rubber based materials irradiated with gamma rays at four medium doses, 30.1 kGy, 60.6 kGy, 91 kGy and 121.8 kGy. The objectives are the following: to identify the optimum radiation dose in order to obtain a good mechanical behaviour and to identify the mechanical behaviour of the material when adding different quantities of natural filler (20 phr, 60 phr and 100 phr).

Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Preliminary results

International Nuclear Information System (INIS)

Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

1993-01-01

Candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at temperatures of either 60 or 250 degrees C. Preliminary results have been obtained for several of these materials irradiated at 60 degrees C. The results show that irradiation at this temperature reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The unloading compliance technique developed for the subsize disk compact specimens works quite well, particularly for materials with lower toughness. Specimens of materials with very high toughness deform excessively, and this results in experimental difficulties

Whole-Genome Sequencing in Microbial Forensic Analysis of Gamma-Irradiated Microbial Materials.

Science.gov (United States)

Broomall, Stacey M; Ait Ichou, Mohamed; Krepps, Michael D; Johnsky, Lauren A; Karavis, Mark A; Hubbard, Kyle S; Insalaco, Joseph M; Betters, Janet L; Redmond, Brady W; Rivers, Bryan A; Liem, Alvin T; Hill, Jessica M; Fochler, Edward T; Roth, Pierce A; Rosenzweig, C Nicole; Skowronski, Evan W; Gibbons, Henry S

2016-01-15

Effective microbial forensic analysis of materials used in a potential biological attack requires robust methods of morphological and genetic characterization of the attack materials in order to enable the attribution of the materials to potential sources and to exclude other potential sources. The genetic homogeneity and potential intersample variability of many of the category A to C bioterrorism agents offer a particular challenge to the generation of attributive signatures, potentially requiring whole-genome or proteomic approaches to be utilized. Currently, irradiation of mail is standard practice at several government facilities judged to be at particularly high risk. Thus, initial forensic signatures would need to be recovered from inactivated (nonviable) material. In the study described in this report, we determined the effects of high-dose gamma irradiation on forensic markers of bacterial biothreat agent surrogate organisms with a particular emphasis on the suitability of genomic DNA (gDNA) recovered from such sources as a template for whole-genome analysis. While irradiation of spores and vegetative cells affected the retention of Gram and spore stains and sheared gDNA into small fragments, we found that irradiated material could be utilized to generate accurate whole-genome sequence data on the Illumina and Roche 454 sequencing platforms. Copyright © 2016, American Society for Microbiology. All Rights Reserved.

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.

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.)

A study on the irradiation effect of reactor materials using a cyclotron

Energy Technology Data Exchange (ETDEWEB)

Hong, Joon Hwa; Ji, Se Hwan; Kang, Yung Hwan; Park, Duk Keun; Park, Jong Man; Lee, Bong Sang; Oh, Jong Myung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-12-01

The objectives of the present study are to develop the simulation techniques of neutron irradiation through ion irradiation using a Cyclotron and small specimen techniques and to evaluate radiation effects of reactor materials. Effects of proton or neutron irradiation on domestic 12Cr-1MoV and SA508-3 steels were evaluated by small scale specimen test techniques, i.e, small punch and miniaturized tensile test. In order to study the radiation damage mechanism, irradiation effects of the steels were investigated by means of property change tests such as microstructure, physical and thermal properties. Feasibility study on application of a magnetic non-destructive methods to evaluate radiation effects on RPV materials was performed. 109 figs, 12 tabs, 102 refs. (Author).

A study on the irradiation effect of reactor materials using a cyclotron

International Nuclear Information System (INIS)

Hong, Joon Hwa; Ji, Se Hwan; Kang, Yung Hwan; Park, Duk Keun; Park, Jong Man; Lee, Bong Sang; Oh, Jong Myung

1994-12-01

The objectives of the present study are to develop the simulation techniques of neutron irradiation through ion irradiation using a Cyclotron and small specimen techniques and to evaluate radiation effects of reactor materials. Effects of proton or neutron irradiation on domestic 12Cr-1MoV and SA508-3 steels were evaluated by small scale specimen test techniques, i.e, small punch and miniaturized tensile test. In order to study the radiation damage mechanism, irradiation effects of the steels were investigated by means of property change tests such as microstructure, physical and thermal properties. Feasibility study on application of a magnetic non-destructive methods to evaluate radiation effects on RPV materials was performed. 109 figs, 12 tabs, 102 refs. (Author)

Effect of γ-irradiation on commercial polypropylene based mono and multi-layered retortable food packaging materials

Science.gov (United States)

George, Johnsy; Kumar, R.; Sajeevkumar, V. A.; Sabapathy, S. N.; Vaijapurkar, S. G.; Kumar, D.; Kchawahha, A.; Bawa, A. S.

2007-07-01

Irradiation processing of food in the prepackaged form may affect chemical and physical properties of the plastic packaging materials. The effect of γ-irradiation doses (2.5-10.0 kGy) on polypropylene (PP)-based retortable food packaging materials, were investigated using Fourier transform infrared (FTIR) spectroscopic analysis, which revealed the changes happening to these materials after irradiation. The mechanical properties decreased with irradiation while oxygen transmission rate (OTR) was not affected significantly. Colour measurement indicated that Nylon 6 containing multilayer films became yellowish after irradiation. Thermal characterization revealed the changes in percentage crystallinity.

Effect of γ-irradiation on commercial polypropylene based mono and multi-layered retortable food packaging materials

International Nuclear Information System (INIS)

George, Johnsy; Kumar, R.; Sajeevkumar, V.A.; Sabapathy, S.N.; Vaijapurkar, S.G.; Kumar, D.; Kchawahha, A.; Bawa, A.S.

2007-01-01

Irradiation processing of food in the prepackaged form may affect chemical and physical properties of the plastic packaging materials. The effect of γ-irradiation doses (2.5-10.0 kGy) on polypropylene (PP)-based retortable food packaging materials, were investigated using Fourier transform infrared (FTIR) spectroscopic analysis, which revealed the changes happening to these materials after irradiation. The mechanical properties decreased with irradiation while oxygen transmission rate (OTR) was not affected significantly. Colour measurement indicated that Nylon 6 containing multilayer films became yellowish after irradiation. Thermal characterization revealed the changes in percentage crystallinity

Effect of {gamma}-irradiation on commercial polypropylene based mono and multi-layered retortable food packaging materials

Energy Technology Data Exchange (ETDEWEB)

George, Johnsy [Defence Food Research Laboratory, Siddarthanagar, Mysore, Karnataka 570011 (India)]. E-mail: g.johnsy@gmail.com; Kumar, R. [Defence Food Research Laboratory, Siddarthanagar, Mysore, Karnataka 570011 (India); Sajeevkumar, V.A. [Defence Food Research Laboratory, Siddarthanagar, Mysore, Karnataka 570011 (India); Sabapathy, S.N. [Defence Food Research Laboratory, Siddarthanagar, Mysore, Karnataka 570011 (India); Vaijapurkar, S.G. [Defence Laboratory, Ratanada Palace, Jodhpur, Rajastan 342011 (India); Kumar, D. [Defence Laboratory, Ratanada Palace, Jodhpur, Rajastan 342011 (India); Kchawahha, A. [Defence Laboratory, Ratanada Palace, Jodhpur, Rajastan 342011 (India); Bawa, A.S. [Defence Food Research Laboratory, Siddarthanagar, Mysore, Karnataka 570011 (India)

2007-07-15

Irradiation processing of food in the prepackaged form may affect chemical and physical properties of the plastic packaging materials. The effect of {gamma}-irradiation doses (2.5-10.0 kGy) on polypropylene (PP)-based retortable food packaging materials, were investigated using Fourier transform infrared (FTIR) spectroscopic analysis, which revealed the changes happening to these materials after irradiation. The mechanical properties decreased with irradiation while oxygen transmission rate (OTR) was not affected significantly. Colour measurement indicated that Nylon 6 containing multilayer films became yellowish after irradiation. Thermal characterization revealed the changes in percentage crystallinity.

Irradiation capability of Japanese materials test reactor for water chemistry experiments

International Nuclear Information System (INIS)

Hanawa, Satoshi; Hata, Kuniki; Chimi, Yasuhiro; Nishiyama, Yutaka; Nakamura, Takehiko

2012-09-01

Appropriate understanding of water chemistry in the core of LWRs is essential as chemical species generated due to water radiolysis by neutron and gamma-ray irradiation govern corrosive environment of structural materials in the core and its periphery, causing material degradation such as stress corrosion cracking. Theoretical model calculation such as water radiolysis calculation gives comprehensive understanding of water chemistry at irradiation field where we cannot directly monitor. For enhancement of the technology, accuracy verification of theoretical models under wide range of irradiation conditions, i.e. dose rate, temperature etc., with well quantified in-pile measurement data is essential. Japan Atomic Energy Agency (JAEA) has decided to launch water chemistry experiments for obtaining data that applicable to model verification as well as model benchmarking, by using an in-pile loop which will be installed in the Japan Materials Testing Reactor (JMTR). In order to clarify the irradiation capability of the JMTR for water chemistry experiments, preliminary investigations by water radiolysis / ECP model calculations were performed. One of the important irradiation conditions for the experiments, i.e. dose rate by neutron and gamma-ray, can be controlled by selecting irradiation position in the core. In this preliminary study, several representative irradiation positions that cover from highest to low absorption dose rate were chosen and absorption dose rate at the irradiation positions were evaluated by MCNP calculations. As a result of the calculations, it became clear that the JMTR could provide the irradiation conditions close to the BWR. The calculated absorption dose rate at each irradiation position was provided to water radiolysis calculations. The radiolysis calculations were performed under various conditions by changing absorption dose rate, water chemistry of feeding water etc. parametrically. Qualitatively, the concentration of H 2 O 2 , O 2 and

Application of electron irradiation to food containers and packaging materials

International Nuclear Information System (INIS)

Ueno, Koji

2010-01-01

Problems caused by microbial contamination and hazardous chemicals have attracted much attention in the food industry. The number of systems such as hygienic management systems and Hazard Analysis Critical Control Point (HACCP) systems adopted in the manufacturing process is increasing. As manufacturing process control has become stricter, stricter control is also required for microbial control for containers and packaging materials (from disinfection to sterilization). Since safe and reliable methods for sterilizing food containers and packaging materials that leave no residue are required, electron beam sterilization used for medical equipment has attracted attention from the food industry. This paper describes an electron irradiation facility, methods for applying electron beams to food containers and packaging materials, and products irradiated with electron beams. (author)

Microstructural evolution of CANDU spacer material Inconel X-750 under in situ ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Zhang, He Ken [Department of Mechanical and Materials Engineering, Queen’s University Kingston, Ontario K7L 3N6 (Canada); Yao, Zhongwen, E-mail: yaoz@me.queensu.ca [Department of Mechanical and Materials Engineering, Queen’s University Kingston, Ontario K7L 3N6 (Canada); Judge, Colin; Griffiths, Malcolm [Deformation Technology Branch, AECL, Chalk River Laboratories Chalk River, Ontario K0J 1J0 (Canada)

2013-11-15

Highlights: •γ′ Disordered at low dose. •Cascade induced SFTs were observed in alloy X-750. •No cavities were found from mono heavy ions irradiated samples. -- Abstract: Work on Inconel® X-750 spacers removed from CANDU® reactors has shown that they become embrittled and there is development of many small cavities within the metal matrix and along grain boundaries. In order to emulate the neutron irradiation induced microstructural changes, heavy ion irradiations (1 MeV Kr{sup 2+} ions) were performed while observing the damage evolution using an intermediate voltage electron microscope (IVEM) operating at 200 kV. The irradiations were carried out at various temperatures 60–400 °C. The principal strengthening phase, γ′, was disordered at low doses (∼0.06 dpa) during the irradiation. M{sub 23}C{sub 6} carbides were found to be stable up to 5.4 dpa. Lattice defects consisted mostly of stacking fault tetrahedras (SFTs), 1/2<1 1 0> perfect loops and small 1/3<1 1 1> faulted Frank loops. The ratio of SFT number density to loop number density for each irradiation condition was found to be neither temperature nor dose dependent. Under the operation of the ion beam the SFT production was very rapid, with no evidence for further growth once formed, indicating that they probably formed as a result of cascade collapse in a single cascade. The number density of the defects was found to saturate at low dose (∼0.68 dpa). No cavities were observed regardless of the irradiation temperature between 60 °C and 400 °C for doses up to 5.4 dpa. In contrast, cavities have been observed after neutron irradiation in the same material at similar doses and temperatures indicating that helium, produce during neutron irradiation, may be essential for the nucleation and growth of cavities.

Gamma irradiation of yellow and blue colorants in polystyrene packaging materials

International Nuclear Information System (INIS)

Komolprasert, V.; Diel, Todd; Sadler, G.

2006-01-01

The effect of 10- and 20-kGy gamma irradiation was studied on chromophtal yellow 2RLTS (Yellow 110-2, 3, 4, 5-tetrachloro-6-cyanobenzoic acid) and Irgalite Blue GBP (copper (II) phthalocyanine blue) colorants, which were added to polystyrene (PS) material used to package food prior to irradiation. Analytical results obtained suggest that irradiation did not generate any new chemicals in the PS polymer containing either yellow or blue colorant at a concentration of up to 1% (w/w). Both yellow and blue colorants are relatively stable to gamma irradiation

Irradiating strand material

International Nuclear Information System (INIS)

Austin, J.R.; Brown, M.J.; Loan, L.D.

1975-01-01

Conductors covered with insulation which is to be irradiated are passed between two groups of coaxial sheaves mounted rotatably individually. Successive sections of the conductors are advanced past the window of one accelerator head, around the associated sheave or sheaves, and then past the window of another accelerator head. The accelerators face in substantially opposite directions and are staggered along the paths of the conductors to avoid any substantial overlap of the electron beams associated therewith. The windows extend vertically to encompass all the generally horizontal passes of the conductors as between the two groups of sheaves. Preferably, conductors are strung-up between the sheaves in a modified figure eight pattern. The pattern is a figure eight modified to intermittently include a pass between the sheaves which is parallel to a line joining the axes of the two groups of sheaves. This reverses the direction of travel of the conductors and optimizes the uniformity of exposure of the cross sectional area of the insulation of the conductors to irradiation. The use of a figure eight path for the conductors causes the successive sections of the conductor to turn about the longitudinal axes thereof as they are advanced around the sheaves. In this way the insulation is more uniformly irradiated. In a preferred embodiment, twisted conductor pairs may be irradiated. The twist accentuates the longitudinal turning of the conductor pair. The irradiation of twisted pairs achieves obvious manufacturing economies while avoiding the necessity of having to twist irradiation cross-linked conductors

Post Irradiation Mechanical Behaviour of Three EUROFER Joints

International Nuclear Information System (INIS)

Lucon, E.; Leenaers, A.; Vandermeulen, W.

2006-01-01

The post-irradiation mechanical properties of three EUROFER joints (two diffusion joints and one TIG weld) have been characterized after irradiation to 1.8 dpa at 300 degrees Celsius in the BR-2 reactor. Tensile, KLST impact and fracture toughness tests have been performed. Based on the results obtained and on the comparison with data from EUROFER base material irradiated under similar conditions, the post-irradiation mechanical behaviour of both diffusion joints (laboratory and mock-up) appears similar to that of the base material. The properties of the TIG joint are affected by the lack of a post-weld heat treatment, which causes the material from the upper part of the weld to be significantly worse than that of the lower region. Thus, specimens from the upper layer exhibit extremely pronounced hardening and embrittlement caused by irradiation. The samples extracted from the lower layer show much better resistance to neutron exposure, although their measured properties do not match those of the diffusion joints. The results presented demonstrate that diffusion joining can be a very promising technique.

Tests on irradiated magnet-insulator materials

International Nuclear Information System (INIS)

Schmunk, R.E.; Miller, L.G.; Becker, H.

1983-01-01

Fusion-reactor coils, located in areas where they will be only partially shielded, must be fabricated from materials which are as resistant to radiation as possible. They will probably incorporate resistive conductors with either water or cryogenic cooling. Inorganic insulators have been recommended for these situations, but the possibility exists that some organic insulators may be usuable as well. Results were previously reported for irradiation and testing of three glass reinforced epoxies: G-7, G-10, and G-11. Thin disks of these materials, nominally 0.5 mm thick by 11.1 mm diameter, were tested in compressive fatigue, a configuration and loading which represents reasonably well the magnet environment. In that work G-10 was shown to withstand repeated loading to moderately high stress levels without failure, and the material survived better at liquid nitrogen temperature than at room temperature

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.)

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

International Nuclear Information System (INIS)

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

1993-01-01

effect on irradiation response was noted and the HSST-02 material's response to irradiation was similar to results from power reactor and other test reactor experiments, thus qualifying the Ford Test Reactor for irradiation experiments such as those conducted for the Yankee Atomic program. (author)

Effect of γ-ray irradiation on properties of castor oil-polyurethane potting materials

International Nuclear Information System (INIS)

Guan Jian; Luo Xianglin; Yue Yilun

2001-01-01

After γ-ray sterilization, the amounts of 4,4'-methylenedianiline (MDA) in the five kinds of synthesized medical castor oil-polyurethane potting materials were detected by HPLC. The influences of γ-ray irradiation on the mechanical performance of the potting materials were also discussed quantitatively.The experimental results show that the amounts of produced MDA increases with γ-ray irradiation dosage. After 25 kGy γ-ray sterilization, the accumulated amounts of MDA in the five kinds of potting materials were 10.33, 10.37, 10.52, 10.59, 10.91 ? μg/g respectively. Those amounts are below the level of harm amount to human body. At the same time, the mechanical properties of the potting materials such as tensile strength, tear strength and hardness are improved because cross-linking happens under irradiation

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.)

TEM examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

Srivastava, D.; Tewari, R.; Dey, G.K.; Sharma, B.P.; Sah, D.N.; Banerjee, Suparna; Sahoo, K.C.

2005-09-01

In the present work, microstructure of the zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water RAPP-1. Reactor (PHWR) has been examined for the first time using transmission electron microscope (TEM). The samples were obtained 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) and expected to have a cumulative radiation damage of about 3 dpa. In this study irradiated microstructure has been characterized and compared it with the microstructure of the unirradiated pressure tube samples. The effect of irradiation on the hydriding behaviour is also studied. (author)

Fracture toughness and strength change of neutron-irradiated ceramic materials

International Nuclear Information System (INIS)

Dienst, W.; Zimmermann, H.

1994-01-01

In order to analyse the results of bending strength measurements on neutron-irradiated samples of Al 2 O 3 , AlN and SiC, fracture toughness measurements were additionally conducted. The neutron fluences concerned were mostly in the range of 0.6 to 3.2x10 26 n/m 2 at irradiation temperatures of 400 to 550 C. A fracture toughness decrease was generally observed for polycrystalline materials which, however, was considerably smaller than the reduction of the fracture strength. Exceptional increase of the fracture toughness seems typical for the effect of rather coarse irradiation defects. The irradiation-induced change of the fracture toughness of single crystal Al 2 O 3 appeared dependent on the crystallographic orientation; both reduced and increased fracture toughness after irradiation was observed. Recent results of neutron irradiation to about 2x10 25 n/m 2 at 100 C showed, that the strength decrease of various Al 2 O 3 grades sets in at (3-5)x10 24 n/m 2 and seems to be little dependent on the irradiation temperature. ((orig.))

Fusion Materials Irradiation Test Facility

International Nuclear Information System (INIS)

Kemp, E.L.; Trego, A.L.

1979-01-01

A Fusion Materials Irradiation Test Facility is being designed to be constructed at Hanford, Washington, The system is designed to produce about 10 15 n/cm-s in a volume of approx. 10 cc and 10 14 n/cm-s in a volume of 500 cc. The lithium and target systems are being developed and designed by HEDL while the 35-MeV, 100-mA cw accelerator is being designed by LASL. The accelerator components will be fabricated by US industry. The total estimated cost of the FMIT is $105 million. The facility is scheduled to begin operation in September 1984

Status of the material capsule irradiation and the development of the new capsule technology in HANARO

International Nuclear Information System (INIS)

Choo, Kee-Nam; Kang, Young-Hwan; Choi, Myoung-Hwan; Cho, Man-Soon; Kim, Bong-Goo

2006-01-01

A material capsule system including a main capsule, fixing, control, cutting, and transport systems was developed for an irradiation test of non-fissile materials in HANARO. 14 irradiation capsules (12 instrumented and 2 non-instrumented capsules) have been designed, fabricated and successfully irradiated in the HANARO CT and IR test holes since 1995. The capsules were mainly designed for an irradiation of the RPV (Reactor Pressure Vessel), reactor core materials, and Zr-based alloys. Most capsules were made for KAERI material research projects, but 5 capsules were made as a part of national projects for the promotion of the HANARO utilization for universities. Based on the accumulated irradiation experience and the user's sophisticated requirements, development of new instrumented capsule technologies for a more precise control of the irradiation temperature and fluence of a specimen irrespective of the reactor operation has been performed in HANARO. (author)

The effect of neutron irradiation on the trapping of tritium in carbon-based materials

International Nuclear Information System (INIS)

Kwast, H.; Werle, H.; Glugla, M.; Wu, C.H.; Federici, G.

1993-11-01

Carbon-based materials are considered for protection of plasma facing components in the next step fusion device. To investigate the effects of neutron damage on the tritium behaviour an experimental study on the tritium retention of various neutron irradiated graphites and carbon/carbon fibre composites was started. The irradiation dose of the specimens ranges from 10 -3 to 3.5 dpa.g and the irradiation temperature from 390 C to 1500 C. A comparison of tritium retention in pre- and post-irradiated carbon-based materials as a function of the sample temperature is reported in this paper and the results are discussed. The first results indicate that the retention of tritium is higher in irradiated graphite than in unirradiated graphite and depends largely on the density and microstructure. The retention is also influenced by the tritium-loading temperature. Graphite of type S 1611, irradiated at 400 C and 600 C up to a damage of 0.1 dpa.g, retained about two times more tritium than the unirradiated material. (orig.)

Analysis of the irradiation data for A302B and A533B correlation monitor materials

International Nuclear Information System (INIS)

Wang, J.A.

1996-04-01

The results of Charpy V-notch impact tests for A302B and A533B-1 Correlation Monitor Materials (CMM) listed in the surveillance power reactor data base (PR-EDB) and material test reactor data base (TR-EDB) are analyzed. The shift of the transition temperature at 30 ft-lb (T 30 ) is considered as the primary measure of radiation embrittlement in this report. The hyperbolic tangent fitting model and uncertainty of the fitting parameters for Charpy impact tests are presented in this report. For the surveillance CMM data, the transition temperature shifts at 30 ft-lb (ΔT 30 ) generally follow the predictions provided by Revision 2 of Regulatory Guide 1.99 (R.G. 1.99). Difference in capsule temperatures is a likely explanation for large deviations from R.G. 1.99 predictions. Deviations from the R.G. 1.99 predictions are correlated to similar deviations for the accompanying materials in the same capsules, but large random fluctuations prevent precise quantitative determination. Significant scatter is noted in the surveillance data, some of which may be attributed to variations from one specimen set to another, or inherent in Charpy V-notch testing. The major contributions to the uncertainty of the R.G. 1.99 prediction model, and the overall data scatter are from mechanical test results, chemical analysis, irradiation environments, fluence evaluation, and inhomogeneous material properties. Thus in order to improve the prediction model, control of the above-mentioned error sources needs to be improved. In general the embrittlement behavior of both the A302B and A533B-1 plate materials is similar. There is evidence for a fluence-rate effect in the CMM data irradiated in test reactors; thus its implication on power reactor surveillance programs deserves special attention

Simulation of the welding of irradiated materials

International Nuclear Information System (INIS)

Lin, Hua Tay

1989-07-01

Helium was uniformly implanted using the ''tritium trick'' technique to levels of 0.18, 2.5, 27, 105 and 256 atomic part per million (appm) for type 316 stainless steel, and 0.3 and 1 appm for Sandvik HT-9 (12 Cr-1MoVW). Both full penetration as well as partial penetration welds were then produced on control and helium-containing materials using the autogenous gas tungsten arc (GTA) welding process under full constraint conditions. For full penetration welds, both materials were successfully welded when they contained less than 0.3 appm helium. However, welds of both materials, when containing greater than 1 appm helium, were found to develop cracks during cooling of the weld. Transmission and scanning electron microscopy indicated that the HAZ cracking was caused by the growth and coalescence of grain boundary (GB) helium bubbles. This cracking occurred as a result of the combination of high temperatures and high shrinkage tensile stresses. The cracking in the fusion zone was found to result from the precipitation of helium along dendrite interfaces. A model based on the kinetics of diffusive cavity growth is presented to explain the observed results. The model proposes a helium bubble growth mechanism which leads to final intergranular rupture in the heat-affected zone. Results of the present study demonstrate that the use of conventional fusion welding techniques to repair materials degraded by exposure to irradiation environments may be difficult if the irradiation results in the generation of helium equal to or greater than 1 appm

Calculations on neutron irradiation damage in reactor materials

International Nuclear Information System (INIS)

Sone, Kazuho; Shiraishi, Kensuke

1976-01-01

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

Analysis of the radiolytic products on high-dose irradiated food and packing materials

International Nuclear Information System (INIS)

Kim, Kyong Su; Shim, Sung Lye; Chung, In Sun

2010-04-01

The aims of this study were to prepare the government approval for the extension of food irradiation item to food or its products, to promote the industrial application of radiation technology, and to apply basic data in policy for introduction of irradiation. The change of hydrocarbons by irradiation was evaluated for the detection of irradiated meat. The results showed that hydrocarbons were detected in all of irradiated samples, but these hydrocarbons were not detected in non-irradiated samples. There were no difference between vacuum and N 2 - packaging. According to fatty acid compounds and degradation pathway of beef and pork, it could be deliberated that a great amount of produced hydrocarbons such as 8-heptadenene and 1,7-hexadecadien were able to be used as identification factor of irradiated meat. Effects of γ-irradiation on the volatile organic compounds in agricultural products were determined by analyzing changes of volatile composition. The composition of volatile organic compounds were little changed, but few specific compounds induced by γ-irradiation were identified. The variations of concentration in irradiated samples identified in this study could be due to the radiation sensitivity of compounds with the dose used. Effects of γ-irradiation on the volatile compounds in packaging materials were determined by analyzing changes of volatile composition. In polyethylene and polypropylene, 1,3-DBB was identified only in irradiated samples. Levels of 1,3-DBB increased with increasing irradiation doses. These results suggest may be useful in evaluation of γ-irradiation effects on food packaging materials

Analysis of the radiolytic products on high-dose irradiated food and packing materials

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyong Su; Shim, Sung Lye; Chung, In Sun [Chosun University, Gwangju (Korea, Republic of)

2010-04-15

The aims of this study were to prepare the government approval for the extension of food irradiation item to food or its products, to promote the industrial application of radiation technology, and to apply basic data in policy for introduction of irradiation. The change of hydrocarbons by irradiation was evaluated for the detection of irradiated meat. The results showed that hydrocarbons were detected in all of irradiated samples, but these hydrocarbons were not detected in non-irradiated samples. There were no difference between vacuum and N{sub 2}- packaging. According to fatty acid compounds and degradation pathway of beef and pork, it could be deliberated that a great amount of produced hydrocarbons such as 8-heptadenene and 1,7-hexadecadien were able to be used as identification factor of irradiated meat. Effects of {gamma}-irradiation on the volatile organic compounds in agricultural products were determined by analyzing changes of volatile composition. The composition of volatile organic compounds were little changed, but few specific compounds induced by {gamma}-irradiation were identified. The variations of concentration in irradiated samples identified in this study could be due to the radiation sensitivity of compounds with the dose used. Effects of {gamma}-irradiation on the volatile compounds in packaging materials were determined by analyzing changes of volatile composition. In polyethylene and polypropylene, 1,3-DBB was identified only in irradiated samples. Levels of 1,3-DBB increased with increasing irradiation doses. These results suggest may be useful in evaluation of {gamma}-irradiation effects on food packaging materials

Inorganic-organic hybrid polymer for preparation of affiliating material using electron beam irradiation

International Nuclear Information System (INIS)

Chung, Jaeseung; Kim, Seongeun; Kim, Byounggak; Lee, Jongchan; Park, Jihyun; Lee, Byeongcheol

2011-01-01

Recently, silver nano materials have gained a lot of attentions in a variety of applications due to the unique biological, optical, and electrical properties. Especially, the antifouling property of these material is considered to be an important character for biomedical field, marine coatings industry, biosensor, and drug delivery. In this study, we design and synthesize the inorganic-organic hybrid polymer for preparation of affiliating materials. Silver nano materials having antifouling property with different shapes are prepared by control the electron beam irradiation conditions. Inorganic-organic hybrid polymer was synthesized and characterized. → Morphology and size controlled nano materials are prepared using electron beam irradiation. → Silver nano materials having various shapes can be used for antifouling material

Newly developed non-destructive testing method for evaluation of irradiation brittleness of structural materials using ultrasonic

International Nuclear Information System (INIS)

Ishii, Toshimitsu; Ooka, Norikazu; Kato, Yoshiaki; Saito, Junichi; Hoshiya, Taiji; Shibata, Saburo; Kobayashi, Hideo

1999-01-01

Surveillance testing is important to evaluate neutron irradiation embrittlement of reactor pressure vessel material for long life operation. An alternative test method for evaluating the irradiation embrittlement of the pressure vessel material will have to be proposed to support the limited number of surveillance test specimens in order to manage the plant life to be extended. In this study, ultrasonic testing for irradiated A533B-1 steel and weld metal was applied to examine material degradation nondestructively. With increasing the shift of Charpy 41 J transition temperature, ultrasonic velocity decreased and attenuation coefficient of ultrasonic wave increased. Especially, the difference of ultrasonic velocity for 5 MHz shear wave between as-received and irradiated material is corresponding to the shift of transition temperature showing material degradation. (author)

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

International Nuclear Information System (INIS)

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

1986-06-01

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

Neutron and gamma irradiation effects on organic insulating materials for fusion magnets

International Nuclear Information System (INIS)

Maurer, W.

1985-10-01

Available low-temperature neutron and gamma irradiation data for organic insulating materials are collected and compared with room temperature data. Only the most promising polymers in terms of mechanical strength for magnet insulation are taken into account. For characterization and comparison of different materials the 75% dose is used, i.e. the dose, where the mechanical strength is reduced by 25%, and 75% is retained. For room temperature special prepared polyimide and epoxy materials reinforced with glass fibre retained 75% of the mechanical strength up to a dose of 7x10 7 Gy. For 5 K irradiation the best epoxy material retained the 75% dose up to 1x10 7 Gy, the best polyimide material up to 1x10 8 Gy. (orig.) [de

The irradiation induced creep in fuel compact materials for H.T.R. applications

International Nuclear Information System (INIS)

Veringa, H.; Blackstone, R.; Loelgen, R.

1976-01-01

Restrained shrinkage experiments up to 3 x 10 21 ncm -2 (DNE) in the temperature range of 600-1,200 0 C on three different dummy coated particle fuel compact materials were performed in the High Flux Reactor at Petten, the Netherlands. The data were evaluated to obtain the steady state irradiation creep coefficient of the compacts. It was found that for the materials investigated, the creep coefficient is temperature-dependent, but no clear relationship to the Young's modulus could be established. Under certain conditions, this irradiation-induced plasticity influences the elastic properties, while also the creep coefficient increases. This effect coincides with the formation and further opening of cracks due to stresses caused by irradiation shrinkage of the matrix material. (orig.) [de

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

Irradiation creep lifetime analysis on first wall structure materials for CFETR

Energy Technology Data Exchange (ETDEWEB)

Ma, Bing; Peng, Lei, E-mail: penglei@ustc.edu.cn; Zhang, Xiansheng; Shi, Jingyi; Zhan, Jie

2017-05-15

Fusion reactor first wall services on the conditions of high surface heat flux and intense neutron irradiation. For China Fusion Engineering Test Reactor (CFETR) with high duty time factor, it is important to analyze the irradiation effect on the creep lifetime of the main candidate structure materials for first wall, i.e. ferritic/martensitic steel, austenite steel and oxide dispersion strengthened steel. The allowable irradiation creep lifetime was evaluated with Larson-Miller Parameter (LMP) model and finite element method. The results show that the allowable irradiation creep lifetime decreases with increasing of surface heat flux, first wall thickness and inlet coolant temperature. For the current CFETR conceptual design, the lifetime is not limited by thermal creep or irradiation creep, which indicated the room for design parameters optimization.

Optical and electrical phenomena in dielectric materials under irradiation

CERN Document Server

Plaksin, O A; Stepanov, P A; Demenkov, P V; Chernov, V M; Krutskikh, A O

2002-01-01

Optical and acoustic properties of the materials based on Al sub 2 O sub 3 , SiO sub 2 and BN under 8 MeV proton irradiation (<10 sup 4 Gy/s) have been measured. Electric charge partitioning has been shown to result in charging the microscopic regions in the bulk of the dielectrics under irradiation, which is due to different mobility of free electrons and holes (sapphire), concentration inhomogeneity in the system of charge carrier traps (alumina), or thermodynamic instability of the homogeneous distribution of the filled traps (silica glasses). Prevalent charge carrier recombination in the grain boundaries causes re-crystallization of pyrolytic boron nitride under irradiation, which shows up as simultaneous decrease of the intensity of radiation-induced luminescence (RIL) of the centres in the grain boundaries and the BN. The local charging results in optical inhomogeneity of the silica glasses which is sustained by the optical loss spectra of the irradiated glasses, features of kinetics of bleaching, RI...

Chemical and physical change of packaging materials for food by gamma-ray irradiation

International Nuclear Information System (INIS)

Kawamura, Yoko; Yamada, Takashi

1999-01-01

Recently, foods are often exposed to radiation under packed states with various wrapping materials. In this study, the effects of γ-ray radiation were investigated on the additives in wrapping materials on the market. 10 - 50 kGy γ radiation was irradiated to samples under sealed condition in a glass-ware. Polyethylene bag and wrapping film, polypropylene wrapping film, cup and sheet, and polystyrene cup were used as samples. And the additives in these materials were analyzed by GC/MS to evaluate the radiation effects on them. The irradiation was found to induce rapid degradation of antioxidants, especially, Irgafos 168. Some fatty acid amides used as a lubricant and a plasticizer, DBP were also reduced, but not aliphatic hydrocarbons. However, all polystyrene products used in this study included no additives. The contents of styrene dimer and trimer in those wrapping materials were not changed by γ-irradiation. (M.N.)

Review Paper: Review of Instrumentation for Irradiation Testing of Nuclear Fuels and Materials

International Nuclear Information System (INIS)

Kim, Bong Goo; Rempe, Joy L.; Villard, Jean-Francois; Solstadd, Steinar

2011-01-01

Over 50 years of nuclear fuels and materials irradiation testing has led to many countries developing significant improvements in instrumentation to monitor physical parameters and to control the test conditions in material test reactors (MTRs). Recently, there is increased interest to irradiate new materials and reactor fuels for advanced pressurized water reactors and Gen-IV reactor systems, such as sodium-cooled fast reactors, very high temperature reactors, supercritical water-cooled reactors, and gas-cooled fast reactors. This review paper documents the current state of instrumentation technologies in MTRs in the world and summarizes ongoing research efforts to deploy new sensors. As described in this paper, a wide range of sensors is available to measure key parameters of interest during fuels and materials irradiations in MTRs. Ongoing development efforts focus on providing MTR users a wider range of parameter measurements with smaller, higher accuracy sensors.

Microstructural processes in irradiated materials

Science.gov (United States)

Byun, Thak Sang; Kaoumi, Djamel; Bai, Xian-Ming

2017-12-01

The 8th symposium on Microstructural Progresses in Irradiated Materials (MPIM) was held at San Diego Convention Center and Marriott Marquis & Marina, San Diego, California, USA, February 26-March 2, 2017, as part of the TMS 2017 146th Annual Meeting and Exhibition. Since 2003, when the first MPIM symposium was held in the same place, the symposium has been held in odd years and has grown to one of the biggest symposia in the TMS Annual Meeting which invites more than sixty symposia. In the 8th MPIM symposium, a total of 106 oral and poster presentations, including 16 invited talks, were delivered for 4 days.

Effect of packaging material on nitrate nitrogen content of irradiated potatoes

International Nuclear Information System (INIS)

Mondy, N.I.; Koushik, S.R.

1990-01-01

The effect of packaging materials on nitrate nitrogen content of irradiated potatoes was investigated. Tubers were irradiated at 10, 30 and 100 Krads and stored for 12 wk at 5 degrees C in paper or plastic bags. Nitrate nitrogen content was significantly (p 0.01) higher in tubers packaged in plastic as compared to those in paper bags. Irradiation significantly (p 0.01) increased nitrate nitrogen content between the lowest and highest levels of treatment in tubers stored in both paper and plastic bags

A correlation between micro- and nano-indentation on materials irradiated by high-energy heavy ions

Science.gov (United States)

Yang, Yitao; Zhang, Chonghong; Ding, Zhaonan; Su, Changhao; Yan, Tingxing; Song, Yin; Cheng, Yuguang

2018-01-01

Hardness testing is an efficient means of assessing the mechanical properties of materials due to the small sampling volume requirement. Previous studies have established the correlation between flow stress and Vickers hardness. However, the damage layer produced by ions irradiation with low energy is too thin to perform Vickers hardness test, which is usually measured by nano-indentation. Therefore, it is necessary to correlate the Vickers hardness and nano-hardness for the convenience of assessing mechanical properties of materials under irradiation. In this study, various materials (pure nickel, nickel base alloys and oxide dispersion strengthened steel) were irradiated with high-energy heavy ions to different damage levels. After irradiation, micro- and nano-indentation were performed to characterize the change in hardness. Due to indentation size effect (ISE), the hardness was dependent of load or depth. Therefore, Nix-Gao model was used to obtain the hardness without ISE (Hv0 and Hnano_0). The determined Hv0 was plotted as a function of the corresponding Hnano_0, then a good linear relation was found between Vickers hardness and nano-hardness, and a coefficient was determined to be 81.0 ± 10.5, namely, Hv 0 = 81.0Hnano _ 0 (Hv0 with unit of kgf/mm2, Hnano_0 with unit of GPa). This correlation was based on the data from various materials, therefore it was independent of materials. Based on the established correlation and nano-indentation results, the change fraction in yield stress of Inconel 718 and pure Ni with ion irradiation was compared with that with neutron irradiation. The data of Inconel 718 with heavy ion irradiation was in good agreement with the data with neutron irradiation, which was a good demonstration for the validation of the established correlation. However, a distinctive difference in change fraction of yield stress was seen for pure Ni under heavy ion irradiation and neutron irradiation, which was attributed to the difference in samples

Irradiation as an alternative environmentally friendly method for microbiological decontamination of herbal raw material

International Nuclear Information System (INIS)

Dragusin, M.; Rotaru, R.

2000-01-01

Microbiological contamination of herbal raw materials is a serious problem in the production of therapeutical preparations. A good quality of the product, according to the pharmaceutical requirements may be achieved by applying suitable methods of decontamination. The decontamination treatments should be fast and effective against all microorganisms. It should ensure the decontamination of both packaging and the microorganisms present and must not reduce the sensory and technological qualities of the commodities. Decontamination of herbal raw materials by irradiation is a method by choice. It is because chemical methods are recognized recently as not safe to the consumer. Irradiation, in turn, is technically feasible, very effective and friendly enough to environment process. Under the prevailing production and handling conditions, most herbs contain a large number of microorganisms what is a serious problem in the production of therapeutical preparations. For several years the most widely used methods for decontamination of herbs was fumigation with ethylene oxide or methyl bromide. Both methods today banned in most countries. Irradiation is an alternative and safe method for effective reducing the microbial contamination of herbal raw materials. The following raw materials have been examined: Folium Cynara, Folium Plantago, Flos Chamomillae, Semen Sylibum Marianum and Folium Farfara. The content of biologically active compounds before and after irradiation of the raw materials did not change in a significant degree after irradiation. The dose of radiation for herbals raw materials was 10 kGy. There are two groups of raw materials: - The raw materials designed for preparing granulates, tablets, dragees, capsules, aqueous extracts, infusions, macerations and preparations for external use; - The raw materials assigned for preparing alcoholic preparations, isolated compounds, oil preparations and essential oils. The medical herbs and herbal raw materials before their

Irradiation effects on C/C composite materials for high temperature nuclear applications

International Nuclear Information System (INIS)

Eto, M.; Ugachi, H.; Baba, S.I.; Ishiyama, S.; Ishihara, M.; Hayashi, K.

2000-01-01

Excellent characteristics such as high strength and high thermal shock resistance of C/C composite materials have led us to try to apply them to the high temperature components in nuclear facilities. Such components include the armour tile of the first wall and divertor of fusion reactor and the elements of control rod for the use in HTGR. One of the most important aspects to be clarified about C/C composites for nuclear applications is the effect of neutron irradiation on their properties. At the Japan Atomic Energy Research Institute (JAERI), research on the irradiation effects on various properties of C/C composite materials has been carried out using fission reactors (JRR-3, JMTR), accelerators (TANDEM, TIARA) and the Fusion Neutronics Source (FNS). Additionally, strength tests of some neutron-irradiated elements for the control rod were carried out to investigate the feasibility of C/C composites. The paper summarises the R and D activities on the irradiation effects on C/C composites. (authors)

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

International Nuclear Information System (INIS)

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

2008-06-01

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

Irradiation tests on bitumen and bitumen coated materials

International Nuclear Information System (INIS)

Tabardel-Brian, R.; Rodier, J.; Lefillatre, G.

1969-01-01

The use of bitumen as a material for coating high-activity products calls for prior study of the resistance of bitumen to irradiation. After giving briefly the methods of preparation of bitumen- coated products, this report lists the equipment which has been used for carrying out the β and γ irradiations of these products, and gives the analytical results obtained as a function of the dose rates chosen and of the total integrated dose. Finally, some conclusions have been drawn concerning the best types of bitumen. It should be stressed that some bitumens apparently underwent no degradation whatsoever nor any volume increase, for a total integrated dose of 1.8 x 10 10 rads. (authors) [fr

IFMIF, a fusion relevant neutron source for material irradiation current status

International Nuclear Information System (INIS)

Knaster, J.; Chel, S.; Fischer, U.; Groeschel, F.; Heidinger, R.; Ibarra, A.; Micciche, G.; Möslang, A.; Sugimoto, M.; Wakai, E.

2014-01-01

The d-Li based International Fusion Materials Irradiation Facility (IFMIF) will provide a high neutron intensity neutron source with a suitable neutron spectrum to fulfil the requirements for testing and qualifying fusion materials under fusion reactor relevant irradiation conditions. The IFMIF project, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the Broader Approach (BA) Agreement between Japan Government and EURATOM, aims at the construction and testing of the most challenging facility sub-systems, such as the first accelerator stage, the Li target and loop, and irradiation test modules, as well as the design of the entire facility, thus to be ready for the IFMIF construction with a clear understanding of schedule and cost at the termination of the BA mid-2017. The paper reviews the IFMIF facility and its principles, and reports on the status of the EVEDA activities and achievements

Stock selection of high-dose-irradiation-resistant materials for filter press under high-dose irradiation operation

International Nuclear Information System (INIS)

Ishiyama, Shintaro; Minami, Mamoru; Hara, Kouji; Yamashita, Manabu

2015-01-01

In a volume reduction process for the decontamination of contained soil, the performance degradation of a filter press is expected owing to material deterioration under high-dose irradiation. Eleven-stock selection of candidate materials including polymers, fibers and rubbers for the filter press was conducted to achieve a high performance of volume reduction of contaminated soil and the following results were derived. Crude rubber and nylon were selected as prime candidates for packing, diaphragm and filter plate materials. Polyethylene was also selected as a prime candidate for the filter cloth material. (author)

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

Development of Environment and Irradiation Effects of High Temperature Materials

International Nuclear Information System (INIS)

Ryu, Woo Seog; Kim, D. W.; Kim, S. H.

2009-11-01

Proposed materials, Mod.9Cr-1Mo steel (32 mm thickness) and 9Cr-1Mo-1W (100 mm thickness), for the reactor vessel were procured, and welded by the qualified welding technologies. Welding soundness was conformed by NDT, and mechanical testings were done along to weld depth. Two new irradiation capsules for use in the OR test hole of HANARO were designed and fabricated. specimens was irradiated in the OR5 test hole of HANARO with a 30MW thermal power at 390±10 .deg. C up to a fast neutron fluence of 4.4x10 19 (n/cm 2 ) (E>1.0 MeV). The dpa was evaluated to be 0.034∼0.07. Base metals and weldments of both Mod.9Cr-1Mo and 9Cr-1Mo-1W steels were tested tensile and impact properties in order to evaluate the irradiation hardening effects due to neutron irradiation. DBTT of base metal and weldment of Mod.9Cr-1Mo steel were -16 .deg. C and 1 .deg. C, respectively. After neutron irradiation, DBTT of weldment of Mod.9Cr-1Mo steel increased to 25 . deg. C. Alloy 617 and several nickel-base superalloys were studied to evaluate high temperature degradation mechanisms. Helium loop was developed to evaluate the oxidation behaviors of materials in the VHTR environments. In addition, creep behaviors in air and He environments were compared, and oxidation layers formed outer surfaces were measured as a function of applied stress and these results were investigated to the creep life

Data on post irradiation experiments of heat resistant ceramic composite materials. PIE for 97M-13A

Energy Technology Data Exchange (ETDEWEB)

Baba, Shin-ichi; Ishihara, Masahiro; Souzawa, Shizuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Sekino, Hajime [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

The research on the radiation damage mechanism of heat resistant ceramic composite materials is one of the research subjects of the innovative basic research in the field of high temperature engineering, using the High Temperature engineering Test Reactor (HTTR). Three series of irradiation tests on the heat resistant ceramic composite materials, first to third irradiation test program, were carried out using the Japan Material Testing Reactor (JMTR). This is a summary report on the first irradiation test program; irradiation induced dimensional change, thermal expansion coefficient, X-ray diffraction and {gamma}-ray spectrum are reported. (author)

Effect of packaging materials on the quality of irradiated ground spices

International Nuclear Information System (INIS)

Saputra, T.S.; Maha, Munsiah; Purwanto, Z.I.

1985-01-01

These experiments were carried out to determine the suitable packaging materials to be used for irradiated ground spices produced in Indonesia. The materials used were white pepper (Piper album), black pepper (Piper nigrum) nutmeg (Myristica fragrans), turmeric (Curcuma domestica), and ginger (Zangiber officinale R.) packaged in transparent polypropylene bottles, in pouches made of cellophane-aluminum foil and lithopaper-polyethylene laminates. The samples were irradiated at 5 kGy, stored at ambient conditions, and then examined every 3 months from 0 up to 9 months of storage. The parameters observed were total bacterial counts, total moulds and yeast counts, water activity (Aw), moisture content, and organoleptic scores of the samples. Piperine content of white pepper and black pepper, colour of turmeric extract, and rancidity of ginger were also determined. The results showed that the packaging materials used had no significant effect on bacterial load of the samples. Prolonged storage, however, could reduce the microbial load of the ground spices. Irradiation at 5 kGy could effectively increase the hygienic condition as well as storage life of the ground spices under investigation without affecting their organoleptic properties. (author). 8 refs

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.

Investigation of high flux test module for the international fusion materials irradiation facilities (IFMIF)

International Nuclear Information System (INIS)

Miyashita, Makoto; Sugimoto, Masayoshi; Yutani, Toshiaki

2007-03-01

This report describes investigation on structure of a high neutron flux test module (HFTM) for the International Fusion Materials Irradiation Facilities (IFMIF). The HFTM is aimed for neutron irradiation of a specimen in a high neutron flux domain of the test cell for irradiation ground of IFMIF. We investigated the overall structure of the HFTM that was able to include specimens in a rig and thermocouple arrangement, an interface of control signal and support structure. Moreover, pressure and the amount of the bend in the module vessel (a rectangular section pressure vessel) were calculated. The module vessel did a rectangular section from limitation of a high neutron flux domain. Also, we investigated damage of thermocouples under neutron irradiation, which was a temperature sensor of irradiation materials temperature control demanded high precision. Based on these results, drawings on the HTFM structure. (author)

Semiconductor-diode-aided dosimetry of the irradiation of pourable bulk material

International Nuclear Information System (INIS)

Gruenewald, T.; Rudolf, M.

1987-01-01

The irradiation of unpackaged pourable bulk material requires the employment of a dosimeter which can be readily transported along with the material. Planar diffused silicon diodes have been found to be suitable for this purpose. To date these have been used solely for the purpose of dose rate measurements; however, it can be shown that the permanent change in reverse recover time at the p-n junction correlates with the absorbed irradiation dose in the range up to 10 kGy. Appropriate selection of the diode and thermal treatment lead to a linear dependence and enable the silicon dosimeter to be reused. (author). 16 refs, 4 figs

Irradiation-accelerated corrosion of reactor core materials

International Nuclear Information System (INIS)

Bartels, David; Was, Gary; Jiao, Zhijie

2012-09-01

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

Surface modification of ceramic materials induced by irradiation of high power pulsed ICP

International Nuclear Information System (INIS)

Ishigaki, Takamasa; Okada, Nobuhiro; Ohashi, Naoki; Haneda, Hajime

2003-01-01

Newly developed pulse-modulated high-power inductively coupled plasma [ICP] is expected to offer the unique physico-chemical condition, such as the increased concentration of chemically reactive species, as well as the appropriate heat flux for materials processing. Two kinds of oxide materials, titanium and zinc oxide, were placed at the downstream of Ar-H 2 ICP and irradiated in the plasma of continuous [CN] and pulse-modulated [PM] modes. The CN-ICP irradiation at the position close to the plasma tail gave rise to the thermal reduction of oxides. In the PM-ICP irradiation, the degree of thermal reduction depended on the lower power level during pulse-off time, as well as the total electric power. Irradiation in PM-ICP led to the increased formation of oxygen vacancies in titanium dioxide. In the case of zinc oxide, the UV emission efficiency was improved by PM-ICP irradiation, while the green emission became predominant by CN-ICP irradiation at the appropriate position. Induced effects in the two oxides by PM-ICP would be related to the high concentration of hydrogen radicals in the plasma. (author)

Study on dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment

International Nuclear Information System (INIS)

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

2001-01-01

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 using fission reactors. 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. The irradiation capsules for in-situ measurement and varying temperature were developed successfully. It was found that insulating ceramics were worked up to 3 dpa. The property changes and related issues in low activation structural materials were summarized. (author)

Project accent: graphite irradiated creep in a materials test reactor

International Nuclear Information System (INIS)

Brooking, M.

2014-01-01

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

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)

Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures

International Nuclear Information System (INIS)

Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.; Shiba, Kiyoyuki

1994-01-01

Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 or 250 degrees C. These specimens have been tested over a temperature range from 20 to 250 degrees C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250 degrees C is more damaging than at 90 degrees C, causing larger decreases in the fracture toughness. Ferritic-martensitic steels are embrittled by the irradiation, and show the lowest toughness at room temperature

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-08-31

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

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.

Irradiation effects on material properties of steels used in nuclear reactors: a literature review

International Nuclear Information System (INIS)

Gerceker, N.; Dara, I. H.

2001-01-01

The structural materials of a nuclear power plant are of vital importance as they provide mechanical strength, structural support and physical containment for the primary reactor components as well as the nuclear power plant itself. These structural materials comprise mainly of metals and their alloys, ceramics and cermets. However, metals and their alloys are the most widely used materials and the irradiation effects are more pronounced on metallic materials as of their high temperature properties are more sensitive (with respect to ceramics and cermets) to any kind of external effects. The wholesale creation of effects on material properties has been studied for over four decades and it is not realistic to attempt to represent even a small part of the field in single poster paper. In the present contribution, a literature review of the irradiation effects on the material properties of different types of steel alloys will be given because steels are widely used as structural materials in reactors and therefore the irradiation effects on steels may be of paramount importance for reactor design, operation and safety concepts which will be discussed about radiation effects on material properties of steels will provide highlights to better understanding of the origins and development of radiation effects in materials

Effect of thermal annealing on property changes of neutron-irradiated non-graphitized carbon materials and nuclear graphite

International Nuclear Information System (INIS)

Matsuo, Hideto

1991-06-01

Changes in dimension of non-graphitized carbon materials and nuclear graphite, and the bulk density, electrical resistivity, Young's modulus and thermal expansivity of nuclear graphite were studied after neutron irradiation at 1128-1483 K and the successive thermal annealing up to 2573 K. Carbon materials showed larger and anisotropic dimensional shrinkage than that of nuclear graphite after the irradiation. The irradiation-induced dimensional shrinkage of carbon materials decreased during annealing at temperatures from 1773 to 2023 K, followed by a slight increase at higher temperatures. On the other hand, the irradiated nuclear graphite hardly showed the changes in length, density and thermal expansivity under the thermal annealing, but the electrical resistivity and Young's modulus showed a gradual decrease with annealing temperature. It has been clarified that there exists significant difference in the effect of thermal annealing on irradiation-induced dimensional shrinkage between graphitized nuclear graphite and non-graphitized carbon materials. (author)

Gamma irradiation induced effects of butyl rubber based damping material

Science.gov (United States)

Chen, Hong-Bing; Wang, Pu-Cheng; Liu, Bo; Zhang, Feng-Shun; Ao, Yin-Yong

2018-04-01

The effects of gamma irradiation on the butyl rubber based damping material (BRP) at various doses in nitrogen were investigated in this study. The results show that irradiation leads to radiolysis of BRP, with extractives increasing from 14.9 ± 0.8% of control to 37.2 ± 1.2% of sample irradiated at 350 kGy, while the swelling ratio increasing from 294 ± 3% to 766 ± 4%. The further investigation of the extractives with FTIR shows that the newly generated extractives are organic compounds containing C-H and C˭C bonds, with molecular weight ranging from 26,500 to 46,300. SEM characterization shows smoother surface with holes disappearing with increasing absorbed doses, consistent with "softer" material because of radiolysis. Dynamic mechanical study of BRP show that tan δ first slightly then obviously increases with increasing absorbed dose, while storage modulus slightly decreases. The tensile testing shows that the tensile strength decreases while the elongation at break increases with increasing dose. The positron annihilation lifetime spectroscopy show no obvious relations between free volume parameters and the damping properties, indicating the complicated influencing factors of damping properties.

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

Electron irradiation experiments in support of fusion materials development

International Nuclear Information System (INIS)

Gelles, D.S.; Ohnuki, S.; Takahashi, H.; Matsui, H.; Kohno, Y.

1991-11-01

Microstructural evolution in response to 1 MeV irradiation has been investigated for three simple ferritic alloys, pure beryllium, pure vanadium, and two simple vanadium alloys over a range of temperatures and doses. Microstructural evolution in Fe-3, -9, and -18Cr ferritic alloys is found to consist of crenulated, faulted a loops and circular, unfaulted a/2 loops at low temperatures, but with only unfaulted loops at high temperatures. The complex dislocation evolution is attributed to sigma phase precipifaults arising from chromium segregation to point defect sinks. Beryllium is found to be resistant to electron damage; the only effect observed was enhanced dislocation mobility. Pure vanadium, V-5Fe, and V-1Ni microstructural response was complicated by precipitation on heating to 400 degrees C and above, but dislocation evolution was investigated in the range of room temperature to 300 degrees C and at 600 degrees C. The three materials behaved similarly, except that pure vanadium showed more rapid dislocation evolution. This difference does not explain the enhanced swelling observed in vanadium alloys

Change in properties of superconducting magnet materials by fusion neutron irradiation

International Nuclear Information System (INIS)

Nishimura, Arata; Nishijima, Shigehiro; Takeuchi, Takao; Nishitani, Takeo

2007-01-01

A fusion reactor will generate a lot of high energy neutron and much energy will be taken out of the neutrons by a blanket system. Since some neutrons will stream out of a plasma vacuum vessel through neutral beam injection ports and penetrate a blanket system, a superconducting magnet system, which provides high magnetic field to confirm high energy particles, will be irradiated by a certain amount of neutrons. By developing the new NBI system or by reducing the penetration, the neutron fluence to the superconducting magnet will be able to be reduced. However, it is not easy to achieve the lower streaming and penetration at the present. Therefore, investigations on irradiation behavior of superconducting magnet materials are desired and some novel researches have been performed from 1970s. In general, the critical current of the superconducting wire increases under fast neutron environment comparing with that of the non-irradiated wire, and then decreased to almost zero as an increase of neutron fluence. On the other hand, the critical temperature of the wire starts to get down around 10 22 n/m 2 of neutron fluence and the temperature margin will be decreased during the operation by the neutron irradiation. In this paper, some aspects of irradiated materials will be overviewed and general tendency will be discussed focussing on knock-on effect of fast neutron and long range ordering of A15 compounds

Staged deployment of the International Fusion Materials Irradiation Facility

International Nuclear Information System (INIS)

Takeuchi, H.; Sugimoto, M.; Nakamura, H.

2001-01-01

The International Fusion Materials Irradiation Facility (IFMIF) employs an accelerator based D-Li intense neutron source as defined in the 1995-96 Conceptual Design Activity (CDA) study. In 1999, IEA mandated a review of the CDA IFMIF design for cost reduction without change to its original mission. This objective was accomplished by eliminating the previously assumed possibility of potential upgrade of IFMIF beyond the user requirements. The total estimated cost was reduced from $797.2 M to $487.8 M. An option of deployment in 3 stages was also examined to reduce the initial investment and annual expenditures during construction. In this scenario, full performance is achieved gradually with each interim stage as follows. 1st Stage: 20% operation for material selection for ITER breeding blanket, 2nd Stage: 50% operation to demonstrate materials performance of a reference alloy for DEMO, 3rd Stage: full performance operation ( 2MW/m 2 at 500cm 3 ) to obtain engineering data for potential DEMO materials under irradiation up to 100-200 dpa. In summary, the new, reduced cost IFMIF design and staged deployment still satisfies the original mission. The estimated cost of the 1st Stage facility is only $303.6 M making it financially much more attractive. Currently, IFMIF Key Element Technology Phase (KEP) is underway to reduce the key technology risk factors. (author)

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

Irradiation-enhanced and-induced mass transport

International Nuclear Information System (INIS)

Rehn, L.E.

1989-01-01

Irradiation can be used to enhance diffusion, that is, to increase the rate at which equilibrium is attained, as well as to induce nonequilibrium changes. The main factors influencing whether irradiation will drive a material toward or away from equilibrium are the initial specimen microstructure and geometry, irradiation temperature, and primary recoil spectrum. This paper summarizes known effects of irradiation temperature and primary recoil spectrum on mass transport during irradiation. In comparison to either electron or heavy-ion irradiation, it is concluded that relatively low-energy, light-ion bombardment at intermediate temperatures offers the greatest potential to enhance the rate at which equilibrium is attained. The greatest departures from equilibrium can be expected from irradiation with similar particles at very low temperatures

Effects of gamma irradiation on raw materials and perfumes

International Nuclear Information System (INIS)

Guillot, M.; Pelpel, A.

1983-01-01

In order to enlight the strange problem of apparent perfume stability observed in manufactured talc powders sterilized by gamma rays, investigations were made on samples of odorant substances (raw materials, essential oils, or elaborated mixtures). As a rule, no immediate adulteration of olfactive caracteristics resulted at once from gamma irradiation. In several cases, a stabilizing effect appeared immediately and remained effective after long storage in various conditions (of temperature, or light, or oxygen exposure). This unexpected effect seems to be in accordance with previous experiments on gamma or electron irradiations of mixtures of organic molecules, reported in litterature: a mutual inhibition was observed to take place [fr

Distribution of products in polymer materials induced by ion-beam irradiation

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Masaki; Kudoh, Hisaaki; Sasuga, Tsuneo; Seguchi, Tadao [Japan Atomic Energy Research Inst., Tokyo (Japan); Hama, Yoshimasa; Hamanaka, Ken-ichi; Matsumoto, Hideya

1997-03-01

The depth profile of double bond formed in low density polyethylene (LDPE) sheet by ion beams irradiation was observed by a micro FT-IR spectrometer in order to investigate the linear energy transfer (LET) dependency on radiation effects to polymer materials. The distribution of double bond formation in LDPE by irradiation of light ions as H+ was found to be same with the dose distribution calculated from TRIM code, and the yield was also same with that by gamma-rays irradiation, which means that the LET dependency is very small. However, the distribution of double bond to depth was much different from the calculated depth-dose in heavy ions irradiation as Ar and Kr. Then, the dose evaluation was difficult from the TRIM code calculation for heavy ions. (author)

Irradiation of gelatin. Important applications for the development of new materials

Energy Technology Data Exchange (ETDEWEB)

Del Mastro, N. L., E-mail: nlmastro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05008-900 Sao Paulo (Brazil)

2011-11-15

Gelatin is obtained from a naturally occurring protein, collagen, by chemical and thermal hydrolysis. As a protein, gelatin is biodegradable. Among biomaterials, gelatin is an interesting material because the dehydrated gelatin is a partially crystalline polymer and has a relatively low melting point. Gelatin melts to a liquid when heated and solidifies when cooled again. Together with water, it forms a semi-solid colloid gel. If gelatin is placed in contact with cold water, some of the material dissolves, their solubility being determined by their manufacturing method. Ionizing radiation acting on aqueous biological systems produces labile intermediates. Macromolecular free radicals may be diffusion ally mobile in homogenous solutions but relatively immobile in gelled systems. In this paper, different works, some of them performed in our laboratory, are firstly described showing gelatin irradiation in diverse systems. We have already studied the mechanical properties of gelatin composites prepared with a natural fiber, plasticizer and treated by electron beam irradiation. Departing from that, the range of novel applications for gelatin composites like gelatin nanoparticles as biodegradable s and low cell toxic alternative carrier delivery systems are outlined. The potential and the possibilities of using gelatin irradiation for important applications for the development of new materials for medical and food industry are presented. (Author)

Binary-collision-approximation simulation for noble gas irradiation onto plasma facing materials

International Nuclear Information System (INIS)

Saito, Seiki; Nakamura, Hiroaki; Takayama, Arimichi; Ito, Atsushi M

2014-01-01

A number of experiments show that helium plasma constructs filament (fuzz) structures whose diameter is in nanometer-scale on the tungsten material under the suitable experimental condition. In this paper, binary-collision-approximation-based simulation is performed to reveal the mechanism and the conditions of fuzz formation of tungsten material under plasma irradiation. The irradiation of the plasma of hydrogen, deuterium, and tritium, and also the plasma of noble gas such as helium, neon, and argon atoms are investigated. The possibility of fuzz formation is discussed on the simulation result of penetration depth of the incident atoms

Instrumentation Technologies for Improving an Irradiation Testing of Nuclear Fuels and Materials at the HANARO

International Nuclear Information System (INIS)

Kim, Bong Goo; Park, Sung Jae; Choo, Ki Nam

2011-01-01

Over 50 years of nuclear fuels and materials irradiation testing has led to many countries developing significant improvements in instrumentation to monitor physical parameters and to control the test conditions in Materials Test Reactors (MTRs) or research reactors. Recent effort to deploy new fuels and materials in existing and advanced reactors has increased the demand for well-instrumented irradiation tests. Specifically, demand has increased for tests with sensors capable of providing real-time measurement of key parameters, such as temperature, geometry changes, thermal conductivity, fission gas release, cracking, coating buildup, thermal and fast flux, etc. This review paper documents the current state of instrumentation technologies in MTRs in the world and summarizes on-going research efforts to deploy new sensors. There is increased interest to irradiate new materials and reactor fuels for advanced PWRs and the Gen-IV reactor systems, such as SFRs (Sodium-cooled Fast Reactors), VHTRs (Very-High-Temperature Reactors), SCWRs (Supercritical-Water-cooled Reactors) and GFRs (Gas-cooled Fast Reactor). This review documents the current state of instrumentation technologies in MTRs in the world, identifies challenges faced by previous testing methods and how these challenges were overcome. A wide range of sensors are available to measure key parameters of interest during fuels and materials irradiations in MTRs. Such sensors must be reliable, small size, highly accurate, and able to withstand harsh conditions. On-going development efforts are focusing on providing MTR users a wider range of parameter measurements with increased accuracy. In addition, development efforts are focusing on reducing the impact of sensor on measurements by reducing sensor size. This report includes not only status of instrumentation using research reactors in the world to irradiate nuclear fuels and materials but also future directions relating to instrumentation technologies for

Inert materials for the GFR fuel. Characterizations, chemical interactions and irradiation damage

International Nuclear Information System (INIS)

Audubert, Fabienne; Carlot, Gaoelle; Lechelle, Jacques; David, Laurent; Gomes, Severine

2005-01-01

In the framework of an extensive R and D Program on GFR fuel, studies on inert materials have been performed at the French Atomic Energy Commission (CEA). The inert materials would be associated with the fuel with the aim of featuring an efficient barrier to radiotoxic species with regard to the cooling circuit of the reactor. Potential matrices identified for dispersion fuels or particles fuels are SiC, TiN, ZrN, ZrC, TiC. Physical microstructural and thermal properties have been determined in order to evaluate elaboration process effects. The evolution under irradiation of thermal properties (such as conductivity, diffusivity) of the materials has been studied using heavy ions to simulate fission product irradiation. After irradiation, scanning thermal microscopy is used to investigate the thermal degradation of the materials. Thermal conductivity variations were obtained on TiC irradiated with krypton ion at an energy of 86 MeV and a fluence of 5.10 15 ions.cm -2 . They are quantified at 19 W.m -1 .K -1 . On other materials such as SiC, ZrC, TiN, no thermal conductivity contrast was shown. Reactivity between the inert matrix (SiC or TiN) and the fuel (U, Pu)N have been evaluated on powders and on ceramic samples in contact by a thermal treatment under several atmospheres. It was shown that SiC reacts with (U, Pu)N in various atmospheres making secondary phases as PuSi 2 , USi 2 , U 20 Si 16 C 3 . TiN behaviour seems to be better: the only reactivity which may take place would be a variation of the nitrogen stoichiometry in TiN and (U, Pu)N at the interface. (author)

Irradiated stainless steel material constitutive model for use in the performance evaluation of PWR pressure vessel internals

Energy Technology Data Exchange (ETDEWEB)

Rashid, J.Y.; Dunham, R.S. [ANATECH (United States); Demma, A. [Electric Power Research Institute - EPRI (United States)

2011-07-01

Demonstration of component functionality requires analytical simulations of reactor internals behavior. Towards that aim, EPRI has undertaken the development of irradiated material constitutive model and damage criteria for use in global and local finite-element based functionality analysis methodology. The constitutive behavioral regimes of irradiated stainless steel types 316 and 304 materials included in the model consist of: elastic-plastic material response considering irradiation hardening of the stress-strain curve, irradiation creep, stress relaxation, and void swelling. IASCC and degradation of ductility with irradiation are the primary damage mechanisms considered in the model. The material behavior model development consists of two parts: the first part is a user-material subroutine that can interface with a general-purpose finite element computer program to adapt it to the special-purpose of functionality analysis of reactor internals. The second part is a user utility in the form of Excel Spread sheets that permit users to extract a given property, e.g. the elastic-plastic stress-strain curve, creep curve, or void-swelling curve, as function of the relevant independent variables. The development of the model takes full advantage of the significant work that has been undertaken within EPRI's Material Reliability Program (MRP) to improve the knowledge of the material properties of irradiated stainless steels. Data from EPRI's MRP database have been utilized to develop equations that characterize the yield strength, ultimate tensile strength, uniform elongation, total elongation, reduction in area, void swelling and irradiation creep of stainless steels in a PWR environment. It is noted that, while the development of the model's equations has been statistically faithful to the material database, approximations were introduced in the model to ensure appropriate conservatism in the model's application consistently with accepted

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)

Transmission electron microscopy of oxide dispersion strengthened (ODS) molybdenum: effects of irradiation on material microstructure

International Nuclear Information System (INIS)

Baranwal, R.; Burke, M.G.

2003-01-01

Oxide dispersion strengthened (ODS) molybdenum has been characterized using transmission electron microscopy (TEM) to determine the effects of irradiation on material microstructure. This work describes the results-to-date from TEM characterization of unirradiated and irradiated ODS molybdenum. The general microstructure of the unirradiated material consists of fine molybdenum grains (< 5 (micro)m average grain size) with numerous low angle boundaries and isolated dislocation networks. 'Ribbon'-like lanthanum oxides are aligned along the working direction of the product form and are frequently associated with grain boundaries, serving to inhibit grain boundary and dislocation movement. In addition to the 'ribbons', discrete lanthanum oxide particles have also been detected. After irradiation, the material is characterized by the presence of nonuniformly distributed large (∼ 20 to 100 nm in diameter), multi-faceted voids, while the molybdenum grain size and oxide morphology appear to be unaffected by irradiation

Dose requirements for microbial decontamination of botanical materials by irradiation

International Nuclear Information System (INIS)

Razem, D.; Katusin-Razem, Branka

2002-01-01

Microbial contamination levels and corresponding resistivities to irradiation (expressed as dose required for the first 90% reduction, D first 9 0% r ed ) were analyzed in a number of various botanical materials. The following generalizations could be made: total aerobic plate count is the most informative measure of contamination; the probability of contamination depends on available surface of the material and processing history: flowers and leaves usually contain more contamination than fruits and seeds, while crude herbs contain more than extracts; liquid extracts are more contaminated than dry ones. At the same time, resistivity to irradiation increases approximately in the reverse order of contamination level on going from flowers and leaves, to fruits and seeds, to liquid and dry extracts. The two quantities, probability of contamination and D first 9 0% r ed being inversely related, the treatment dose needed to reduce initial contamination to tolerable level amounts to between 4 and 30 kGy under a typical scenario, and between 8 and 40 kGy under the worst-case scenario for the whole range of raw materials and botanical products

Genetic similarity among commercial oil palm materials based on microsatellite markers

Directory of Open Access Journals (Sweden)

Diana Arias

2012-08-01

Full Text Available Microsatellite markers are used to determine genetic similarities among individuals and might be used in various applications in breeding programs. For example, knowing the genetic similarity relationships of commercial planting materials helps to better understand their responses to environmental, agronomic and plant health factors. This study assessed 17 microsatellite markers in 9 crosses (D x P of Elaeis guineensis Jacq. from various commercial companies in Malaysia, France, Costa Rica and Colombia, in order to find possible genetic differences and/or similarities. Seventy-seven alleles were obtained, with an average of 4.5 alleles per primer and a range of 2-8 amplified alleles. The results show a significant reduction of alleles, compared to the number of alleles reported for wild oil palm populations. The obtained dendrogram shows the formation of two groups based on their genetic similarity. Group A, with ~76% similarity, contains the commercial material of 3 codes of Deli x La Mé crosses produced in France and Colombia, and group B, with ~66% genetic similarity, includes all the materials produced by commercial companies in Malaysia, France, Costa Rica and Colombia

Investigation of cryogenic irradiation influence on mechanical and physical properties of ITER magnetic system insulation materials

International Nuclear Information System (INIS)

Kozlov, A.V.; Scherbacov, E.N.; Dudchenko, N.A.; Shihalev, V.S.; Bedin, V.V.; Paltusov, N.A.; Korsunskiy, V.E.

1998-01-01

A set of methods of cryogenic irradiation influence test on mechanical and physical properties of insulation of ITER magnetic system are presented in this paper. Investigations are carried out without intermediate warming up of samples. A Russian insulating composite material was irradiated in the IVV-2M reactor. The ratio of energy absorbed by insulation materials from neutron irradiation to that from gamma irradiation can be varied from ∝(25:75)% to ∝(50:50)% in the reactor. The test results on the thermal expansion, thermal conductivity and gas evolution of the above material are presented. It was shown, that cryogenic irradiation up to the fluence ∝2 x 10 22 n/m 2 (E ≥ 0.1 MeV) leads to 0.27% linear size changes along layers of fiber-glass, the thermal conductivity coefficient is decreased on 15% at 100 k in perpendicular direction to fiber-glass plane, and thermal coefficient of linear expansion (TCLE) has anomalous temperature dependence. (orig.)

Femtosecond Laser Irradiation of Plasmonic Nanoparticles in Polymer Matrix: Implications for Photothermal and Photochemical Material Alteration

Directory of Open Access Journals (Sweden)

Anton A. Smirnov

2014-11-01

Full Text Available We analyze the opportunities provided by the plasmonic nanoparticles inserted into the bulk of a transparent medium to modify the material by laser light irradiation. This study is provoked by the advent of photo-induced nano-composites consisting of a typical polymer matrix and metal nanoparticles located in the light-irradiated domains of the initially homogeneous material. The subsequent irradiation of these domains by femtosecond laser pulses promotes a further alteration of the material properties. We separately consider two different mechanisms of material alteration. First, we analyze a photochemical reaction initiated by the two-photon absorption of light near the plasmonic nanoparticle within the matrix. We show that the spatial distribution of the products of such a reaction changes the symmetry of the material, resulting in the appearance of anisotropy in the initially isotropic material or even in the loss of the center of symmetry. Second, we analyze the efficiency of a thermally-activated chemical reaction at the surface of a plasmonic particle and the distribution of the product of such a reaction just near the metal nanoparticle irradiated by an ultrashort laser pulse.

Bulk-shield design for the Fusion Materials Irradiation Test facility

International Nuclear Information System (INIS)

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

1982-07-01

The accelerator-based Fusion Materials Irradiation Test (FMIT) facility will provide a high-fluence, fusion-like radiation environment for the testing of materials. While the neutron spectrum produced in the forward direction by the 35 MeV deuterons incident upon a flowing lithium target is characterized by a broad peak around 14 MeV, a high energy tail extends up to about 50 MeV. Some shield design considerations are reviewed

TEM investigation of plant-irradiated NPP bolt material

International Nuclear Information System (INIS)

Pakarinen, J.; Ehrnsten, U.; Keinaenen, H.; Karlsen, W.; Karlsen, T.

2015-01-01

Analytical transmission electron microscopy (ATEM) was used to examine irradiation-induced damage in material removed from two different bolts from two different nuclear power plants. One section came from a French PWR, was made of CW AISI 316, and included a section of the bolt that had accumulated a dose of approximately 15 dpa during 19 operation cycles at 350 - 390 C. degrees. Another section came from a VVER bolt that was removed from the plant due to indications found in non-destructive examinations (NDE). The VVER bolt was made of solution annealed titanium stabilized 0X18H10T (corresponding to Type AISI 321) and had accumulated a fluence of 2.9 dpa. During the removal of that bolt, it was found that the bolt washer had been inappropriately spot welded to the shielding plate during assembly. Destructive investigations showed that the bolt had two large intergranular cracks, and the TEM samples were prepared from the material adjacent to those cracks. The PWR bolt had not failed, although cracks in the bolts with a similar history had been found previously. The fluence for the cold-worked AISI 316 PWR bolt was estimated to be about 15 dpa. Both the examined bolts showed a clear radiation induced segregation of alloying elements at the grain boundaries (GB-RIS), the presence of dislocation loops, the formation of precipitates, and linear deformation microstructures. Additionally, voids were found from the PWR bolt and the VVER bolt had a high density of dislocations. (authors)

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)

STUDY STRUCTURE OF THREE-COMPONENT POLYMERIC MATERIAL UNDER INFLUENCE OF γ-IRRADIATION

Directory of Open Access Journals (Sweden)

V. T. Tarasyuk

2017-01-01

Full Text Available The polymer material (РА/РЕ/Eva with a width of 55 μm was studied. Sterilization was carried out on the unit GU–200 at doses from 3 to 18 kGy in the Research Institute of Technical Physics and Automation, Rosatom, Moscow, Russia. The structure of the polymermaterial samples was studied by IR spectroscopy before and after irradiation in a range of 400–5000 сm–1. According to the results of the analysis of the IR spectrum structure, the changes in the structure were insignificant upon irradiation at doses up to 6 kGy. Upon irradiation at doses from 9 kGy and higher, an increase in quantity of ester groups (2340 сm–1 and insignificant increase in other functional groups were observed, which can suggest a simultaneous process of intra-molecular cross-linking with the intermediatestage of cross-linking occurring with formation of vinylene groups. This causes destruction of a polymer material and radiation oxidation. These disorders can lead to changes in physico-mechanical and barrier parameters of a polymer material, which can be notably reflected in the shelf life of agricultural products.

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.

Effect of Fast Neutron Irradiation on Current Transport Properties of HTS Materials

CERN Document Server

Ballarino, A; Kruglov, V S; Latushkin, S T; Lubimov, A N; Ryazanov, A I; Shavkin, S V; Taylor, T M; Volkov, P V

2004-01-01

The effect of fast neutron irradiation with energy up to 35 MeV and integrated fluence of up to 5 x 10**15 cm-2 on the current transport properties of HTS materials Bi-2212 and Bi-2223 has been studied, both at liquid nitrogen and at room temperatures. The samples irradiated were selected after verification of the stability of their superconducting properties after temperature cycling in the range of 77 K - 293 K. It has been found that the irradiation by fast neutrons up to the above dose does not produce a significant degradation of critical current. The effect of room temperature annealing on the recovery of transport properties of the irradiated samples is also reported, as is a preliminary microstructure investigation of the effect of irradiation on the soldered contacts.

Effects of irradiation on four solid breeder materials

International Nuclear Information System (INIS)

Hollenberg, G.W.

1984-01-01

The tritium breeding material with the highest lithium atom density, Li 2 O has been observed to incur significant swelling (>4%) under fast reactor irradiation. Such swelling, if unrestrained leads to either unacceptable, induced-strains in adjacent structural material or undesirable design compromises. Fortunately, however, Li 2 O deforms at low temperatures so that swelling strains may be internally accommodated. Laboratory dilational creep experiments were conducted on unirraciated Li 2 O between 500 and 700 0 C in order to provide data for structural analysis of in-reactor experiments and blanket design studies. A densification model agreed with most of the available data. 15 refs

Contamination confinement system of irradiated materials handling laboratories

International Nuclear Information System (INIS)

Lobao, A. dos S.T.; Araujo, J.A. de; Camilo, R.L.

1988-06-01

A study to prevent radioctivity release in lab scale is presented. As a basis for the design all the limits established by the IAEA for ventilation systems were observed. An evaluation of the different parameters involved in the design have been made, resulting in the especification of the working areas, ducts and filtering systems in order to get the best conditions for the safe handling of irradiated materials. (author) [pt

Testing capabilities of Los Alamos National Laboratory for irradiated materials

International Nuclear Information System (INIS)

Maloy, S.A.; James, M.R.; Sommer, W.F.

1999-01-01

Spallation neutron sources expose materials to high energy (>100 MeV) proton and neutron spectra. Although numerous studies have investigated the effects of radiation damage in a lower energy neutron flux from fission or fusion reactors on the mechanical properties of materials, very little work has been performed on the effects that exposure to a spallation neutron spectrum has on the mechanical properties of materials. These effects can be significantly different than those observed in a fission or fusion reactor spectrum because exposure to high energy protons and neutrons produces more He and H along with the atomic displacement damage. Los Alamos National Laboratory has unique facilities to study the effects of spallation radiation damage on the mechanical properties of materials. The Los Alamos Neutron Science Center (LANSCE) has a pulsed linear accelerator which operates at 800 MeV and 1 mA. The Los Alamos Spallation Radiation Effect Facility (LASREF) located at the end of this accelerator is designed to allow the irradiation of components in a proton beam while water cooling these components and measuring their temperature. After irradiation, specimens can be investigated at hot cells located at the Chemical Metallurgy Research Building. Wing 9 of this facility contains 16 hot cells set up in two groups of eight, each having a corridor in the center to allow easy transfer of radioactive shipments into and out of the hot cells. These corridors have been used to prepare specimens for shipment to collaborating laboratories such as PNNL, ORNL, BNL, and the Paul Scherrer Institute to perform specialized testing at their hot cells. The LANL hot cells contain capabilities for opening radioactive components and testing their mechanical properties as well as preparing specimens from irradiated components

A TEM method for analyzing local strain fields in irradiated materials

International Nuclear Information System (INIS)

Bennetch, J.I.; Jesser, W.A.

1983-01-01

Of great interest to the field of fracture mechanics is the strain field in front of a crack tip. In irradiated materials, cavities which naturally form as a result of radiation provide convenient internal markers. If a miniaturized irradiated tensile sample is pulled in situ in a transmission electron microscope (TEM), both the relative displacement of these cavities and their distortion in shape provide information on localized strain on a microscopic level. In addition, the TEM method allows direct correlation of active slip systems with crack propagation characteristics. To illustrate this method a strain field map was constructed about a crack propagating in a helium irradiated type 316 stainless steel sample containing large cavities. (orig.)

Assessment of repair welding technologies of irradiated materials

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Damages on reactor internals of stainless steels caused by stress corrosion cracking and fatigue were identified in aged BWR plants. Repair-welding is one of the practical countermeasure candidates to restore the soundness of components and structures. The project of 'Assessment of Repair welding Technologies of Irradiated Materials' has been carried out to develop the technical guideline regarding the repair-welding of reactor internals. In FY 2011, we investigated the fatigue strength of stainless steel SUS316L irradiated by YAG laser welding. Furthermore, revision of the technical guideline regarding the repair-welding of reactor internals was discussed. Diagram of tungsten inert gas (TIG) weld cracking caused by entrapped Helium was modified. Helium concentration for evaluation-free of TIG weld cracking caused by entrapped Helium was revised to 0.007appm from 0.01appm. (author)

The development of fuel pins and material specimens mixed loading irradiation test rig in the experimental fast reactor Joyo. The development of the fuel-material hybrid rig

International Nuclear Information System (INIS)

Oyamatsu, Yasuko; Someya, Hiroyuki

2013-02-01

In the experimental fast reactor Joyo, there were many tests using the irradiation rigs that it was possible to be set irradiation conditions for each compartment independently. In case of no alternative fuel element to irradiate after unloading the irradiated compartments, the irradiation test was restarted with the dummy compartment which the fuel elements was not mounted. If the material specimens are mounted in this space, it is possible to use the irradiation space effectively. For these reasons, the irradiation rig (hybrid rig) is developed that is consolidated with material specimens compartment and fuel elements compartment. Fuel elements and material specimens differ greatly with heat generation, so that the most important issue in developing of hybrid rig is being able to distribute appropriately the coolant flow which satisfies irradiation conditions. The following is described by this report. (1) It was confirmed that the flow distribution of loading the same irradiation rig with the compartment from which a flow demand differs could be satisfied. (2) It was confirmed that temperature setting range of hybrid rig could be equivalent to that of irradiation condition. (3) By standardizing the coolant entrance structure of the compartment lower part, the prospect which can perform easily recombination of the compartment from which a type differs between irradiation rigs was acquired. (author)

Reactivation of X-irradiated cell material during limb regeneration in Urodeles Amphibians

International Nuclear Information System (INIS)

Desselle, J.C.

1979-10-01

In amputated members irradiated with X-rays the regeneration power is inhibited. This power is restored by grafts of healthy tissue in the irradiated members. The origin of the cell material of the restored regeneration blastema has been studied by an original labelling technique. The different amounts of DNA in the graft cells and those of the stump mark the graft cells during the regeneration process. It was shown that the graft causes a reactivation of the inhibited stump cells and the reactivation stages are the same as the activation stages of the member regenerating normally. It was also established that during restored regeneration the cell material implanted in the irradiated members contributes, by the 160th day of regeneration, 4.5% of the cartilaginous regenerate cells and 12% of the muscle cells. All the other regenerate cells are supplied by the cells of the stump; these are reactivated and together with the activated graft cells lead to the restitution of the amputated member [fr

Evaluation of thermal shock strengths for graphite materials using a laser irradiation method

International Nuclear Information System (INIS)

Kim, Jae Hoon; Lee, Young Shin; Kim, Duck Hoi; Park, No Seok; Suh, Jeong; Kim, Jeng O.; Il Moon, Soon

2004-01-01

Thermal shock is a physical phenomenon that occurs during the exposure to rapidly high temperature and pressure changes or during quenching of a material. The rocket nozzle throat is exposed to combustion gas of high temperature. Therefore, it is important to select suitable materials having the appropriate thermal shock resistance and to evaluate these materials for rocket nozzle design. The material of this study is ATJ graphite, which is the candidate material for rocket nozzle throat. This study presents an experimental method to evaluate the thermal shock resistance and thermal shock fracture toughness of ATJ graphite using laser irradiation. In particular, thermal shock resistance tests are conducted with changes of specimen thickness, with laser source irradiated at the center of the specimen. Temperature distributions on the specimen surface are detected using type K and C thermocouples. Scanning electron microscope (SEM) is used to observe the thermal cracks on specimen surface

Botryllus schlosseri (Tunicata) whole colony irradiation: Do senescent zooid resorption and immunological resorption involve similar recognition events

International Nuclear Information System (INIS)

Rinkevich, B.; Weissman, I.L.

1990-01-01

The colonial tunicate Botryllus schlosseri undergoes cyclic blastogenesis where feeding zooids are senescened and resorbed and a new generation of zooids takes over the colony. When non-identical colonies come into direct contact, they either reject each other or fuse. Fusion is usually followed by the resorption of one of the partners in the chimera (immunological resorption). The striking morphological similarities between the two resorption phenomena suggest that both may involve tissue destruction following self-nonself recognition events. Here we attempt to modify these two events by whole colony gamma irradiation assays. Three sets of experiments were performed: (1) different doses of whole colony irradiation for determination of irradiation effects (110 colonies); (2) pairs of irradiated-nonirradiated isografts of clonal replicates for the potential of reconstruction of the irradiated partners (23 pairs); (3) chimeras of irradiated-nonirradiated partners for analysis of resorption hierarchy. Mortality increased with the irradiation dose. All colonies exposed to more than 5,000 rads died within 19 days, while no colony died below 2,000 rads. The average mortality periods, in days, for doses of 6,000-8,000, 5,000, and 2,500-4,000 rads were 14.4 +/- 3.1 (n = 24), 19.8 +/- 6.0 (n = 15), and 19.6 + 5.1 (n = 22), respectively. Younger colonies (3-6 months old) may survive radiation better than older ones (more than 13 months). Many morphological alterations were recorded in irradiated colonies: ampullar contraction and/or dilation, accumulation of pigment cells within ampullae, abnormal bleeding from blood vessels, sluggish blood circulation, necrotic zones, reduction in bud number, and irregularities in zooid and system structures. With doses of 3,000-4,000 rads and above, irradiation arrested the formation of new buds and interrupted normal takeover

Irradiation as an alternative environment friendly method for microbiological decontamination of herbal raw material

International Nuclear Information System (INIS)

Gorecki, P.; Kedzia, B.; Migdal, W.; Owczarczyk, H.B.

1998-01-01

Microbiological contamination of herbal raw materials is a serious problem in the production of therapeutical preparations. A good quality of the product, according to the pharmaceutical requirements may be achieved by applying suitable methods of decontamination. The decontamination treatments should be fast and effective against all microorganisms. It should ensure the decontamination of both packaging and the product in order to act effectively against all the microorganisms present and must not reduce the sensory and technological qualities of the commodities. In the paper, the results of comparative investigations on the microbiological decontamination of herbal raw materials by chemical (ethylene oxide, methyl bromide) and physical method (irradiation) are presented. Decontamination of herbal raw materials by irradiation is a method by choice. It is because chemical methods have been recognized recently as not safe to the consumer. Irradiation, in turn, is technically feasible, very effective and friendly enough to environment process

Remote-handling demonstration tests for the Fusion Materials Irradiation Test (FMIT) Facility

International Nuclear Information System (INIS)

Shen, E.J.; Hussey, M.W.; Kelly, V.P.; Yount, J.A.

1982-01-01

The mission of the Fusion Materials Irradiation Test (FMIT) Facility is to create a fusion-like environment for fusion materials development. Crucial to the success of FMIT is the development and testing of remote handling systems required to handle materials specimens and maintenance of the facility. The use of full scale mock-ups for demonstration tests provides the means for proving these systems

Binary-collision-approximation-based simulation of noble gas irradiation to tungsten materials

International Nuclear Information System (INIS)

Saito, Seiki; Takayama, Arimichi; Ito, Atsushi M.; Nakamura, Hiroaki

2013-01-01

To reveal the possibility of fuzz formation of tungsten material under noble gas irradiation, helium, neon, and argon atom injections into tungsten materials are performed by binary-collision-approximation-based simulation. The penetration depth is strongly depends on the structure of the target material. Therefore, the penetration depth for amorphous and bcc crystalline structure is carefully investigated in this paper

Irradiation testing of stainless steel plate material and weldments. Report on ITER Task T14, Part B. Tensile properties after 0.5 and 5 dpa at 350 and 500 K

International Nuclear Information System (INIS)

Rensman, J.W.; Boskeljon, J.; Horsten, M.G.; De Vries, M.I.

1997-10-01

The tensile properties of unirradiated and neutron irradiated type 316L(N)-SPH stainless steel plate, EB weldments, 16-8 TIG-weldments, and full 16-8 TIG-deposits have been measured. Miniature 4 mm diameter test specimens of the European Reference Heat 1 and 2 (ERH), and 4 mm and some 8 mm diameter specimens of the weldments mentioned above, were irradiated in the High Flux Reactor (HFR) in Petten, The Netherlands, simulating the first wall conditions by a combination of high displacement damage with high amounts of helium. The irradiation conditions were 0.5 and 5 displacements per atom (dpa) at 350K and 0.5 and 5 dpa at 500K. Testing temperatures ranged from 300K to 850K. This work was performed as part of the European Fusion Technology Programme for ITER as 'Irradiation testing of stainless steel' The report contains the experimental conditions and summarises the results. The tensile properties of the unirradiated ERH's 1 and 2 plate materials were found to differ slightly but significantly: ERH2 has a lower UTS, but higher yield strength and ductility than ERH1. The plate materials have lower yield strength in the unirradiated condition than all of the weldments (EB, TIG-weld and TIG-deposit), accompanied by a higher ductility of the plate materials. When irradiated at 350K the differences in strength between the plate and weld materials decrease, but the ductility of the plate remains higher than that of the weldments. A saturation of irradiation damage has taken place already at about 0.5 dpa. When irradiated at 500K the plate material continuously hardens up to 5 dpa, where it has lost all uniform plastic ductility. The weldments show similar but less dramatic hardening and loss of ductility as the plate material for both irradiation conditions. 54 figs., 17 tabs., 21 refs

Effect of material variables on the irradiation performance of boron carbide

International Nuclear Information System (INIS)

Basmajian, J.A.; Hollenberg, G.W.

1980-01-01

Boron carbide pellets were fabricated with variations in material parameters. These pellets were irradiated in the Experimental Breeder Reactor-II (EBR-II) to determine the effect of these variations on the performance. Helium release from the material and swelling of the pellets are the primary measures of performance. It was determined that material with a smaller grain size released more helium and swelled less. The pellets with boron-to-carbon ratios greater than 4 to 1 did not perform well. Iron additions improved the performance of the material while density variations had little effect

Pyrolysis responses of kevlar/epoxy composite materials on laser irradiating

Science.gov (United States)

Liu, Wei-ping; Wei, Cheng-hua; Zhou, Meng-lian; Ma, Zhi-liang; Song, Ming-ying; Wu, Li-xiong

2017-05-01

The pyrolysis responses of kevlar/epoxy composite materials are valuable to study in a case of high temperature rising rate for its widely application. Distinguishing from the Thermal Gravimetric Analysis method, an apparatus is built to research the pyrolysis responses of kevlar/epoxy composite materials irradiated by laser in order to offer a high temperature rising rate of the sample. By deploying the apparatus, a near real-time gas pressure response can be obtained. The sample mass is weighted before laser irradiating and after an experiment finished. Then, the gas products molecular weight and the sample mass loss evolution are derived. It is found that the pressure and mass of the gas products increase with the laser power if it is less than 240W, while the molecular weight varies inversely. The variation tendency is confusing while the laser power is bigger than 240W. It needs more deeper investigations to bring it to light.

Multi-MW accelerator target material properties under proton irradiation at Brookhaven National Laboratory linear isotope producer

Science.gov (United States)

Simos, N.; Ludewig, H.; Kirk, H.; Dooryhee, E.; Ghose, S.; Zhong, Z.; Zhong, H.; Makimura, S.; Yoshimura, K.; Bennett, J. R. J.; Kotsinas, G.; Kotsina, Z.; McDonald, K. T.

2018-05-01

The effects of proton beams irradiating materials considered for targets in high-power accelerator experiments have been studied using the Brookhaven National Laboratory's (BNL) 200 MeV proton linac. A wide array of materials and alloys covering a wide range of the atomic number (Z) are being scoped by the high-power accelerator community prompting the BNL studies to focus on materials representing each distinct range, i.e. low-Z, mid-Z and high-Z. The low range includes materials such as beryllium and graphite, the midrange alloys such as Ti-6Al-4V, gum metal and super-Invar and finally the high-Z range pure tungsten and tantalum. Of interest in assessing proton irradiation effects are (a) changes in physiomechanical properties which are important in maintaining high-power target functionality, (b) identification of possible limits of proton flux or fluence above which certain materials cease to maintain integrity, (c) the role of material operating temperature in inducing or maintaining radiation damage reversal, and (d) phase stability and microstructural changes. The paper presents excerpt results deduced from macroscopic and microscopic post-irradiation evaluation (PIE) following several irradiation campaigns conducted at the BNL 200 MeV linac and specifically at the isotope producer beam-line/target station. The microscopic PIE relied on high energy x-ray diffraction at the BNL NSLS X17B1 and NSLS II XPD beam lines. The studies reveal the dramatic effects of irradiation on phase stability in several of the materials, changes in physical properties and ductility loss as well as thermally induced radiation damage reversal in graphite and alloys such as super-Invar.

Size-Tuned Plastic Flow Localization in Irradiated Materials at the Submicron Scale

Science.gov (United States)

Cui, Yinan; Po, Giacomo; Ghoniem, Nasr

2018-05-01

Three-dimensional discrete dislocation dynamics (3D-DDD) simulations reveal that, with reduction of sample size in the submicron regime, the mechanism of plastic flow localization in irradiated materials transitions from irradiation-controlled to an intrinsic dislocation source controlled. Furthermore, the spatial correlation of plastic deformation decreases due to weaker dislocation interactions and less frequent cross slip as the system size decreases, thus manifesting itself in thinner dislocation channels. A simple model of discrete dislocation source activation coupled with cross slip channel widening is developed to reproduce and physically explain this transition. In order to quantify the phenomenon of plastic flow localization, we introduce a "deformation localization index," with implications to the design of radiation-resistant materials.

Ion-irradiation-induced damage in nuclear materials: Case study of a-SiO2 and MgO

International Nuclear Information System (INIS)

Bachiller-Perea, Diana

2016-01-01

One of the most important challenges in Physics today is the development of a clean, sustainable, and efficient energy source that can satisfy the needs of the actual and future society producing the minimum impact on the environment. For this purpose, a huge international research effort is being devoted to the study of new systems of energy production; in particular, Generation IV fission reactors and nuclear fusion reactors are being developed. The materials used in these reactors will be subjected to high levels of radiation, making necessary the study of their behavior under irradiation to achieve a successful development of these new technologies. In this thesis two materials have been studied: amorphous silica (a-SiO 2 ) and magnesium oxide (MgO). Both materials are insulating oxides with applications in the nuclear energy industry. High-energy ion irradiations have been carried out at different accelerator facilities to induce the irradiation damage in these two materials; then, the mechanisms of damage have been characterized using principally Ion Beam Analysis (IBA) techniques. One of the challenges of this thesis was to develop the Ion Beam Induced Luminescence or iono-luminescence (which is not a widely known IBA technique) and to apply it to the study of the mechanisms of irradiation damage in materials, proving the power of this technique. For this purpose, the iono-luminescence of three different types of silica (containing different amounts of OH groups) has been studied in detail and used to describe the creation and evolution of point defects under irradiation. In the case of MgO, the damage produced under 1.2 MeV Au + irradiation has been characterized using Rutherford backscattering spectrometry in channeling configuration and X-ray diffraction. Finally, the iono-luminescence of MgO under different irradiation conditions has also been studied.The results obtained in this thesis help to understand the irradiation-damage processes in materials

Consequences of the improvement of fast reactor material behavior under irradiation on fuel element performance

International Nuclear Information System (INIS)

Leclere, J.; Dupouy, J.M.; Marcon, J.P.

1979-01-01

The most important problems in fast reactor fuel element come from the excessive swelling of the structural materials used. The limitations of irradiation time for a given reactor result from the cladding or hexagonal wrapper deformations. Irradiation creep plays a major role, either in inducing additional deformations, or in providing possible ways of accommodation of bending stresses. Progress has been made in designing swelling resistant and/or low irradiation creep modulus materials. For instance in FRANCE, annealed 316 SS has been eliminated from pin and subassembly, and replaced by cold worked 316; we are now considering introduction of stabilizing elements in 316 SS as a further improvement and studying different alloys (nickel alloys, or ferritic steels). It has to be checked that the improvement of irradiation characteristic is not counterbalanced by losses on other properties (embrittlement for instance). Considering that pushing off or eliminating a limit may lead to the onset of a new one, it is porposed to make a review of the consequences of substantial improvement of structural material behavior

Modelling irradiation effects in fusion materials

International Nuclear Information System (INIS)

Victoria, M.; Dudarev, S.; Boutard, J.L.; Diegele, E.; Laesser, R.; Almazouzi, A.; Caturla, M.J.; Fu, C.C.; Kaellne, J.; Malerba, L.; Nordlund, K.; Perlado, M.; Rieth, M.; Samaras, M.; Schaeublin, R.; Singh, B.N.; Willaime, F.

2007-01-01

We review the current status of the European fusion materials modelling programme. We describe recent findings and outline potential areas for future development. Large-scale density functional theory (DFT) calculations reveal the structure of the point defects in α-Fe, and highlight the crucial part played by magnetism. The calculations give accurate migration energies of point defects and the strength of their interaction with He atoms. Kinetic models based on DFT results reproduce the stages of radiation damage recovery in iron, and stages of He-desorption from pre-implanted iron. Experiments aimed at validating the models will be carried out in the future using a multi-beam ion irradiation facility chosen for its versatility and rapid feedback

Modelling irradiation effects in fusion materials

Energy Technology Data Exchange (ETDEWEB)

Victoria, M. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, c/Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Dudarev, S. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Oxfordshire OX14 3DB, UK and Department of Physics, Imperial College, Exhibition Road, London SW7 2AZ (United Kingdom); Boutard, J.L. [EFDA-CSU Garching, Boltzmannstrasse 2, D-85748 Garching (Germany)], E-mail: jean-louis.boutard@tech.efda.org; Diegele, E.; Laesser, R. [EFDA-CSU Garching, Boltzmannstrasse 2, D-85748 Garching (Germany); Almazouzi, A. [Structural Materials Expert Group, Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Caturla, M.J. [Departamento de Fisica Aplicada, Universidad de Alicante, 03690 San Vicente de Raspeig (Spain); Fu, C.C. [Service de Metallurgie Physique, CEA/Saclay, F-91191 Gif sur Yvette Cedex (France); Kaellne, J. [Department of Engineering Sciences, Uppsala University, Box 534, S-751 21 Uppsala (Sweden); Malerba, L. [Structural Materials Expert Group, Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Nordlund, K. [Association EURATOM-Tekes, Accelerator Laboratory, P.O. Box 43, 00014 University of Helsinki (Finland); Perlado, M. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, c/Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Rieth, M. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung I, P.O. Box 3640, D-76021 Karlsruhe (Germany); Samaras, M. [Paul Scherrer Institute, Nuclear Energy and Safety Department, CH-5232 Villigen PSI (Switzerland); Schaeublin, R. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-5232 Villigen PSI (Switzerland); Singh, B.N. [Department of Materials Research, Risoe National Laboratory, DK-4000 Roskilde (Denmark); Willaime, F. [Service de Metallurgie Physique, CEA/Saclay, F-91191 Gif sur Yvette Cedex (France)

2007-10-15

We review the current status of the European fusion materials modelling programme. We describe recent findings and outline potential areas for future development. Large-scale density functional theory (DFT) calculations reveal the structure of the point defects in {alpha}-Fe, and highlight the crucial part played by magnetism. The calculations give accurate migration energies of point defects and the strength of their interaction with He atoms. Kinetic models based on DFT results reproduce the stages of radiation damage recovery in iron, and stages of He-desorption from pre-implanted iron. Experiments aimed at validating the models will be carried out in the future using a multi-beam ion irradiation facility chosen for its versatility and rapid feedback.

Radiation damage and materials performance in irradiation environment

International Nuclear Information System (INIS)

Singh, B.N.

2009-01-01

Collisions of energetic projectile particles with host atoms produce atomic displacements in the target materials. Subsequently, some of these displacements are transformed into lattice defects and survive in the form of single defects and of defect clusters. Depending on the ambient temperature, these defects and their clusters diffuse, interact, annihilate, segregate and accumulate in various forms and are responsible for the evolution of the irradiation-induced microstructure. Naturally, both physical and mechanical properties and thereby the performance and lifetime of target materials are likely to be determined by the nature and the magnitude of the accumulated defects and their spatial dispositions. The defect accumulation, microstructural evolution and the resulting materials response gets very complicated particularly under the reactor operational conditions. The complication arises from the fact that the materials used in the structural components will experience concurrently generation of defects produced by the flux of neutrons and generation of dislocations due to plastic deformation. In other words, the defect accumulation will have to be considered under the conditions of two interactive reaction kinetics operating simultaneously. Both materials and experimental variables are likely to affect the damage accumulation and thereby the materials performance. Experimental and theoretical results pertaining to effects of major materials and experimental variables on materials performance will be briefly examined. (au)

IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

International Nuclear Information System (INIS)

Martone, M.

1997-01-01

This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

Energy Technology Data Exchange (ETDEWEB)

Martone, M [ENEA, Centro Ricerche Frascati, Rome (Italy)

1997-01-01

This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

How can we predict microstructural changes caused by the multiscale irradiation process occurred in materials having complicated and hierarchical structures?

International Nuclear Information System (INIS)

Morishita, Kazunori; Watanabe, Yoshiyuki; Yoshimatsu, Jun-ichi

2008-01-01

Challenging efforts are discussed to establish an advanced methodology for prediction of material's property and performance changes by irradiation, which will be necessary by all means for the advanced reactor maintenance technology in the future. The changes of material's properties and performance caused by irradiation, such as irradiation-induced hardening, ductility loss, and material's degradation leading to reduction in reactor lifetime, are primarily determined by microstructural changes in materials during irradiation, where athermal lattice defects are continuously produced by collisions between an irradiating particle and a target material atom, and subsequently the defects are aggregated via diffusion in the form of dislocation loops, voids, and solute precipitation. These radiation damage processes are in essence multiscale phenomena, which involve varying time- and length-scales, from ballistic binary collisions to collective atomic motion in the thermal spike stage followed by the thermal activation process. In this report, the multiscale modeling approach is proposed to understand the processes in materials having complicated and hierarchical structures. (author)

Method and equipment to lead a cable-like material under an irradiation source

International Nuclear Information System (INIS)

Riesselmann, F.J.

1975-01-01

When irradiating cable-like material (cable jacketed with polyethylene) which is led through an irradiation source and is thus turned and twisted, no uniform irradiation and twist changes have so far been obtained. It is suggested to twist the cable before the first circuit by about 45 0 in one direction, after turning and the second circuit, to twist by about 90 0 in the other direction and to follow with a further two circuits with twisting. A suitable cable twisting device which works with discrete clamping jaw is described in detail. (UWI) [de

Mimicking lizard-like surface structures upon ultrashort laser pulse irradiation of inorganic materials

Science.gov (United States)

Hermens, U.; Kirner, S. V.; Emonts, C.; Comanns, P.; Skoulas, E.; Mimidis, A.; Mescheder, H.; Winands, K.; Krüger, J.; Stratakis, E.; Bonse, J.

2017-10-01

Inorganic materials, such as steel, were functionalized by ultrashort laser pulse irradiation (fs- to ps-range) to modify the surface's wetting behavior. The laser processing was performed by scanning the laser beam across the surface of initially polished flat sample material. A systematic experimental study of the laser processing parameters (peak fluence, scan velocity, line overlap) allowed the identification of different regimes associated with characteristic surface morphologies (laser-induced periodic surface structures, grooves, spikes, etc.). Analyses of the surface using optical as well as scanning electron microscopy revealed morphologies providing the optimum similarity to the natural skin of lizards. For mimicking skin structures of moisture-harvesting lizards towards an optimization of the surface wetting behavior, additionally a two-step laser processing strategy was established for realizing hierarchical microstructures. In this approach, micrometer-scaled capillaries (step 1) were superimposed by a laser-generated regular array of small dimples (step 2). Optical focus variation imaging measurements finally disclosed the three dimensional topography of the laser processed surfaces derived from lizard skin structures. The functionality of these surfaces was analyzed in view of wetting properties.

Effect of irradiation temperature on microstructural changes in self-ion irradiated austenitic stainless steel

Science.gov (United States)

Jin, Hyung-Ha; Ko, Eunsol; Lim, Sangyeob; Kwon, Junhyun; Shin, Chansun

2017-09-01

We investigated the microstructural and hardness changes in austenitic stainless steel after Fe ion irradiation at 400, 300, and 200 °C using transmission electron microscopy (TEM) and nanoindentation. The size of the Frank loops increased and the density decreased with increasing irradiation temperature. Radiation-induced segregation (RIS) was detected across high-angle grain boundaries, and the degree of RIS increases with increasing irradiation temperature. Ni-Si clusters were observed using high-resolution TEM in the sample irradiated at 400 °C. The results of this work are compared with the literature data of self-ion and proton irradiation at comparable temperatures and damage levels on stainless steels with a similar material composition with this study. Despite the differences in dose rate, alloy composition and incident ion energy, the irradiation temperature dependence of RIS and the size and density of radiation defects followed the same trends, and were very comparable in magnitude.

IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

International Nuclear Information System (INIS)

1997-01-01

This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-01-01

This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

Irradiation Creep in Graphite

Energy Technology Data Exchange (ETDEWEB)

Ubic, Rick; Butt, Darryl; Windes, William

2014-03-13

An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

International Nuclear Information System (INIS)

Sugimoto, Masayoshi

2001-01-01

International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

Dielectric changes in neutron-irradiated rf window materials

International Nuclear Information System (INIS)

Frost, H.M.; Clinard, F.W. Jr.

1987-01-01

Ceramics used for windows in ECRH heating systems for magnetically-confined fusion reactors must retain adequate properties during and after intense neutron irradiation. Of particular concern is a decrease in transmissivity, a parameter inversely related to the product of dielectric constant K and loss tangent tanδ. Samples of polycrystalline Al 2 O 3 and BeO were irradiated to 1 x 10 26 n/m 2 at 660K in the EBR-II fission reactor, and the above properties subsequently measured at 95 GHz. It was found that ktanδ for both materials doubled, implying a doubling of thermal stresses and a consequent reduction of time-to-failure from an assumed one year to 20 min for beryllia and 2 s for alumina. In the case of BeO, a large increase in reflectance of the incident millimeter-wave power results from dielectrically uncompensated swelling. This phenomenon could significantly degrade source performance

Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

International Nuclear Information System (INIS)

Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

2009-01-01

The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

Construction of irradiated material examination facility-basic design

International Nuclear Information System (INIS)

Ro, Seung Gy; Kim, Eun Ka; Hong, Gye Won; Herr, Young Hoi; Hong, Kwon Pyo; Lee, Myeong Han; Baik, Sang Youl; Choo, Yong Sun; Baik, Seung Je

1989-02-01

The basic design of the hot cell facility which has the main purpose of doing mechanical and physical property tests of irradiated materials, the examination process, and the annexed facility has been made. Also basic and detall designs for the underground excavation work have been performed. The project management and tasks required for the license application have been carried out in due course. The facility is expected to be completed by the end of 1992, if the budgetary support is sufficient. (Author)

Irradiation and testing of compact ignition tokamak toroidal field coil insulation materials

International Nuclear Information System (INIS)

Kanemoto, G.K.; Sherick, M.J.; Sparks, D.C.

1990-05-01

This report documents the results of an irradiation and testing program performed on behalf of Martin Marietta Energy Systems, Inc. in support of the Compact Ignition Tokamak Research and Development program. The purpose of the irradiation and testing program was to determine the effects of neutron and gamma irradiation on the mechanical and electrical properties of candidate toroidal field coil insulation materials. Insulation samples were irradiated in the Advanced Test Reactor (ATR) in a large I-hole. The insulation samples were irradiated within a lead shield to reduce exposure to gamma radiation to better approximate the desired ration of neutron to gamma exposure. Two different exposure levels were specified for the insulation samples. To accomplish this, the samples were encapsulated in two separate aluminum capsules; the capsules positioned at the ATR core mid-plane and at the top of the fueled region to take advantage of the axial cosine distribution of the neutron and gamma flux; and by varying the length of irradiation time of the two capsules. Disassembly of the irradiated capsules and testing of the insulation samples were performed at the Test Reactor Area (TRA) Hot Cell Facilities. Testing of the samples included shear compression static, shear compression fatigue, flexure static, and electrical resistance measurements

Fatigue behavior of type 316 stainless steel following neutron irradiation inducing helium

International Nuclear Information System (INIS)

Grossbeck, M.L.; Liu, K.C.

1980-01-01

Since a tokamak fusion reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially the first wall and blanket. Type 316 stainless steel in the 20% cold-worked condition has been irradiated in the HFIR in order to introduce helium as well as displacement damage. A miniature hourglass specimen was developed for the reactor irradiations and subsequent fully reversed low cycle fatigue testing. For material irradiated and tested at 430 0 C in vacuum to a damage level of 7 to 15 dpa and containing 200 to 1000 appm He, a reduction in life by a factor of 3 to 10 was observed. An attempt was made to predict irradiated fatigue life by fitting data from irradiated material to a power law equation similar to the universal slopes equation and using ductility ratios from tensile tests to modify the equation for irradiated material

Effects of antioxidant and package materials on the quality of irradiated rugao ham

International Nuclear Information System (INIS)

Cao Hong; Chen Xiulan; Bao Jianzhong; Han Yan; Jiang Yunsheng; Wang Zhijun; Dong Jie; Yang Hairong; Xi Jun

2008-01-01

Irradiation could extend the shelf life of ham, but irradiation also facilitates the oxidation of fat. Different packaging materials and combination of antioxidants were used to deal with Rugao ham in order to lower the level of antioxidation caused by irradiation treatment. The peroxide value of fat was detected as the reference index. The results were indicated that the fat peroxide value of all samples increased within the storage of 100d, and then decreased. Aluminum film compound packaging showed a better effect than polyethylene plastic bag. The antioxideant combination of 0.5% tea-polyphenol, 0.5% Vc, 0.5% citric acid, 5% sodium alginate, applied on 4 kGy irradiated samples was measured the lowest peroxide value of fat among all the treatments. (authors)

Effect of low temperature reactor irradiation on organic insulators in superconducting magnets, (4)

International Nuclear Information System (INIS)

Kato, Teruo; Takamura, Saburo

1983-01-01

In order to study effects of irradiation at low temperature on insulating materials of superconducting magnets, flexural and impact tests are carried out at 4.2K without warmup after low temperature irradiation for several fiber reinforced plastics. The used materials are glass fiber reinforced epoxies and polyimide, and carbon fiber reinforced epoxies. After irradiation of 1.1 X 10 9 rad, the reduction in flexural strength of G-10 CR is about 70% and that of G-11 CR about 25%. No change are observed in strength of glass fiber reinforced polyimide by low temperature irradiation. Other kinds of glass fiber reinforced epoxies show a reduction in strength but the flexural strength of carbon fiber reinforced epoxies increases a small by irradiation. Irradiation effect of these materials on impact value is similar to that on flexural strength. (author)

Spherical nanoindentation of proton irradiated 304 stainless steel: A comparison of small scale mechanical test techniques for measuring irradiation hardening

Science.gov (United States)

Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley; Vo, Hi T.; Maloy, Stuart A.; Hosemann, Peter; Mara, Nathan A.

2017-09-01

Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current work focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-induced increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa-30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. The disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.

Materials characterization center workshop on the irradiation effects in nuclear waste forms

International Nuclear Information System (INIS)

Roberts, F.P.; Turcotte, R.P.; Weber, W.J.

1981-01-01

The Workshop on Irradiation Effects in Nuclear Waste Forms sponsored by the Materials Characterization Center (MCC) brought together experts in radiation damage in materials and waste-management technology to review the problems associated with irradiation effects on waste-form integrity and to evaluate standard methods for generating data to be included in the Nuclear Waste Materials Handbook. The workshop reached the following conclusions: the concept of Standard Test for the Effects of Alpha-Decay in Nuclear Waste Solids, (MCC-6) for evaluating the effects of alpha decay is valid and useful, and as a result of the workshop, modifications to the proposed procedure will be incorpoated in a revised version of MCC-6; the MCC-6 test is not applicable to the evaluation of radiation damage in spent fuel; plutonium-238 is recommended as the dopant for transuranic and defense high-level waste forms, and when high doses are required, as in the case of commercial high-level waste forms, 244 Cm can be used; among the important property changes caused by irradiation are those that lead to greater leachability, and additionally, radiolysis of the leachant may increase leach rates; research is needed in this area; ionization-induced changes in physical properties can be as important as displacement damage in some materials, and a synergism is also likely to exist from the combined effects of ionization and displacement damage; and the effect of changing the temperature and dose rates on property changes induced by radiation damage needs to be determined

Small Punch Test Techniques for Irradiated Materials in Hot Cell

International Nuclear Information System (INIS)

Kim, Do Sik; Ahn, S. B.; Oh, W. H.; Yoo, B. O.; Choo, Y. S.

2006-06-01

Detailed procedures of the small punch test including the apparatus, the definition of small punch-related parameters, and the interpretation of results were presented. The testing machine should have a capability of the compressive loading and unloading at a given deflection level. The small punch specimen holder consists of an upper and lower die and clamping screws. The clamped specimen is deformed by using ball or spherical head punch. Two type of specimens with a circular and a square shape were used. The irradiated small punch specimen is made from the undamaged portion of the broken CVN bars or prepared by the irradiation of the specimen fabricated from the fresh materials. The heating and cooling devices should have the capability of the temperature control within ±2 .deg. C for the target value during the test. Based on the load-deflection data obtained from the small punch test. the empirical correlation between the small punch related parameters and a tensile properties such as 0.2% yield strength and ultimate tensile strength, fracture toughness, ductile-brittle transition temperature and creep properties determined from the standard test method is established and used to evaluate the mechanical properties of an irradiated materials. In addition, from the quantitative fractographic assessment of small punch test specimens, the relationship between the small punch energy and the quantity of ductile crack growth is obtained. Analytical formulations demonstrated good agreement with experimental load-deflection curves

Containment system of contamination in irradiated materials handling laboratories

International Nuclear Information System (INIS)

Lobao, A.S.T.; Araujo, J.A. de; Camilo, R.L.

1988-01-01

A study to prevent radiactivity release in lab scale is presented. As a basis for the design all the limits established by the IAEA for ventilation systems were observed. An evaluation of the different parameters involved in the design have been made, resulting in the specification of the working areas, ducts and filtering systems in order to get the best conditions for the safe handling of irradiated materials. (author) [pt

Temperature response of biological materials to pulsed non-ablative CO2 laser irradiation

NARCIS (Netherlands)

Brugmans, M. J.; Kemper, J.; Gijsbers, G. H.; van der Meulen, F. W.; van Gemert, M. J.

1991-01-01

This paper presents surface temperature responses of various tissue phantoms and in vitro and in vivo biological materials in air to non-ablative pulsed CO2 laser irradiation, measured with a thermocamera. We studied cooling off behavior of the materials after a laser pulse, to come to an

Simulated Irradiation of Samples in HFIR for use as Possible Test Materials in the MPEX (Material Plasma Exposure Experiment) Facility

Energy Technology Data Exchange (ETDEWEB)

Ellis, Ronald James [ORNL; Rapp, Juergen [ORNL

2014-01-01

The importance of Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) facility will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. The project presented in this paper involved performing assessments of the induced radioactivity and resulting radiation fields of a variety of potential fusion reactor materials. The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR; generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. These state-of-the-art simulation methods were used in addressing the challenge of the MPEX project to minimize the radioactive inventory in the preparation of the samples for inclusion in the MPEX facility.

Design and Demonstration of a Material-Plasma Exposure Target Station for Neutron Irradiated Samples

International Nuclear Information System (INIS)

Rapp, Juergen; Aaron, A. M.; Bell, Gary L.; Burgess, Thomas W.; Ellis, Ronald James; Giuliano, D.; Howard, R.; Kiggans, James O.; Lessard, Timothy L.; Ohriner, Evan Keith; Perkins, Dale E.; Varma, Venugopal Koikal

2015-01-01

Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma-material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panel reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a ''. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.'' The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma-material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL's proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL's strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the ''signature facility'' FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material-Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing steady-state heat fluxes of

Importance diagrams - a novel presentation of the response of a material to neutron irradiation

International Nuclear Information System (INIS)

Forrest, R.A.

1998-01-01

Activation of fusion materials following neutron irradiation is of great technological importance, especially in the study of safety and environmental impacts. Currently, activation calculations are performed for a particular neutron spectrum, appropriate to a region in a particular fusion device, which makes it difficult to extract generic information. The present work gives details of a method to present the dominant nuclides for the radiological responses or an irradiated material in a fashion that is independent of the neutron spectrum and almost independent of the flux. The importance diagrams show regions in the decay time versus neutron energy space where a nuclide contributes >50% of the response. The importance diagrams for pure iron and SS316 are described, and it is noted that the shapes of the various regions vary very little with the total neutron flux. Variation of the diagrams with irradiation time occurs at short decay times in a systematic fashion. The use of the diagrams in a realistic spectrum relies on an expansion, which while not generally true, does hold approximately for many of the nuclides of interest. The diagrams are therefore a valuable summary of the universal, device-independent, response of the materials, and when combined with pathway information give a comprehensive description of activation for that material. (orig.)

Material correlations and models for the irradiation behavior of fissile and fertile material in SNR-300, Mark-II and KNK II, third core

International Nuclear Information System (INIS)

Fenneker; Steinmetz; Toebbe

1986-07-01

The report contains the material correlations and models used in the fuel pin design code IAMBUS for the irradiation behavior of PuO 2 -UO 2 fissile materials and UO 2 fertile materials of the SNR-300 Mark-II reload and the KNK II third core. They are applicable for pellet densities of more than 90 % of the theoretical density. The presented models of the fuel behavior and the applied material correlations have been derived either from single experiments or from the comparison of theoretically predicted integral fuel behavior with the results of fuel pin irradiation experiments. The material correlations have been examined and extended in the frame of the collaborations INTERATOM/KWU and INTERATOM/KfK. French and British results were included, when available from the European fast reactor knowledge exchange [de

Development of in-pile instruments for fuel and material irradiation tests

Energy Technology Data Exchange (ETDEWEB)

Shibata, Akira; Kitagishi, Shigeru; Kimura, Nobuaki; Saito, Takashi; Nakamura, Jinichi; Ohmi, Masao; Izumo, Hironobu; Tsuchiya, Kunihiko [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

To get measurement data with high accuracy for fuel and material behavior studies in irradiation tests, two kinds of measuring equipments have been developed; these are the Electrochemical Corrosion Potential (ECP) sensor and the Linear Voltage Differential Transformer (LVDT) type gas pressure gauge. The ECP sensor has been developed to determine the corrosive potential under high temperature and high pressure water conditions. The structure of the joining parts was optimized to avoid stress concentration. The ECP sensor showed enough performance at 288degC and at 9MPa conditions. The LVDT type rod inner gas pressure gauge has been developed to measure gas pressure in a fuel element during neutron irradiation. To perform stable measurements with high accuracy under high temperature, high pressure and high dosed environment, the coil material of LVDT was changed to MI cable. As a result of this development, the LVDT type gas pressure gauge showed high accuracy within 1.8% of a full scale, and good stability. (author)

Development of in-pile instruments for fuel and material irradiation tests

International Nuclear Information System (INIS)

Shibata, Akira; Kitagishi, Shigeru; Kimura, Nobuaki; Saito, Takashi; Nakamura, Jinichi; Ohmi, Masao; Izumo, Hironobu; Tsuchiya, Kunihiko

2012-01-01

To get measurement data with high accuracy for fuel and material behavior studies in irradiation tests, two kinds of measuring equipments have been developed; these are the Electrochemical Corrosion Potential (ECP) sensor and the Linear Voltage Differential Transformer (LVDT) type gas pressure gauge. The ECP sensor has been developed to determine the corrosive potential under high temperature and high pressure water conditions. The structure of the joining parts was optimized to avoid stress concentration. The ECP sensor showed enough performance at 288degC and at 9MPa conditions. The LVDT type rod inner gas pressure gauge has been developed to measure gas pressure in a fuel element during neutron irradiation. To perform stable measurements with high accuracy under high temperature, high pressure and high dosed environment, the coil material of LVDT was changed to MI cable. As a result of this development, the LVDT type gas pressure gauge showed high accuracy within 1.8% of a full scale, and good stability. (author)

Behavior of structural and target materials irradiated in spallation neutron environments

Energy Technology Data Exchange (ETDEWEB)

Stubbins, J.F. [Univ. of Illinois, Urbana, IL (United States); Wechsler, M. [North Carolina State Univ., Raleigh, NC (United States); Borden, M. [Los Alamos National Laboratory, NM (United States)] [and others

1995-10-01

This paper describes considerations for selection of structural and target materials for accelerator-driven neutron sources. Due to the operating constraints of proposed accelerator-driven neutron sources, the criteria for selection are different than those commonly applied to fission and fusion systems. Established irradiation performance of various alloy systems is taken into account in the selection criteria. Nevertheless, only limited materials performance data are available which specifically related to neutron energy spectra anticipated for spallation sources.

Behavior of structural and target materials irradiated in spallation neutron environments

International Nuclear Information System (INIS)

Stubbins, J.F.; Wechsler, M.; Borden, M.

1995-01-01

This paper describes considerations for selection of structural and target materials for accelerator-driven neutron sources. Due to the operating constraints of proposed accelerator-driven neutron sources, the criteria for selection are different than those commonly applied to fission and fusion systems. Established irradiation performance of various alloy systems is taken into account in the selection criteria. Nevertheless, only limited materials performance data are available which specifically related to neutron energy spectra anticipated for spallation sources

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.

Facts about food irradiation: Packaging of irradiated foods

International Nuclear Information System (INIS)

1991-01-01

This fact sheet considers the effects on packaging materials of food irradiation. Extensive research has shown that almost all commonly used food packaging materials toted are suitable for use. Furthermore, many packaging materials are themselves routinely sterilized by irradiation before being used. 2 refs

Effect of ion-irradiation on the microstructure and microhardness of the W-2Y2O3 composite materials fabricated by sintering and hot forging

International Nuclear Information System (INIS)

Battabyal, M.; Spätig, P.; Baluc, N.

2013-01-01

Highlights: • W-2Y 2 O 3 material is fabricated using sintering and hot forging method with 99.3 vol.% density. • Microstructure and microhardness of the material after heavy ion irradiation are almost similar irrespective of the sample holder heating temperatures. • Dislocation loops are found on the W grains of irradiated sample where as radiation induced fine voids are observed on yttria particles. • We also observe few radiation loops on yttria particles. • No surface crack at the grain boundary is observed and significant difference in radiation hardening is confirmed. -- Abstract: A W-2Y 2 O 3 material was developed in collaboration with the Plansee Company (Austria). An ingot of the material having approximate dimension of 95 mm × 20 mm was fabricated by mixing the elemental powders followed by pressing, sintering and hot forging. The microstructure of the W-2Y 2 O 3 composite was investigated using transmission electron microscopy (TEM). The microhardness was studied using nano-indentation technique. We observed that the W-grains having a mean size of about 1 μm already formed and these grains contain very low density of dislocations. The size of the yttria particles was between 300 nm and 1 μm and the Berkovich hardness was about 4.8 GPa. The specimens were irradiated/implanted with Fe and He ions at JANNuS facility located at Orsay/Saclay, France. The TEM disks kept were irradiated/implanted at 300 and 700 °C using Fe and He ions with an energy of 24 and 2 MeV, respectively. The calculated radiation dose was about 5 dpa produced by Fe ions and total He content is 75 appm at both 300 and 700 °C. From the TEM investigation of irradiated samples, few radiation loops are present on the W grains, whereas on yttria particles, the radiation induced damages appear as voids. Berkovich hardness of the irradiated sample is higher than that of the non-irradiated sample. Results on the microstructure and microhardness of the ion-irradiated W-2Y 2 O 3

International fusion materials irradiation facility and neutronic calculations for its test modules

International Nuclear Information System (INIS)

Sokcic-Kostic, M.

1997-01-01

The International Fusion Material Irradiation Facility (IFMIF) is a projected high intensity neutron source for material testing. Neutron transport calculations for the IFMIF project are performed for variety of here explained reasons. The results of MCNP neutronic calculations for IFMIF test modules with NaK and He cooled high flux test cells are presented in this paper. (author). 3 refs., 2 figs., 3 tabs

New insight on bubble-void transition effects in irradiated materials

International Nuclear Information System (INIS)

Dubinko, V.I.

1993-01-01

An account of elastic interaction between cavities and point defects is shown to result in new critical quantities for bubblevoid transition effects in irradiated cubic crystals. In contrast to previous theories, the present one gives not only critical quantities which determine the onset of bias-driven void swelling but the maximum stationary number density and the corresponding mean radius of voids as well as the duration of the bimodal regime. The void density and swelling rate are shown to be independent from the gas level. In the region of low temperatures/high dose rates, the void density appears to be independent from irradiation parameters as well. The relationships among material constants are found at which the stabilization of gas bubbles occurs via the dislocation loop punching mechanism resulting in a drastic change in the cavity behaviour under irradiation such as the saturation (or even suppression) of void swelling and void lattice formation. The theoretical results are compared with experimental data and further experimental tests are proposed. (author). 38 refs., 1 tab., 11 figs

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

International Nuclear Information System (INIS)

Petrequin, P.

1995-01-01

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

Irradiation effects on the ductility of fusion reactor structural materials

International Nuclear Information System (INIS)

Boudamous, F.

1986-10-01

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

Improving the thermal stability and electrical parameters of a liquid crystalline material 4-n-(nonyloxy) benzoic acid by using Li ion beam irradiation

Science.gov (United States)

Kumar, Satendra; Verma, Rohit; Dwivedi, Aanchal; Dhar, R.; Tripathi, Ambuj

2018-05-01

Li ion beam irradiation studies on a liquid crystalline material 4-n-(nonyloxy) benzoic acid (NOBA) have been carried out. The material has phase sequence of I-N-SmC-Cr. Thermodynamic studies demonstrate that an irradiation fluence of 1×1013 ions-cm-2 results in the increased thermal stability of the smectic C (SmC) phase of the material. Dielectric measurements illustrate that the transverse component of the dielectric permittivity and hence the dielectric anisotropy of the material in the nematic (N) and SmC phases are increased as compared to those of the pure material due to irradiation. UV-Visible spectrum of the irradiated material shows an additional peak along with the peak of the pure material. The observed change in the thermodynamic and electrical parameters is attributed to the conversion of some of the dimers of NOBA to monomers of NOBA due to irradiation.

Investigation of special capsule technologies for material in-pile irradiation test and development plan in HANARO

Energy Technology Data Exchange (ETDEWEB)

Cho, M. S.; Son, J. M.; Kim, D. S.; Park, S. J.; Cho, Y. G.; Seo, C. K.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

2002-10-01

In-pile test for several materials such as Zr alloy, stainless steel, Cr-Ni steel etc. which are used as structural material of the advanced reactor and KNGR(Korea Next Generation Reactor) like SMART, is necessary to produce the design data for developing new reactor materials. Advanced countries like USA, Europe and Japan etc. are not only performing the simple irradiation test for materials, but developing many kinds of special capsule to perform in-pile test having special purpose. For the special test items of fuel rod, fission products, total heat generation, swelling, deformation, sweep gas, temperature ramping and BOCA etc. are being actively concerned. There are capsules measuring creep, fatigue, crack growth, and controlling fluence etc. for special irradiation test of materials. In addition, the advanced countries are developing several instrument technologies suitable for the special capsules. In HANARO, non-instrumented, instrumented material capsules and non-instrumented fuel capsule have been developed and they have been utilized in the irradiation test for users, and creep capsule loading single specimen was made and is planned to test in the reactor soon. For some forthcoming years, special capsules not only measuring creep deformation with multi-specimens, fatigue, controlling fluence but crack propagation and gas sweep considering the requirements of users will be developed in HANARO.

Radiation research of materials using irradiation capsules

International Nuclear Information System (INIS)

Chamrad, B.

1976-01-01

The methods are briefly characterized of radiation experiments on the WWR-S research reactor. The irradiation capsule installed in the reactor including the electronic instrumentation is described. Irradiated samples temperature is stabilized by an auxiliary heat source placed in the irradiation space. The electronic control equipment of the system is automated. In irradiation experiments, experimental and operating conditions are recorded by a digital measuring centre with electric typewriter and paper tape data recording and by an analog compensating recorder. The irradiation experiment control system controls irradiated sample temperature, the supply current size and the heating element temperature of the auxiliary stabilizing source, inert and technological pressures of the capsule atmosphere and the thermostat temperature of the thermocouple junctions. (O.K.)

Heavy ion irradiations on synthetic hollandite-type materials: Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16} (A=Cr, Fe, Al)

Energy Technology Data Exchange (ETDEWEB)

Tang, Ming, E-mail: mtang@lanl.gov [Materials Science & Technology Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Tumurugoti, Priyatham; Clark, Braeden; Sundaram, S.K. [Kazuo Inamori School of Engineering, The New York State College of Ceramics, Alfred University, Alfred, NY 14802 (United States); Amoroso, Jake; Marra, James [Materials Science & Technology Directorate, Savannah River National Laboratory, Aiken, SC 29808 (United States); Sun, Cheng [Materials Science & Technology Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Lu, Ping [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Wang, Yongqiang [Materials Science & Technology Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Jiang, Ying-Bing [TEM Laboratory, University of New Mexico, Albuquerque, NM 87131 (United States)

2016-07-15

The hollandite supergroup of minerals has received considerable attention as a nuclear waste form for immobilization of Cs. The radiation stability of synthetic hollandite-type compounds described generally as Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16} (A=Cr, Fe, Al) were evaluated by heavy ion (Kr) irradiations on polycrystalline single phase materials and multiphase materials incorporating the hollandite phases. Ion irradiation damage effects on these samples were examined using grazing incidence X-ray diffraction (GIXRD) and transmission electron microscopy (TEM). Single phase compounds possess tetragonal structure with space group I4/m. GIXRD and TEM observations revealed that 600 keV Kr irradiation-induced amorphization on single phase hollandites compounds occurred at a fluence between 2.5×10{sup 14} Kr/cm{sup 2} and 5×10{sup 14} Kr/cm{sup 2}. The critical amorphization fluence of single phase hollandite compounds obtained by in situ 1 MeV Kr ion irradiation was around 3.25×10{sup 14} Kr/cm{sup 2}. The hollandite phase exhibited similar amorphization susceptibility under Kr ion irradiation when incorporated into a multiphase system. - Graphical abstract: 600 keV Kr irradiation-induced amorphization on single phase hollandites compounds occurred at a fluence between 2.5×10{sup 14} Kr/cm{sup 2} and 5×10{sup 14} Kr/cm{sup 2}. The hollandite phase exhibited similar amorphization susceptibility under Kr ion irradiation when incorporated into a multiphase system. This is also the first time that the critical amorphization fluence of single phase hollandite compounds were determined at a fluence of around 3.25×10{sup 14} Kr/cm{sup 2} by in situ 1 MeV Kr ion irradiation. Display Omitted.

INAA study of Hg, Se, As, and Br irradiation losses from l-cysteine treated and untreated reference materials

International Nuclear Information System (INIS)

Anderson, D.L.

2013-01-01

U. S. Food and Drug Administration in-house reference material (RM) Cocoa Powder and National Institute of Standards and Technology Standard RMs (SRMs) 1515 apple leaves, 1547 peach leaves, 1571 orchard leaves, 1566a oyster tissue, and 1568a rice flour were co-irradiated together with polyethylene blanks and analyzed for Hg and Se by anticoincidence instrumental neutron activation analysis. The three botanical SRM portions showed a combined Hg recovery of 70 % while the other portions showed a combined Hg recovery of 169 %, indicating that volatile Hg was lost from botanical SRMs and absorbed by the other irradiated portions. Total Hg recovery for all portions was 82 %. Se results showed no evidence of cross-contamination and all results agreed with certified and known values. National Research Council of Canada Certified RMs DOLT-3 dogfish liver, TORT-2 lobster hepatopancreas, and DORM-3 fish protein were separately analyzed either with no treatment or after treatment with l-cysteine solutions followed by drying over magnesium perchlorate. Each set of portions was co-irradiated with polyethylene and treated filter blanks. Analysis of all components of each treated portion irradiation package showed that essentially all Hg was retained within the package. Treated DOLT-3 portions (inorganic Hg content 53 %) showed a tenfold improvement with 99 % Hg retention. Hg retention for DORM-3 (7 % inorganic Hg) was 85 % (a twofold improvement) while retention for TORT-2 (44 % inorganic Hg), was 94 %, similar to that for untreated portions (96 %). Small irradiation losses (≤0.5 %) of volatile species of Se, As, and Br were observed. (author)

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)

Effect of irradiation on the microbiological status and flavouring materials of selected spices

International Nuclear Information System (INIS)

Farag, S.E.D.A.; Aziz, N.H.; Attia, E.S.A.

1995-01-01

Spices from Egyptian local markets were irradiated with different recommended doses (0, 5, 10, 20 and 30 kGy). The spices tested included dried leaves of marjoram (Majorana hortensis Moench), rhizomes of ginger (Zingiber officinale Roscoe) and powdered hot pepper (Capsicum annum L.). The study included the isolation and identification of micro-organisms in spices following their irradiation, as well as gas chromatographic (GLC) chemical analysis for the presence and structure of volatile oils, pungent and pigment materials. The results showed that hot pepper was contaminated more (9.2x10 5 /g) than marjoram (4.2x10 3 /g) and ginger (14.3x10 3 /g) with respect to total aerobic bacterial content. The total contents of moulds were 4.8x10 3 /g, 5.7x10 3 /g and 19x10 3 /g in the same spices, respectively, but the pathogenic moulds and bacterial strains differed according to the type of spice. Irradiation at 10, 20 and 30 kGy caused complete elimination of mirco-organisms, whereas 5 kGy was less effective. With the GLC method chosen 18 and 50 compounds could be detected in the extracts of marjoram and ginger, respectively; γ-terpinen and zingiberen being the major compounds in marjoram and ginger, respectively. A noticeable reduction was observed in the amount of terpenes present in irradiated marjoram; they were converted to monoterpensalcohols. Ginger was more sensitive to irradiation, especially at high doses, but moderate changes were detected at low doses (5 and 10 kGy). A slight, but significant effect on the capsaicin (pungent compound) in hot pepper was observed following irradiation, whereas no changes in total pigments resulted at any dose. These results prove that 10 kGy is a sufficiently high dose to eliminate the microorganisms in spices, causing only slight changes in the flavouring materials. (orig.)

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

International Nuclear Information System (INIS)

Poirier, J.; Dupouy, J.M.

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

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

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

Defect studies in electron-irradiated ZnO and GaN

International Nuclear Information System (INIS)

Tuomisto, F.; Look, D.C.; Farlow, G.C.

2007-01-01

We present experimental results obtained with positron annihilation spectroscopy in room-temperature electron-irradiated n-type ZnO and GaN. The cation vacancies act as important compensating centers in 2 MeV electron-irradiated samples, even though their introduction rates are different by 2 orders of magnitude. In addition, negatively charged non-open volume defects that also compensate the n-type conductivity are produced together with the cation vacancies at similar introduction rates. The low introduction rates of compensating defects in ZnO demonstrate the radiation hardness of the material. Isochronal thermal annealings were performed to study the dynamics of the irradiation-induced defects. In 2 MeV electron-irradiated ZnO, all the defects introduced in the irradiation disappear already at 600 K, while 1100 K is needed in GaN. Several separate annealing stages of the defects are observed in both materials, the first at 400 K

Defect studies in electron-irradiated ZnO and GaN

Energy Technology Data Exchange (ETDEWEB)

Tuomisto, F. [Laboratory of Physics, Helsinki University of Technology, 02015 TKK Espoo (Finland)], E-mail: filip.tuomisto@tkk.fi; Look, D.C. [Semiconductor Research Center, Wright State University, Dayton, OH 45435 (United States); Materials and Manufacturing Directorate, Air Force Research Laboratory, Wright-Patterson Air Force Base, OH 45433 (United States); Farlow, G.C. [Physics Department, Wright State University, Dayton, OH 45435 (United States)

2007-12-15

We present experimental results obtained with positron annihilation spectroscopy in room-temperature electron-irradiated n-type ZnO and GaN. The cation vacancies act as important compensating centers in 2 MeV electron-irradiated samples, even though their introduction rates are different by 2 orders of magnitude. In addition, negatively charged non-open volume defects that also compensate the n-type conductivity are produced together with the cation vacancies at similar introduction rates. The low introduction rates of compensating defects in ZnO demonstrate the radiation hardness of the material. Isochronal thermal annealings were performed to study the dynamics of the irradiation-induced defects. In 2 MeV electron-irradiated ZnO, all the defects introduced in the irradiation disappear already at 600 K, while 1100 K is needed in GaN. Several separate annealing stages of the defects are observed in both materials, the first at 400 K.

Design and Demonstration of a Material-Plasma Exposure Target Station for Neutron Irradiated Samples

Energy Technology Data Exchange (ETDEWEB)

Rapp, Juergen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Aaron, A. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bell, Gary L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burgess, Thomas W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kiggans, James O. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lessard, Timothy L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ohriner, Evan Keith [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Perkins, Dale E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Varma, Venugopal Koikal [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-10-20

Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma–material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panel reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a “. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.” The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma–material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL’s proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL’s strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the “signature facility” FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material–Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing steady

Influence of specimen size/type on the fracture toughness of five irradiated RPV materials

International Nuclear Information System (INIS)

Sokolov, Mikhail A; Lucon, Enrico

2015-01-01

The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8 x 10 11 n/cm 2 /s (>1 MeV) to fluences from 0.5 to 3.4 10 19 n/cm 2 and at 288 °C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5 x 10-mm three-point bend specimens to SCK-CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes > 10 13 n/cm 2 /s and subsequent testing by SCK-CEN. The BR2 irradiations were conducted at about 2 and 4 x 10 13 n/cm 2 /s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and 10 x 10 19 n/cm 2 . The irradiation-induced shifts of the Master Curve reference temperatures, ΔT 0 , for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5 x 10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, ΔT 0 , 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ΔT 0 , were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

Ion irradiation studies of construction materials for high-power accelerators

Science.gov (United States)

Mustafin, E.; Seidl, T.; Plotnikov, A.; Strašík, I.; Pavlović, M.; Miglierini, M.; Stanćek, S.; Fertman, A.; Lanćok, A.

The paper reviews the activities and reports the current results of GSI-INTAS projects that are dealing with investigations of construction materials for high-power accelerators and their components. Three types of materials have been investigated, namely metals (stainless steel and copper), metallic glasses (Nanoperm, Finemet and Vitrovac) and organic materials (polyimide insulators and glass fiber reinforced plastics/GFRP). The materials were irradiated by different ion beams with various fluencies and energies. The influence of radiation on selected physical properties of these materials has been investigated with the aid of gamma-ray spectroscopy, transmission Mössbauer spectroscopy (TMS), conversion electrons Mössbauer spectroscopy (CEMS), optical spectroscopy (IR and UV/VIS) and other analytical methods. Some experiments were accompanied with computer simulations by FLUKA, SHIELD and SRIM codes. Validity of the codes was verified by comparison of the simulation results with experiments. After the validation, the codes were used to complete the data that could not be obtained experimentally.

Process for the irradiation of a film-like material

International Nuclear Information System (INIS)

Takimoto, Kazuo; Inoue, Takashi.

1969-01-01

Herein provided is a process for curing a polymerizable coating applied to a strip-like material by irradiating the film with high energy radiation. A plurality of rollers are arranged on both sides of the radiation path in a rectangular configuration such that only the underside of the film contacts the rollers as it is unwound in spiral fashion from a feed bobbin and rewound by a take-up bobbin located within the rectangle. The rollers are further positioned to feed the film in a direction perpendicular to the radiation beam path and to assure that successive levels of the strip superimposed while being inwardly wound are mutually parallel, uniformly spaced and adjusted to precisely intercept the radiation beam. Such an arrangement prevents a polymerizable liquid coating applied to the surface of the strip from contacting the rollers and allows effective repetitive irradiation of the strip as it passes through successive levels of the spiral before being rewound. (Owens, K. J.)

Dynamic nuclear polarization of irradiated target materials

International Nuclear Information System (INIS)

Seely, M.L.

1982-01-01

Polarized nucleon targets used in high energy physics experiments usually employ the method of dynamic nuclear polarization (DNP) to polarize the protons or deuterons in an alcohol. DNP requires the presence of paramagnetic centers, which are customarily provided by a chemical dopant. These chemically doped targets have a relatively low polarizable nucleon content and suffer from loss of polarization when subjected to high doses of ionizing radiation. If the paramagnetic centers formed when the target is irradiated can be used in the DNP process, it becomes possible to produce targets using materials which have a relatively high polarizable nucleon content, but which are not easily doped by chemical means. Furthermore, the polarization of such targets may be much more radiation resistant. Dynamic nuclear polarization in ammonia, deuterated ammonia, ammonium hydroxide, methylamine, borane ammonia, butonal, ethane and lithium borohydride has been studied. These studies were conducted at the Stanford Linear Accelerator Center using the Yale-SLAC polarized target system. Results indicate that the use of ammonia and deuterated ammonia as polarized target materials would make significant increases in polarized target performance possible

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

Nano lead oxide and epdm composite for development of polymer based radiation shielding material: Gamma irradiation and attenuation tests

Science.gov (United States)

Özdemir, T.; Güngör, A.; Akbay, I. K.; Uzun, H.; Babucçuoglu, Y.

2018-03-01

It is important to have a shielding material that is not easily breaking in order to have a robust product that guarantee the radiation protection of the patients and radiation workers especially during the medical exposure. In this study, nano sized lead oxide (PbO) particles were used, for the first time, to obtain an elastomeric composite material in which lead oxide nanoparticles, after the surface modification with silane binding agent, was used as functional material for radiation shielding. In addition, the composite material including 1%, 5%, 10%, 15% and 20% weight percent nano sized lead oxide was irradiated with doses of 81, 100 and 120 kGy up to an irradiation period of 248 days in a gamma ray source with an initial dose rate of 21.1 Gy/h. Mechanical, thermal properties of the irradiated materials were investigated using DSC, DMA, TGA and tensile testing and modifications in thermal and mechanical properties of the nano lead oxide containing composite material via gamma irradiation were reported. Moreover, effect of bismuth-III oxide addition on radiation attenuation of the composite material was investigated. Nano lead oxide and bismuth-III oxide particles were mixed with different weight ratios. Attenuation tests have been conducted to determine lead equivalent values for the developed composite material. Lead equivalent thickness values from 0.07 to 0.65 (2-6 mm sample thickness) were obtained.

Mechanical properties of organic composite materials irradiated with 2 MeV electrons

International Nuclear Information System (INIS)

Egusa, S.; Kirk, M.A.; Birtcher, R.C.; Argonne National Lab., IL; Hagiwara, M.; Kawanishi, S.

1983-01-01

Four kinds of cloth-filled organic composites (filter: glass or carbon fiber; matrix; epoxy or polyimide resin) were irradiated with 2 MeV electrons at room temperature, and were examined with regard to the mechanical properties. Following irradiation the Young's (tensile) modulus of these composites remains practically unchanged even after irradiation up to 15.000 Mrad. The shear modulus and the ultimate strength, on the other hand, begin to decrease after the absorbed dose reaches about 2.000 Mrad for the glass/epoxy composite and about 5.000-10.000 Mrad for the other composites. This result is ascribed to the decrease in the capacity of load transfer from the matrix to the fiber due to the radiation damage at the interface, and the dose dependence is interpreted and formulated based on the mechanics of composite materials and the target theory used in radiation biology. As to the fracture behavior, the propagation energy increases from the beginning of irradiation. This result is attributed to the radiation-induced decrease in the bonding energy at the interface. (orig.)

First results of the post-irradiation examination of the Ceramic Breeder materials from the Pebble Bed Assemblies Irradiation for the HCPB Blanket concept

International Nuclear Information System (INIS)

Hegeman, J.; Magielsen, A.J.; Peeters, M.; Stijkel, M.P.; Fokkens, J.H.; Laan, J.G. van der

2006-01-01

In the framework of developing the European Helium Cooled Pebble-Bed (HCPB) blanket an irradiation test of pebble-bed assemblies is performed in the HFR Petten. The experiment is focused on the thermo-mechanical behavior of the HCPB type breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. To achieve representative conditions a section of the HCPB is simulated by EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. Floating Eurofer-97 steel plates separate the pebble-beds. The structural integrity of the ceramic breeder materials is an issue for the design of the Helium Cooled Pebble Bed concept. Therefore the objective of the post irradiation examination is to study deformation of pebbles and the pebble beds and to investigate the microstructure of the ceramic pebbles from the Pebble Bed Assemblies. This paper concentrates on the Post Irradiation Examination (PIE) of the four ceramic pebble beds that have been irradiated in the Pebble Bed Assembly experiment for the HCPB blanket concept. Two assemblies with Li 4 SiO 4 pebble-beds are operated at different maximum temperatures of approximately 600 o C and 800 o C. Post irradiation computational analysis has shown that both have different creep deformation. Two other assemblies have been loaded with a ceramic breeder bed of two types of Li 2 TiO 3 beds having different sintering temperatures and consequently different creep behavior. The irradiation maximum temperature of the Li 2 TiO 3 was 800 o C. To support the first PIE result, the post irradiation thermal analysis will be discussed because thermal gradients have influence on the pebble-bed thermo-mechanical behavior and as a result it may have impact on the structural integrity of the ceramic breeder materials. (author)

Control of helium effects in irradiated materials based on theory and experiment

International Nuclear Information System (INIS)

Mansur, L.K.; Lee, E.H.; Maziasz, P.J.; Rowcliffe, A.F.

1986-01-01

Helium produced in materials by (n,α) transmutation reactions during neutron irradiations or subjected in ion bombardment experiments causes substantial changes in the response to displacement damage. In particular, swelling, phase transformations and embrittlement are strongly affected. Present understanding of the mechanisms underlying these effects is reviewed. Key theoretical relationships describing helium effects on swelling and helium diffusion are described. Experimental data in the areas of helium effects on swelling and precipitation is reviewed with emphasis on critical experiments that have been designed and evaluated in conjunction with theory. Confirmed principles for alloy design to control irradiation performance are described

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

Self-shielding and burn-out effects in the irradiation of strongly-neutron-absorbing material

International Nuclear Information System (INIS)

Sekine, T.; Baba, H.

1978-01-01

Self-shielding and burn-out effects are discussed in the evaluation of radioisotopes formed by neutron irradiation of a strongly-neutron-absorbing material. A method of the evaluation of such effects is developed both for thermal and epithermal neutrons. Gadolinium oxide uniformly mixed with graphite powder was irradiated by reactor-neutrons together with pieces of a Co-Al alloy wire (the content of Co being 0.475%) as the neutron flux monitor. The configuration of the samples and flux monitors in each of two irradiations is illustrated. The yields of activities produced in the irradiated samples were determined by the γ-spectrometry with a Ge(Li) detector of a relative detection efficiency of 8%. Activities at the end of irradiation were estimated by corrections due to pile-up, self-absorption, detection efficiency, branching ratio, and decay of the activity. Results of the calculation are discussed in comparison with the observed yields of 153 Gd, 160 Tb, and 161 Tb for the case of neutron irradiation of disc-shaped targets of gadolinium oxide. (T.G.)

Effect of low temperature neutron irradiation on the magnetoresistivity in stabilizer materials for a superconducting magnet

International Nuclear Information System (INIS)

Nakata, Kiyotomo; Tada, Naobumi; Masaoka, Isao; Takamura, Saburo.

1985-01-01

Magnetoresistivity changes caused by neutron irradiation at 5 K, annealing up to 300 K and cyclic irradiation are studied in copper and aluminuim stabilizer materials at 4.2 K. The radiation-induced resistivity in Al is about three times as large as that in Cu, and the resistivities in both Al and Cu are independent of the purity and the degree of cold-work of the samples. The radiation-induced magnetoresistivity of the high purity Cu with R.R.R. (R sub(298 K)/R sub(4.2 K)) of 1400 is larger than that of the impure Cu with R.R.R. of 300 and 280. The magnetoresistivities of the high purity Cu and Al with R.R.R. of 1500 increase with the magetic field. Magnetoresistivity change with the magnetic field in the irradiated Cu mostly follows Kohler's rule, and that in the irradiated Al does not follow the rule at high magnetic fields. By the annealing at 300 K after the irradiation, the radiation-induced resistivity is completely annihilated in the Al, but about 20 % of the resistivity retains in the full-annealed Cu and the retained resistivity is accumulated during the cyclic irradiation. Though the accumulated resistivity in the cold-worked Cu is smaller than that in the full-annealed one, the resistivity before irradiation in the cold-worked samples is very large. From the above results, the full-annealed Cu with R.R.R. of about 300 is considered to be the best material as a stabilizer used under irradiation. (author)

Materials Characterization Center. Second workshop on irradiation effects in nuclear waste forms. Summary report

International Nuclear Information System (INIS)

Weber, W.J.; Turcotte, R.P.

1982-01-01

The purpose of this second workshop on irradiations effects was to continue the discussions initiated at the first workshop and to obtain guidance for the Materials Characterization Center in developing test methods. The following major conclusions were reached: Ion or neutron irradiations are not substitutes for the actinide-doping technique, as described by the MCC-6 Method for Preparation and Characterization of Actinide-Doped Waste Forms, in the final evaluation of any waste form with respect to the radiation effects from actinide decay. Ion or neutron irradiations may be useful for screening tests or more fundamental studies. The use of these simulation techniques as screening tests for actinide decay requires that a correlation between ion or neutron irradiations and actinide decay be established. Such a correlation has not yet been established and experimental programs in this area are highly recommended. There is a need for more fundamental studies on dose-rate effects, temperature dependence, and the nature and importance of alpha-particle effects relative to the recoil nucleus in actinide decay. There are insufficient data presently available to evaluate the potential for damage from ionizing radiation in nuclear waste forms. No additional test methods were recommended for using ion or neutron irradiations to simulate actinide decay or for testing ionization damage in nuclear waste forms. It was recognized that additional test methods may be required and developed as more data become available. An American Society for Testing and Materials (ASTM) Task Group on the Simulation of Radiation Effects in Nuclear Waste Forms (E 10.08.03) was organized to act as a continuing vehicle for discussions and development of procedures, particularly with regard to ion irradiations

Cesium glass irradiation sources

International Nuclear Information System (INIS)

Plodinec, M.J.

1982-01-01

The precipitation process for the decontamination of soluble SRP wastes produces a material whose radioactivity is dominated by 137 Cs. Potentially, this material could be vitrified to produce irradiation sources similar to the Hanford CsCl sources. In this report, process steps necessary for the production of cesium glass irradiation sources (CGS), and the nature of the sources produced, are examined. Three options are considered in detail: direct vitrification of precipitation process waste; direct vitrification of this waste after organic destruction; and vitrification of cesium separated from the precipitation process waste. Direct vitrification is compatible with DWPF equipment, but process rates may be limited by high levels of combustible materials in the off-gas. Organic destruction would allow more rapid processing. In both cases, the source produced has a dose rate of 2 x 10 4 rads/hr at the surface. Cesium separation produces a source with a dose rate of 4 x 10 5 at the surface, which is nearer that of the Hanford sources (2 x 10 6 rads/hr). Additional processing steps would be required, as well as R and D to demonstrate that DWPF equipment is compatible with this intensely radioactive material

Development for advanced materials and testing techniques

Energy Technology Data Exchange (ETDEWEB)

Hishinuma, Akimichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Recent studies using a JMTR and research reactors of JRR-2 and JRR-3 are briefly summarized. Small specimen testing techniques (SSTT) required for an effective use of irradiation volume and also irradiated specimens have been developed focussing on tensile test, fatigue test, Charpy test and small punch test. By using the small specimens of 0.1 - several mm in size, similar values of tensile and fatigue properties to those by standard size specimens can be taken, although the ductile-brittle transition temperature (DBTT) depends strongly on Charpy specimen size. As for advanced material development, R and D about low activation ferritic steels have been done to investigate irradiation response. The low activation ferritic steel, so-called F82H jointly-developed by JAERI and NKK for fusion, has been confirmed to have good irradiation resistance within a limited dose and now selected as a standard material in the fusion material community. It is also found that TiAi intermetallic compounds, which never been considered for nuclear application in the past, have an excellent irradiation resistance under an irradiation condition. Such knowledge can bring about a large expectation for developing advanced nuclear materials. (author)

Optical transmittance investigation of 1-keV ion-irradiated sapphire crystals as potential VUV to NIR window materials of fusion reactors

Directory of Open Access Journals (Sweden)

Keisuke Iwano

2016-10-01

Full Text Available We investigate the optical transmittances of ion-irradiated sapphire crystals as potential vacuum ultraviolet (VUV to near-infrared (NIR window materials of fusion reactors. Under potential conditions in fusion reactors, sapphire crystals are irradiated with hydrogen (H, deuterium (D, and helium (He ions with 1-keV energy and ∼ 1020-m-2 s-1 flux. Ion irradiation decreases the transmittances from 140 to 260 nm but hardly affects the transmittances from 300 to 1500 nm. H-ion and D-ion irradiation causes optical absorptions near 210 and 260 nm associated with an F-center and an F+-center, respectively. These F-type centers are classified as Schottky defects that can be removed through annealing above 1000 K. In contrast, He-ion irradiation does not cause optical absorptions above 200 nm because He-ions cannot be incorporated in the crystal lattice due to the large ionic radius of He-ions. Moreover, the significant decrease in transmittance of the ion-irradiated sapphire crystals from 140 to 180 nm is related to the light scattering on the crystal surface. Similar to diamond polishing, ion irradiation modifies the crystal surface thereby affecting the optical properties especially at shorter wavelengths. Although the transmittances in the VUV wavelengths decrease after ion irradiation, the transmittances can be improved through annealing above 1000 K. With an optical transmittance in the VUV region that can recover through simple annealing and with a high transparency from the ultraviolet (UV to the NIR region, sapphire crystals can therefore be used as good optical windows inside modern fusion power reactors in terms of light particle loadings of hydrogen isotopes and helium.

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

Irradiation growth in zirconium alloys: a review

International Nuclear Information System (INIS)

Fidleris, V.

1980-09-01

The change in shape during irradiation without external stress, irradiation growth, was first discovered in uranium and later in graphite, zirconium and other core materials which exhibit anisotropic physical properties. The direction of maximum growth of metals invariably corresponds with the direction of minimum thermal expansion. In polycrystalline zirconium alloys growth is positive in the direction of maximum deformation during fabrication and in other directions it can be either positive or negative depending on the preferred orientation of grains (crystallographic texture). Growth increases gradually with temperature between 300 K and 620 K and rapidly with fluence up to about 1 x 10 25 n.m. -2 (Eμ1 MeV). At higher fluences the growth appears to saturate in annealed materials and reach a steady rate approximately proportional to dislocation density in cold-worked materials. Above 600 K both annealed and cold-worked materials have similar steady growth rates. Irradiation growth is caused by the segregation to different sinks of the vacancies and interstitials generated by irradiation, but the dominant types of sinks for each type of point defect and the mode of transport of the point defects to sinks cannot therefore be predicted theoretically. For the purpose of designing reactor core components empirical equations have been derived that can satisfactorily predict the steady state growth behaviour from texture and microstructure. (auth)

Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

Energy Technology Data Exchange (ETDEWEB)

Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: luigi.debarberis@cec.eu.int; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)

2006-04-15

It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

The Influence of Irradiation Regimes on Retention Hydrogen Isotopes in Structural Materials

International Nuclear Information System (INIS)

Zaluzhnyi, A.

2007-01-01

Full text of publication follows: In the present work was investigated the influence of irradiation regimes on retention hydrogen isotopes in samples of austenitic steel during heating. The samples of studied materials were irradiated both in the reactor and by hydrogen isotopes ions of different energies and fluencies bombardment in an accelerator. Kinetic of hydrogen release from the samples worked with deuterium plasma was investigated. The following results were obtained. Heating the irradiate d samples of steel (irradiated in the reactor or by hydrogen isotopes ions bombardment), which have been kept in normal temperature during quite a long period after the irradiation, a shift of the diffusion peak of hydrogen release to higher temperatures, comparing to no irradiated samples, was observed. It means that atoms of hydrogen in the irradiated sample were caught by radiation defects, which are very effective as traps for hydrogen atoms till quite high temperatures (700 K). The worked out analysis of the received results supposes that vacancy complexes. On thermodesorption curves of hydrogen release from irradiated samples of austenitic steels a high temperature peak (900-1000 K) was observed because of dissociation of hydrogen containing compounds in micro pores. During investigations of hydrogen release from irradiated samples of austenitic steel, after it had been saturated with hydrogen plasma, abnormally big blisters were registered with cover thickness of about 1 mkm. Three peaks were observed on the thermodesorption curves of hydrogen release from irradiated samples, contained blisters. The low temperature spike (∼500 K) was showed to correspond to hydrogen release because of its resolution from blisters, where it was in molecular form. The high temperature peak (∼900 K) corresponds to hydrogen release from dissociating blisters, which contain hydrocarbons. The mechanism of abnormal blisters generation is offered. Inasmuch methane is not soluble in

Direct or indirect UV-Irradiation in insect killers and similar equipment

International Nuclear Information System (INIS)

Heinz, G.

1978-01-01

UV sterilisation equipment is used in the refrigerating and storage rooms of food and meat processing factories. The UV radiation used has a wavelength of 254 nm. Due to possible side effects on the food, the packing is directly irradiated but the food itself only indirectly. This method makes high irradiation energies necessary for the inhibition of bacterial growth on the foodstuffs. (AJ) 891 AJ [de

Effect of using type A radiation for dose reconstruction in type B irradiated material: A microdosimetry approach

International Nuclear Information System (INIS)

Piters, T.M.; Chernov, V.

2008-01-01

A model is proposed to explain that in previously γ irradiated calcite, the yield after additive β irradiation tends to incline to the saturation yield of the β radiation even if that yield is lower than the yield after the γ irradiation. However, the proposed model is not specific for calcite and in fact all calculations are done in a fictive material. The proposed model considers, in contrast to existing models, the track nature of γ and β radiations and that these different types of radiations can be distinguished by the dose distribution inside their tracks. The determination of the dose distribution in the tracks for the different types of irradiations is quite complicated and instead we approximate the γ and β tracks by type A and B tracks that have different but homogeneously distributed dose in their track volumes. The trapping of generated free charges in the track was calculated with a simple one electron-one hole trap model. To obtain the total dose response (the average concentration of occupied traps as a function of dose), the yield in one point was averaged over all possible configurations of track overlapping in that point. We determined the slope of the initial part of the response curve (low dose sensitivity) and the saturation yield as function of the track dose. It is observed that the low dose sensitivity and saturation yield both decrease with increasing track dose. Simulations of the response to sequential irradiation first by type A radiation with a 64 Gy track dose and then followed by type B radiation with a track dose of 128 Gy using our model show a similar effect as observed in calcite demonstrating that the track nature of radiation is a plausible cause for the observed effect

Evaluation of irradiation hardening of proton irradiated stainless steels by nanoindentation

International Nuclear Information System (INIS)

Yabuuchi, Kiyohiro; Kuribayashi, Yutaka; Nogami, Shuhei; Kasada, Ryuta; Hasegawa, Akira

2014-01-01

Ion irradiation experiments are useful for investigating irradiation damage. However, estimating the irradiation hardening of ion-irradiated materials is challenging because of the shallow damage induced region. Therefore, the purpose of this study is to prove usefulness of nanoindentation technique for estimation of irradiation hardening for ion-irradiated materials. SUS316L austenitic stainless steel was used and it was irradiated by 1 MeV H + ions to a nominal displacement damage of 0.1, 0.3, 1, and 8 dpa at 573 K. The irradiation hardness of the irradiated specimens were measured and analyzed by Nix–Gao model. The indentation size effect was observed in both unirradiated and irradiated specimens. The hardness of the irradiated specimens changed significantly at certain indentation depths. The depth at which the hardness varied indicated that the region deformed by the indenter had reached the boundary between the irradiated and unirradiated regions. The hardness of the irradiated region was proportional to the inverse of the indentation depth in the Nix–Gao plot. The bulk hardness of the irradiated region, H 0 , estimated by the Nix–Gao plot and Vickers hardness were found to be related to each other, and the relationship could be described by the equation, HV = 0.76H 0 . Thus, the nanoindentation technique demonstrated in this study is valuable for measuring irradiation hardening in ion-irradiated materials

General corrosion, irradiation-corrosion, and environmental-mechanical evaluation of nuclear-waste-package structural-barrier materials. Progress report

International Nuclear Information System (INIS)

Westerman, R.E.; Pitman, S.G.; Nelson, J.L.

1982-09-01

Pacific Northwest Laboratory is studying the general corrosion, irradiation-corrosion, and environmentally enhanced crack propagation of five candidate materials in high-temperature aqueous environments simulating those expected in basalt and tuff repositories. The materials include three cast ferrous materials (ductile cast iron and two low-alloy Cr-Mo cast steels) and two titanium alloys, titanium Grade 2 (commercial purity) and Grade 12 (a Ti-Ni-Mo alloy). The general corrosion results are being obtained by autoclave exposure of specimens to slowly replenished simulated ground water flowing upward through a bed of the appropriate crushed rock (basalt or tuff), which is maintained at the desired test temperature (usually 250 0 C). In addition, tests are being performed in deionized water. Metal penetration rates of iron-base alloys are being derived by stripping off the corrosion product film and weighing the specimen after the appropriate exposure time. The corrosion of titanium alloy specimens is being determined by weight gain methods. The irradiation-corrosion studies are similar to the general corrosion tests, except that the specimen-bearing autoclaves are held in a 60 Co gamma radiation field at dose rates up to 2 x 10 6 rad/h. For evaluating the resistance of the candidate materials to environmentally enhanced crack propagation, three methods are being used: U-bend and fracture toughness specimens exposed in autoclaves; slow strain rate studies in repository-relevant environments to 300 0 C; and fatigue crack growth rate studies at ambient pressure and 90 0 C. The preliminary data suggest a 1-in. corrosion allowance for iron-base barrier elements intended for 1000-yr service in basalt or tuff repositories. No evidence has yet been found that titanium Grade 2 or Grade 12 is susceptible to environmentally induced crack propagation or, by extension, to stress corrosion cracking

H2 formation by electron irradiation of SBA-15 materials and the effect of Cu(II) grafting

International Nuclear Information System (INIS)

Brodie-Linder, N.; Le Caer, S.; Shahdo Alam, M.; Renault, J.P.; Alba-Simionesco, Ch.

2010-01-01

Measurement of H 2 production from electron irradiation (10 MeV) on SBA-15 materials has shown that adsorbed water is attacked preferentially. Silanol groups are only attacked when they are in the majority with respect to adsorbed water, however they are much less efficient at producing H 2 . The comparison between water content before and after electron irradiation and the corresponding H 2 production indicates that water desorption is the main route to adsorbed water loss for SBA-15 materials. On the other hand, surface silanol groups are more susceptible to attack,leading to H 2 production when SBA-15 samples have undergone extensive thermal treatment. Electron irradiation of SBA-15-Cu materials has shown that the presence of Cu(II) on the surface reduces and inhibits the production of H 2 . This inhibiting power affects adsorbed water bonded to grafted copper but not surface silanol groups. (authors)

Laser irradiation of carbon–tungsten materials

International Nuclear Information System (INIS)

Marcu, A; Lungu, C P; Ursescu, D; Porosnicu, C; Grigoriu, C; Avotina, L; Kizane, G; Marin, A; Osiceanu, P; Grigorescu, C E A; Demitri, N

2014-01-01

Carbon–tungsten layers deposited on graphite by thermionic vacuum arc (TVA) were directly irradiated with a femtosecond terawatt laser. The morphological and structural changes produced in the irradiated area by different numbers of pulses were systematically explored, both along the spots and in their depths. Although micro-Raman and Synchrotron-x-ray diffraction investigations have shown no carbide formation, they have shown the unexpected presence of embedded nano-diamonds in the areas irradiated with high fluencies. Scanning electron microscopy images show a cumulative effect of the laser pulses on the morphology through the ablation process. The micro-Raman spatial mapping signalled an increased percentage of sp 3 carbon bonding in the areas irradiated with laser fluencies around the ablation threshold. In-depth x-ray photoelectron spectroscopy investigations suggested a weak cumulative effect on the percentage increase of the sp 2 -sp 3 transitions with the number of laser pulses just for nanometric layer thicknesses. (paper)

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Evidence of different red emissions in irradiated germanosilicate materials

Energy Technology Data Exchange (ETDEWEB)

Alessi, A., E-mail: antonino.alessi@univ-st-etienne.fr [Univ-Lyon, Laboratoire H. Curien, UMR CNRS 5516, Université Jean Monnet, 18 rue du Pr. Benoît Lauras, 42000 Saint-Etienne (France); Di Francesca, D. [Univ-Lyon, Laboratoire H. Curien, UMR CNRS 5516, Université Jean Monnet, 18 rue du Pr. Benoît Lauras, 42000 Saint-Etienne (France); Agnello, S. [Dipartimento di Fisica e Chimica, Università di Palermo, I-90123 Palermo (Italy); Girard, S. [Univ-Lyon, Laboratoire H. Curien, UMR CNRS 5516, Université Jean Monnet, 18 rue du Pr. Benoît Lauras, 42000 Saint-Etienne (France); Cannas, M. [Dipartimento di Fisica e Chimica, Università di Palermo, I-90123 Palermo (Italy); Richard, N. [CEA, DAM, DIF, F91297 Arpajon (France); Boukenter, A.; Ouerdane, Y. [Univ-Lyon, Laboratoire H. Curien, UMR CNRS 5516, Université Jean Monnet, 18 rue du Pr. Benoît Lauras, 42000 Saint-Etienne (France)

2016-09-15

This experimental investigation is focused on a radiation induced red emission in Ge doped silica materials, elaborated with different methods and processes. The differently irradiated samples as well as the pristine ones were analyzed with various spectroscopic techniques, such as confocal microscopy luminescence (CML), time resolved luminescence (TRL), photoluminescence excitation (PLE) and electron paramagnetic resonance (EPR). Our data prove that irradiation induces a red luminescence related to the presence of the Ge atoms. Such emission features a photoexcitation spectrum in the UV-blue spectral range and, TRL measurements show that its decrease differs from a single exponential law with a lifetime of tens of nanoseconds. CML measurements under laser at 633 nm evidenced the lack of correlation of the emission here reported with that of the Ge- or Si- non bridging oxygen hole centers. Moreover, our EPR experiments highlighted the lack of correlation between the red emitting defect with other radiation induced paramagnetic centers such as the E′Ge and Ge(2). The relation of the investigated emission with the H(II) defects, previously considered as responsible for a red emission, can not be totally excluded. - Highlights: • Composite nature of the red emission in Ge-doped doped silica materials. • Experimental study with various spectroscopic techniques and on different samples. • Time resolved and stationary characterization of an new red emission. • Study of the spatial distributions of diverse red emissions in optical fibers.

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

Irradiation-induced modification of the material parameters in magnesium-doped lithium niobate

International Nuclear Information System (INIS)

Jentjens, Lena

2010-01-01

In the framework of this thesis the material properties of lithium niobate are directedly influenced by the irradiation with 3 He ions with an energy of 40 MeV. In the first part the irradiation-induced material changes are intensively studied. Long-time stable changes of the refractive index are measured in the range of up to 6.10 -3 , which depend on the radiation dose and exhibit until now no saturation behaviour. Accompanied is this change by an also dose-dependent deformation as well as a brownish change of color of the crystals. Furthermore a by several orders of magnitude increased electrical dark- and photoconductivity, which depends on the ion dose and exhibits until now also no saturation behaviour. An effect independent on the ion dose is the reduction of the coercive field strength by about 10%. Furthermore it was stated the quantity of the effects not only depends on the absolute dose, but also on the irradiation direction in view of the crystallographic c-axis. The second part of this thesis deals with the generation of microscopic structures in lithium niobate. By an ion microbeam respectively a shiftable slit aperture the fabrication of refractive-index gratings is pursued. Grating with periodicity lengths in the range of 12-160 μm could until now be detected and promise in comparison with photorefractive gratings the advance of larger stability.

Effect of gamma rays on crystalline materials during irradiation in a reactor

International Nuclear Information System (INIS)

Nikolaenko, V.A.; Karpukhin, V.I.; Gordeev, V.G.

1995-01-01

The article presents and discusses the results of experiments to determine the effect of gamma rays on the change in the properties of diamond, graphite, and structural steel. The materials were irradiated in a VVER type reactor. For diamonds, the effect on the annealing of defects was investigated. As gamma ray intensity increased, the crystal lattice expansion and defect concentration increased. Graphite lattice expansion and the mechanical properties of structural steel were also examined. Graphite lattice expansion increased with increased neutron flux and decreased irradiation temperature. Changes in the impact toughness of structural steel correlated precisely to the gamma ray flux in the experiments. 6 refs., 3 figs

Effect of irradiation on the microbiological status and flavouring materials of selected spices

Energy Technology Data Exchange (ETDEWEB)

Farag, S.E.D.A. [National Centre for Radiation Research and Technology, Cairo (Egypt); Aziz, N.H. [National Centre for Radiation Research and Technology, Cairo (Egypt); Attia, E.S.A. [Egyptian Starch and Glucose Manufacturing Co., Cairo (Egypt)

1995-12-31

Spices from Egyptian local markets were irradiated with different recommended doses (0, 5, 10, 20 and 30 kGy). The spices tested included dried leaves of marjoram (Majorana hortensis Moench), rhizomes of ginger (Zingiber officinale Roscoe) and powdered hot pepper (Capsicum annum L.). The study included the isolation and identification of micro-organisms in spices following their irradiation, as well as gas chromatographic (GLC) chemical analysis for the presence and structure of volatile oils, pungent and pigment materials. The results showed that hot pepper was contaminated more (9.2x10{sup 5}/g) than marjoram (4.2x10{sup 3}/g) and ginger (14.3x10{sup 3}/g) with respect to total aerobic bacterial content. The total contents of moulds were 4.8x10{sup 3}/g, 5.7x10{sup 3}/g and 19x10{sup 3}/g in the same spices, respectively, but the pathogenic moulds and bacterial strains differed according to the type of spice. Irradiation at 10, 20 and 30 kGy caused complete elimination of mirco-organisms, whereas 5 kGy was less effective. With the GLC method chosen 18 and 50 compounds could be detected in the extracts of marjoram and ginger, respectively; {gamma}-terpinen and zingiberen being the major compounds in marjoram and ginger, respectively. A noticeable reduction was observed in the amount of terpenes present in irradiated marjoram; they were converted to monoterpensalcohols. Ginger was more sensitive to irradiation, especially at high doses, but moderate changes were detected at low doses (5 and 10 kGy). A slight, but significant effect on the capsaicin (pungent compound) in hot pepper was observed following irradiation, whereas no changes in total pigments resulted at any dose. These results prove that 10 kGy is a sufficiently high dose to eliminate the microorganisms in spices, causing only slight changes in the flavouring materials. (orig.)

Thermal and irradiation effects on high-temperature mechanical properties of materials for SCWR fuel cladding

International Nuclear Information System (INIS)

Kano, F.; Tsuchiya, Y.; Oka, K.

2009-01-01

The thermal and irradiation effects on high-temperature mechanical properties are examined for candidate alloys for fuel cladding of supercritical water-cooled reactors (SCRWs). JMTR (Japan Materials Testing Reactor) and Experimental Fast Reactor JOYO were utilized for neutron irradiation tests, considering their fluence and temperature. Irradiation was performed with JMTR at 600degC up to 4x10 24 n/m 2 and with JOYO at 600degC and 700degC up to 6x10 25 n/m 2 . Tensile test, creep test and hardness measurement were carried out for high-temperature mechanical properties. Based on the uniaxial creep test, the extrapolation curves were drawn with time-temperature relationships utilizing the Larson and Miller Parameter. Several candidate alloys are expected to satisfy the design requirement from the estimation of the creep rupture stress for 50000 hours. Comparing the creep strengths under irradiated and unirradiated conditions, it was inferred that creep deformation was dominated by the thermal effect rather than the irradiation at SCWR core condition. The microstructure was examined using transmission electron microscope (TEM) analysis, focusing on void swelling and helium (He) bubble formation. Void formation was observed in the materials irradiated with JOYO at 600degC but not at 700degC. However, its effect on the deformation of components was estimated to be tolerable since their size and density were negligibly small. The manufacturability of the thin-wall, small-diameter tube was confirmed for the potential candidate alloys through the trial tests in the factory where the fuel cladding tube is manufactured. (author)

EFFECT OF IRRADIATION AND PACKAGING MATERIALS TYPES ON SHELF-LIFE AND QUALITY ATTRIBUTES OF MINCED MEAT DURING COLD STORAGE

International Nuclear Information System (INIS)

OSHEBA, A.S.; NAGY, KH.S.; ANWAR, M.M.

2008-01-01

Minced meat is considered one of the most meat products that exposed to contamination which led to many changes in its quality and reduced its shelf-life.Therefore, this investigation was carried out to extend the shelf-life of minced meat for consumption and maintaining its quality during cold storage by using irradiation with various doses (3, 6 and 9 kGy) and different packing materials. The results indicated that irradiation,especially at 3 and 6 kGy, had no effect on chemical composition and some physical properties of minced meat. On the other hand, pH values of all irradiated samples were slightly decreased with decreasing irradiation doses.Irradiation at the highest dose used, i.e. 9 kGy, slightly increased total volatile nitrogen (TVN) of minced meat. Thiobarbituric acid (TBA) value of irradiated samples was tended to increase with increasing irradiation dose from 3 to 9 kGy either directly after irradiation or during storage.Regardless of irradiation effect on TVN and TBA values at zero time, there were no marked differences in TVN and TBA values of irradiated minced meat according to differentiate packaging materials (PE, PA/PE and PET/Al/PE). During cold storage, the TVN and TBA values of all minced meat samples either non-irradiated or irradiated were progressively increased as the time of cold storage increased. The higher increasing rate in TVN and TBA of irradiated samples was recorded for samples packaged in PE (one layer) followed by PA/PE (two layers) and finally PET/Al/PE (three layers) at the same irradiation dose. Irradiation of minced meat with 3 kGy reduced the counts of total bacteria, coliform bacteria, Staphylococcus aureus and yeasts and molds counts as well as eliminating Salmonella spp. Irradiation doses of 6 and 9 kGy completely eliminated coliform bacteria, Staphylococcus aureus and yeasts and molds. Also, type of packaging materials which used had no effect on counts of all studied microorganisms. Irradiation of minced meat with

Positron lifetime study of copper irradiated by energetic protons or energetic neutrons

International Nuclear Information System (INIS)

Howell, R.H.

1979-03-01

Positron lifetime measurements of pure copper damaged by irradiation with energetic protons and neutrons are presented. Lifetime determinations of the bulk material and various traps were made, and the dependence of the trapping rate on dose and irradiation energy were investigated. The results from the neutron- and proton-irradiated samples point to the existence of traps with similar but distinct lifetime parameters, not varying greatly from values reported in deformation studies. Also, a trap with long lifetime is seen for some proton irradiations, but is never seen for the neutron irradiations. The trapping rate of the short-lifetime trap is a linear function of dose for proton-irradiated samples and nearly so for the neutron irradiation. 1 figure

Influence of specimen size/type on the fracture toughness of five irradiated RPV materials

Energy Technology Data Exchange (ETDEWEB)

Sokolov, Mikhail A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lucon, Enrico [National Inst. of Standards and Technology (NIST), Boulder, CO (United States)

2015-01-01

The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8 x 10¹¹ n/cm²/s (>1 MeV) to fluences from 0.5 to 3.4 10¹⁹ n/cm² and at 288 °C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5 x 10-mm three-point bend specimens to SCK-CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes > 10¹³ n/cm²/s and subsequent testing by SCK-CEN. The BR2 irradiations were conducted at about 2 and 4 x 10¹³ n/cm²/s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and 10 x 10¹⁹n/cm². The irradiation-induced shifts of the Master Curve reference temperatures, ΔT₀, for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5 x 10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, T₀, 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ΔT₀, were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

HiRadMat: A high‐energy, pulsed beam, material irradiation facility

CERN Multimedia

Charitonidis, Nikolaos

2016-01-01

HiRadMat is a facility constructed in 2011, designed to provide high-intensity pulsed beams to an irradiation area where different material samples or accelerator components can be tested. The facility, located at the CERN SPS accelerator complex, uses a 440 GeV proton beam with a pulse length up to 7.2 μs and a maximum intensity up to 1E13 protons / pulse. The facility, a unique place for performing state-of-the art beam-to-material experiments, operates under transnational access and welcomes and financially supports, under certain conditions, experimental teams to perform their experiments.

Mechanical-property changes of structural composite materials after low-temperature proton irradiation: Implications for use in SSC magnet systems

International Nuclear Information System (INIS)

Morena, J.; Snead, C.L. Jr.; Czajkowski, C.; Skaritka, J.

1993-01-01

Longterm physical, mechanical, electrical, and other properties of advanced composites, plastics, and other polymer materials are greatly affected by high-energy proton, neutron, electron, and gamma radiation. The effects of high-energy particles on materials is a critical design parameter to consider when choosing polymeric structural, nonstructural, and elastomeric matrix resin systems. Polymer materials used for filled resins, laminates, seals, gaskets, coatings, insulation and other nonmetallic components must be chosen carefully, and reference data viewed with caution. Most reference data collected in the high-energy physics community to date reflects material property degradation using other than proton irradiations. In most instances, the data were collected for room-temperature irradiations, not 4.2 K or other cryogenic temperatures, and at doses less than 10 8 --10 9 Rad. Energetic proton (and the accompanying spallation-product particles) provide good simulation fidelity to the expected radiation fields predicted for the cold-mass regions of the SSC magnets, especially the corrector magnets. The authors present here results for some structural composite materials which were part of a larger irradiation-characterization of polymeric materials for SSC applications

Growth and instability of charged dislocation loops under irradiation in ceramic materials

CERN Document Server

Ryazanov, A I; Kinoshita, C; Klaptsov, A V

2002-01-01

We have investigated the physical mechanisms of the growth and stability of charged dislocation loops in ceramic materials with very strong different mass of atoms (stabilized cubic zirconia) under different energies and types of irradiation conditions: 100-1000 keV electrons, 100 keV He sup + and 300 keV O sup + ions. The anomalous formation of extended defect clusters (charged dislocation loops) has been observed by TEM under electron irradiation subsequent to ion irradiation. It is demonstrated that very strong strain field (contrast) near charged dislocation loops is formed. The dislocation loops grow up to a critical size and after then become unstable. The instability of the charged dislocation loop leads to the multiplication of dislocation loops and the formation of dislocation network near the charged dislocation loops. A theoretical model is suggested for the explanation of the growth and stability of the charged dislocation loop, taking the charge state of point defects. The calculated distribution...

Investigation of electrophysical properties of electrical insulating materials under neutron irradiation

International Nuclear Information System (INIS)

Skornyakov, Yu.A.; Stepanov, A.N.; Lapenas, A.A.

1978-01-01

The possibilities of applicaiton of insulating materials on the basis of glass cloths in electric windings for operation under neutron radiation of thermonuclear devices are studied. Changes in the specimen resistance, tangent of the angle of dielectric losses, electric strength according to the value of neutron fluence are determined. The temperature regimes are also studied. The data indicate the irreversible changes in the composition and structure of the polymer material under irradiation. The LSMI 228L-80 glass cloth has the highest radiation resistance. The necessity of forced cooling of large-sized specimens under the neutron radiation the IRT-200 reactor is established. The presence of impurities leading to the long-term induced activity of the insulating materials ( 59 Fe, 60 Co) is determined

Status and possible prospects of an international fusion materials irradiation facility

International Nuclear Information System (INIS)

Cozzani, F.

1999-01-01

Structural materials for future DT fusion power reactors will have to operate under intense neutron fields with energies up to 14 MeV and fluences in the order of 2 MW/m 2 per year. As environmental acceptability, safety considerations and economic viability will be ultimately the keys to the widespread introduction of fusion power, the development of radiation-resistant and low activation materials would contribute significantly to fusion development. For this purpose, testing of materials under irradiation conditions close to those expected in a fusion power station would require the availability, in an appropriate time framework, of an intense, high-energy neutron source. Recent advances in linear accelerator technology, in small specimens testing technology, and in the comprehension of damage phenomena, lead to the conclusion that an accelerator-based D-Li neutron source, with beam energy variability, would provide the most realistic option for a fusion materials testing facility. Under the auspices of the IEA, an international effort (EU, Japan, US, RF) to carry out the conceptual design activities (CDA) of an international fusion materials irradiation facility (IFMIF), based on the D-Li concept, have been carried out successfully. A final conceptual design report was produced at the end of 1996. A phase of conceptual design evaluation (CDE), presently underway, is extending and further refining some of the conceptual design details of IFMIF. The results indicate that an IFMIF-class installation would be technically feasible and could meet its mission objectives. However, a suitable phase of Engineering Validation, to carry out some complementary R and D and prototyping, would still be needed to resolve a few key technical uncertainties before the possibility to proceed toward detailed design and construction could be explored. (orig.)

Delayed hydride cracking in irradiated Zr-2.5 % Nb pressure tubes

International Nuclear Information System (INIS)

Cirimello, Pablo; Coronel, Pascual; Haddad, Roberto; Lafont, Claudio; Mizrahi, Rafael

2003-01-01

Pressure tubes in CANDU nuclear power plants are made of Zr-2.5 % Nb alloy, which is susceptible to a cracking process called Delayed Hydride Cracking (DHC). Measurement of DHC velocity on irradiated pressure tubes is essential to assure the validity of the Leak Before Break criterion. This work was performed on samples from two pressure tubes taken out of the Embalse NPP in 1995, belonging to fuel channels A-14 and L-12. DHC velocity in the axial direction was measured at 211 C degrees for samples taken from different axial positions, which allowed to study its dependence on fast neutron fluency and irradiation temperature. Non-irradiated material was also tested. It was found that DHC velocity results for the tested material were similar to those obtained for a great number of tubes irradiated in other CANDU plants. (author)

Report on the program of 4 K irradiation of insulating materials for the Superconducting Super Collider

International Nuclear Information System (INIS)

Spindel, A.

1993-07-01

This report is intended to serve as an aid to material selection. The results reported herein are the product of a careful investigation and can be used with confidence in their validity. The selection of materials based on this data, however, is not the responsibility of the author. This report will not approve or disapprove any specific material for use in the Super Collider. The author of this report does not assume any design responsibility or responsibility for material selection for any application. It is, therefore, very important that those with design responsibility use this report wisely. For this reason, the following informational guide to the material selection process has been provided. There are several issues to take into account when evaluating a material for radiation resistance. It is very important that the design criteria and operating loads for the application be known. For many applications the actual loading, and therefore required properties, are unknown. Certain materials have empirically been used successfully in a similar application and those materials have often been selected on that basis. Both percent degradation and the magnitude of the actual properties after irradiation need to be considered. Consider the scenario where two materials are being compared that both have acceptable properties after exposure to 10 9 rads. It is preferable to choose the material with less degradation because degradation tends to be a threshold phenomena with properties declining rapidly with dose after a certain threshold dose. The properties of the initially strong material, therefore, will be extremely sensitive to dose in that dose range and slight magnet-to-magnet differences in dose may, depending on the application, lead to performance variations

Preparation of a new gamma irradiated PVC-Olive oil cake plastic composite material

International Nuclear Information System (INIS)

Messaud, F.A.; Almsmary, Y.A.; Elwerfalli, S.M.; Benayad, S.M.; Haraga, S.O.; Benfaid, N.A.; Kabar, Y.M.

2003-01-01

This paper dealt with the investigation on preparing new plastic composite material, utilizing polyvinyl chloride polymer (a commercial product in abu-kammash chemical complex) and olive oil cake (a waste of many olive oil production factories), followed by gamma irradiation (26.3 Kg ry) o induce crosslinking of the polymer. The new material possess good, electrical and mechanical properties as compared to plastic products of (PVC plastic pipe factory), and which could be used as new construction anti corrosive material, such as special roofing and partitioning or household goods

Neutron irradiation facility and its characteristics

International Nuclear Information System (INIS)

Oyama, Yukio; Noda, Kenji

1995-01-01

A neutron irradiation facility utilizing spallation reactions with high energy protons is conceived as one of the facilities in 'Proton Engineering center (PEC)' proposed at JAERI. Characteristics of neutron irradiation field of the facility for material irradiation studies are described in terms of material damage parameters, influence of the pulse irradiation, irradiation environments other than neutronics features, etc., comparing with the other sorts of neutron irradiation facilities. Some perspectives for materials irradiation studies using PEC are presented. (author)

Study on ceramic breeder and related materials by means of work function measurement under irradiation

International Nuclear Information System (INIS)

Luo, G.N.; Terai, T.; Yamawaki, M.; Yamaguchi, K.

2002-01-01

Ceramic breeder materials, Li 2 O, LiAlO 2 and Li 4 SiO 4 , under irradiation have been studied using a Kelvin probe that measures work function changes of materials. Surface charging was observed to influence greatly the probe output, which can be explained qualitatively employing a model concerning induction electric field due to external field and free charges on ceramic surface. It is found that the insulating ceramics could not be studied properly with the Kelvin probe. A probable solution is to heat the ceramics, so as to raise their electric conductivities high enough to root out the surface charging. Also briefly discussed is the application of the probe to metals under ion irradiation. (orig.)

Material modifications in lithium niobate and lithium tantalate crystals by ion irradiation

International Nuclear Information System (INIS)

Raeth, Niels Lennart

2017-01-01

The artificially produced crystals lithium niobate (LiNbO 3 ) and the closely related lithium tantalate (LiTaO 3 ) are proven starting materials for producing active and passive devices that can guide, amplify, switch and process light. For this purpose, it is often necessary to be able to influence the refractive index of the substrate targeted, which is possible in addition to other methods by irradiation of the materials with fast light ions. In this work, lithium niobate and lithium tantalate crystals are irradiated with alpha particles, 3 He ions, deuterons, and protons at projectile energies of up to 14 MeV / nucleon. Energy and crystal thickness are chosen so that the projectiles penetrate the entire sample and are not implanted. All isotopes responsible for the unwanted nuclear activation of the crystals due to the irradiation are relatively short-lived and overall the activation decreases fast enough to allow the safe handling of the irradiated samples after a storage period of a few days to a few weeks. The refractive index changes produced in lithium niobate and lithium tantalate by irradiation with the different projectiles are determined interferometrically and can also be measured by suitable choice of the sample geometry as a function of the ion penetration depth: In LiNbO 3 the ordinary refractive index decreases, the extraordinary increases equally. In LiTaO 3 , both the ordinary and the extraordinary refractive indices decrease as a result of the irradiation; the ordinary refractive index change is many times stronger than the extraordinary one. There is an enormous long-term stability at room temperature for both crystal systems: Even after eleven (LiNbO 3 ) or three (LiTaO 3 ) years, no decrease in the ion beam-induced refractive index change can be observed. The ion beam-induced refractive index changes are probably the result of atomic displacements such as vacancies, defect clusters or ''latent tracks''. An explanation for

AGC-2 Specimen Post Irradiation Data Package Report

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-01

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One-half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time-of-flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between

Precipitation response of annealed type 316 stainless steel in HFIR irradiations at 550 to 6800C

International Nuclear Information System (INIS)

Maziasz, P.J.

1978-01-01

Precipitation in annealed type 316 stainless steel after HFIR irradiation at 550--680 0 C to fluences producing 2000--3300 at. ppM He and 30--47 dpa is changed relative to fast reactor or thermal aging exposure to similar temperatures and times. The phases observed after HFIR irradiation are the same as those observed after aging to temperatures 70--200 0 C higher or for much longer times. There is a similar temperature shift in addition to different phases observed for HFIR irradiation compared with EBR-II. The changes observed are coincident with including simultaneous helium production to high levels in the irradiation damage products of the material

Post-Irradiation Properties of Candidate Materials for High-Power Targets

International Nuclear Information System (INIS)

Kirk, H.G.; Ludewig, H.; Mausner, L.F.; Simos, N.; Thieberger, P.; Brookhaven; Hayato, Y.; Yoshimura, K.; McDonald, K.T.; Sheppard, J.; Trung, L.P.

2006-01-01

The desire of the high-energy-physics community for more intense secondary particle beams motivates the development of multi-megawatt, pulsed proton sources. The targets needed to produce these secondary particle beams must be sufficiently robust to withstand the intense pressure waves arising from the high peak-energy deposition which an intense pulsed beam will deliver. In addition, the materials used for the targets must continue to perform in a severe radiation environment. The effect of the beam-induced pressure waves can be mitigated by use of target materials with high-yield strength and/or low coefficient of thermal expansion (CTE) [1, 2, 3]. We report here first results of an expanded study of the effects of irradiation on several additional candidate materials with high strength (AlBeMet, beryllium, Ti-V6-Al4) or low CTE (a carbon-carbon composite, a new Toyota ''gum'' metal alloy [4], Super-Invar)

Influence of Curing Humidity on the Compressive Strength of Gypsum-Cemented Similar Materials

Directory of Open Access Journals (Sweden)

Weiming Guan

2016-01-01

Full Text Available The analogous simulation experiment is widely used in geotechnical and mining engineering. However, systematic errors derived from unified standard curing procedure have been underestimated to some extent. In this study, 140 gypsum-cemented similar material specimens were chosen to study their curing procedure with different relative humidity, which is 10%–15%, 40%, 60%, and 80%, respectively. SEM microstructures and XRD spectra were adopted to detect the correlation between microstructures and macroscopic mechanical strength during curing. Our results indicated that the needle-like phases of similar materials began to develop in the early stage of the hydration process through intersecting with each other and eventually transformed into mat-like phases. Increase of humidity may inhibit the development of needle-like phases; thus the compressive strength changes more smoothly, and the time required for the material strength to reach the peak value will be prolonged. The peak strength decreases along with the increase of humidity while the humidity is higher than 40%; however, the reverse tendency was observed if the humidity was lower than 40%. Finally, we noticed that the material strength usually reaches the peak value when the water content continuously reduces and tends towards stability. Based on the above observation, a curing method determination model and experimental strength predication method for gypsum-cemented similar materials were proposed.

Thermoluminescence response of gamma-irradiated sesame with mineral dust

Energy Technology Data Exchange (ETDEWEB)

Rodriguez L, Y. [CSIC, Instituto de Estructura de la Materia, Calle Serrano 121, 28006 Madrid (Spain); Correcher, V. [CIEMAT, Av. Complutense 22, 28040 Madrid (Spain); Garcia G, J. [CSIC, Museo Nacional de Ciencias Naturales, Calle Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Cruz Z, E., E-mail: y.r.l@csic.es [UNAM, Instituto de Ciencias Nucleares, Circuito Exterior s/n, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

2011-10-15

The thermoluminescence (Tl) emission of minerals isolated from Mexican and Indian sesame seeds appear as a good tool to discern between irradiated and non-irradiated samples. According to the X-ray diffraction and environmental scanning microscope, the adhered dust in both samples is mainly composed by different amounts of quartz and feldspars. These mineral phases exhibit (i) enough sensitivity to ionizing radiation inducing good Tl intensity, (ii) high stability of the Tl signal during the storage of the material (i.e. low fading) and (iii) are thermally and chemically stable. Blind tests performed under laboratory conditions, but simulating industrial preservation processes (similar temperature and moisture, and presence of white light), allows to distinguish between 1 KGy gamma-irradiated and non-irradiated samples even 11000 hours (15 months) after the irradiation proceeding. (Author)

Thermoluminescence response of gamma-irradiated sesame with mineral dust

International Nuclear Information System (INIS)

Rodriguez L, Y.; Correcher, V.; Garcia G, J.; Cruz Z, E.

2011-10-01

The thermoluminescence (Tl) emission of minerals isolated from Mexican and Indian sesame seeds appear as a good tool to discern between irradiated and non-irradiated samples. According to the X-ray diffraction and environmental scanning microscope, the adhered dust in both samples is mainly composed by different amounts of quartz and feldspars. These mineral phases exhibit (i) enough sensitivity to ionizing radiation inducing good Tl intensity, (ii) high stability of the Tl signal during the storage of the material (i.e. low fading) and (iii) are thermally and chemically stable. Blind tests performed under laboratory conditions, but simulating industrial preservation processes (similar temperature and moisture, and presence of white light), allows to distinguish between 1 KGy gamma-irradiated and non-irradiated samples even 11000 hours (15 months) after the irradiation proceeding. (Author)

Effects of wearing bio-active material coated fabric against γ-irradiation-induced cellular damaged in Sprague-Dawley rats

International Nuclear Information System (INIS)

Kang, Jung Ae; Kim, Hye Rim; Yoon, Sun Hye; Nam, Sang Hyun; Park, Sang Hyun; Jang, Beom Su; Go, Kyung Chan; Yang, Gwang Wung; Rho, Young Hwan; Park, Hyo Suk

2016-01-01

Ionizing radiation causes cellular damage and death through the direct damage and/or indirectly the production of ROS, which induces oxidative stress. This study was designed to evaluate the in vivo radioprotective effects of a bio-active material coated fabric (BMCF) against γ-irradiation-induced cellular damage in Sprague-Dawley (SD) rats. Healthy male SD rats wore bio-active material coated (concentrations in 10% and 30%) fabric for 7 days after 3 Gy of γ-irradiation. Radioprotective effects were evaluated by performing various biochemical assays including spleen and thymus index, WBC count, hepatic damage marker enzymes [aspartate transaminase (AST) and alanine transaminase (ALT)] in plasma, liver antioxidant enzymes, and mitochondrial activity in muscle. Exposure to γ-irradiation resulted in hepatocellular and immune systemic damage. Gamma-irradiation induced decreases in antioxidant enzymes. However, wearing the BMCF-30% decreased significantly AST and ALT activities in plasma. Furthermore, wearing the BMCF-30% increased SOD (superoxide dismutase) and mitochondrial activity. These results suggest that wearing BMCF offers effective radioprotection against γ-irradiation-induced cellular damage in SD rats

Effects of wearing bio-active material coated fabric against γ-irradiation-induced cellular damaged in Sprague-Dawley rats

Energy Technology Data Exchange (ETDEWEB)

Kang, Jung Ae; Kim, Hye Rim; Yoon, Sun Hye; Nam, Sang Hyun; Park, Sang Hyun; Jang, Beom Su [Korea Atomic Energy Research Institute, Jeongeup (Korea, Republic of); Go, Kyung Chan; Yang, Gwang Wung; Rho, Young Hwan; Park, Hyo Suk [Research and Development Center, VENTEX Co. Ltd., Seoul (Korea, Republic of)

2016-09-15

Ionizing radiation causes cellular damage and death through the direct damage and/or indirectly the production of ROS, which induces oxidative stress. This study was designed to evaluate the in vivo radioprotective effects of a bio-active material coated fabric (BMCF) against γ-irradiation-induced cellular damage in Sprague-Dawley (SD) rats. Healthy male SD rats wore bio-active material coated (concentrations in 10% and 30%) fabric for 7 days after 3 Gy of γ-irradiation. Radioprotective effects were evaluated by performing various biochemical assays including spleen and thymus index, WBC count, hepatic damage marker enzymes [aspartate transaminase (AST) and alanine transaminase (ALT)] in plasma, liver antioxidant enzymes, and mitochondrial activity in muscle. Exposure to γ-irradiation resulted in hepatocellular and immune systemic damage. Gamma-irradiation induced decreases in antioxidant enzymes. However, wearing the BMCF-30% decreased significantly AST and ALT activities in plasma. Furthermore, wearing the BMCF-30% increased SOD (superoxide dismutase) and mitochondrial activity. These results suggest that wearing BMCF offers effective radioprotection against γ-irradiation-induced cellular damage in SD rats.

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

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

In-reactor precipitation and ferritic transformation in neutron--irradiated stainless steels

International Nuclear Information System (INIS)

Porter, D.L.; Wood, E.L.

1978-01-01

Ferritic transformation (γ → α) was observed in Type 304L, 20% cold-worked AISI 316, and solution-annealed AISI 316 stainless steels subjected to fast neutron irradiation. Each material demonstrated an increasing propensity for transformation with increasing irradiation temperature between 400 and 550 0 C. Irradiation-induced segregation of Ni solute to precipitates was found not to influence the transformation kinetics in 304L. Similar composition data from 316 materials demonstrates a much greater temperature dependence of precipitation reactions in the process of matrix Ni depletion during neutron irradiation. The 316 data establishes a strong link between such depletion and the observed γ → α transformation. Moreover, the lack of correlation between precipitate-related Ni depletion and the γ → α transformation in 304L can be related to the fact that irradiation-induced voids nucleate very quickly in 304L steel during irradiation. These voids present preferential sites for Ni segregation through a defect trapping mechanism, and hence Ni segregates to voids rather than to precipitates, as evidenced by observed stable γ shells around voids in areas of complete transformation

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.

Observation of He bubbles in ion irradiated fusion materials by conductive atomic force microscopy

Energy Technology Data Exchange (ETDEWEB)

Fan, Hongyu [School of Physics and Materials Engineering, Dalian Nationalities University, Dalian 116600 (China); Li, Ruihuan [School of Physics and Optoelectronic Engineering, Dalian University of Technology, Dalian 116024 (China); Yang, Deming [School of Physics and Materials Engineering, Dalian Nationalities University, Dalian 116600 (China); School of Science, Changchun University of Science and Technology, Changchun, Jilin 130022 (China); Wu, Yunfeng; Niu, Jinhai; Yang, Qi [School of Physics and Materials Engineering, Dalian Nationalities University, Dalian 116600 (China); Zhao, Jijun [School of Physics and Optoelectronic Engineering, Dalian University of Technology, Dalian 116024 (China); Liu, Dongping, E-mail: dongping.liu@dlnu.edu.cn [School of Physics and Materials Engineering, Dalian Nationalities University, Dalian 116600 (China); Fujian Key Laboratory for Plasma and Magnetic Resonance, Department of Electronic Science, Aeronautics, School of Physics and Mechanical and Electrical Engineering, Xiamen University, Xiamen, Fujian 361005 (China)

2013-10-15

Using a non-destructive conductive atomic force microscope combined with the Ar{sup +} etching technique, we demonstrate that nanoscale and conductive He bubbles are formed in the implanted layer of single-crystalline 6H-SiC irradiated with 100 keV He{sup +}. We find that the surface swelling of irradiated SiC samples is well correlated with the growth of elliptic He bubbles in the implanted layer. First-principle calculations are performed to estimate the internal pressure of the He bubble in the void of SiC. Analysis indicates that nanoscale He bubbles acting as a captor capture the He atoms diffusing along the implanted layer at an evaluated temperature and result in the surface swelling of irradiated SiC materials.

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

CERN Document Server

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

2000-01-01

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

Displacement per atom profile in carbon nanotube bulk material under gamma irradiation

International Nuclear Information System (INIS)

Leyva, A.; Pinnera, I.; Leyva, D.; Cruz, C.; Abreu, Y.

2011-01-01

Taking into account the physical properties and the displacement threshold energy values reported in literature for C atoms in single and multiple walled carbon nanotubes, the effective atomic displacement cross-section in carbon nanotube bulk materials exposed to the gamma rays were calculated. Then, using the mathematical simulation of photons and particles transport in the matter, energy fluxes distribution of electrons and positrons within the irradiated object were also calculated. Finally, considering both results, the atomic displacement damage profiles inside the analyzed carbon nanotube bulk materials were determined. (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

Van de Graaff Irradiation of Materials

Energy Technology Data Exchange (ETDEWEB)

Quigley, Kevin [Argonne National Lab. (ANL), Argonne, IL (United States); Chemerisov, Sergey [Argonne National Lab. (ANL), Argonne, IL (United States); Tkac, Peter [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-10-01

Through irradiations using our 3 MeV Van de Graaf accelerator, Argonne is testing the radiation stability of components of equipment that are being used to dispense molybdenum solutions for use as feeds to ^99mTc generators and in the ^99mTc generators themselves. Components have been irradiated by both a direct electron beam and photons generated from a tungsten convertor.

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-02-16

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

Selective Adsorption of Nano-bio materials and nanostructure fabrication on Molecular Resists Modified by proton beam irradiation

International Nuclear Information System (INIS)

Lee, H. W.; Kim, H. S.; Kim, S. M.

2008-04-01

The purpose of this research is the fabrication of nanostructures on silicon substrate using proton beam and selectively adsorption of bio-nano materials on the patterned substrate. Recently, the miniaturization of the integrated devices with fine functional structures was intensively investigated, based on combination of nanotechnology (NT), biotechnology (BT) and information technology (IT). Because of the inherent limitation in optical lithography, large variety of novel patterning technologies were evolved to construct nano-structures onto a substrate. Atomic force microscope-based nanolithography has readily formed sub-50 nm patterns by the local modification of a substrate using a probe with a curvature of 10 nm. The surface property was regarded as one of the most important factors for AFM-based nanolithography as well as for other novel nanolithographies. The molecular thin films such as a self-assembled monolayer or a polymer resist layer have been used as an alternative to modifying the surface property. Although proton or ion beam irradiation has been used as an efficient tool to modify the physical, chemical and electrical properties of a surface, the nano-patterning on the substrate or the molecular film modified with the beam irradiation has hardly been studied at both home and abroad. The selective adsorption of nano-bio materials such as carbon nanotubes and proteins on the patterns would contribute to developing the integrated devices. The polystyrene nanoparticles (400 nm) were arrayed on al silicon surface using nanosphere lithography and the various nanopatterns were fabricated by proton beam irradiation on the polystyrene nanoparticles arrayed silicon surface. We obtained the two different nanopatterns such as polymer nanoring patterns and silicon oxide patterns on the same silicon substrate. The polymer nanoring patterns formed by the crosslinkage of polystyrene when proton beam was irradiated at the triangular void spaces that are enclosed by

Review of recent irradiation-creep results

International Nuclear Information System (INIS)

Coghlan, W.A.

1982-05-01

Materials deform faster under stress in the presence of irradiation by a process known as irradiation creep. This phenomenon is important to reactor design and has been the subject of a large number of experimental and theoretical investigations. The purpose of this work is to review the recent experimental results to obtain a summary of these results and to determine those research areas that require additional information. The investigations have been classified into four subgroups based on the different experimental methods used. These four are: (1) irradiation creep using stress relaxation methods, (2) creep measurements using pressurized tubes, (3) irradiation creep from constant applied load, and (4) irradiation creep experiments using accelerated particles. The similarity and the differences of the results from these methods are discussed and a summary of important results and suggested areas for research is presented. In brief, the important results relate to the dependence of creep on swelling, temperature, stress state and alloying additions. In each of these areas new results have been presented and new questions have arisen which require further research to answer. 65 references

The effects of packaging materials on microbe population in irradiated traditional herbal medicines

International Nuclear Information System (INIS)

Bagiawati, Sri; Hilmy, Nazly

1983-01-01

Microbial population and moisture content of traditional herbal medicines contaminated with 3 kinds of aerobic microbes, packed in 5 kinds of plastic packaging materials, followed by irradiation at minimum dose of 5 kGy and stored for 6 months were investigated. The highest reduction of microbial counts during storage was observed on samples packed in polyethylene bags. All of packaging materials used were found to be impermeable to microbes and water vapour. Radiation and packaging materials used acted synergistically to inactivate microbes durind storage. The microbial counts decreased as much as 2 to 4 log cycles during storage. (author)

Disintegration of C60 by Xe ion irradiation

International Nuclear Information System (INIS)

Kalish, R.; Samoiloff, A.; Hoffman, A.; Uzan-Saguy, C.

1993-01-01

The Changes in resistivity of fullerene (C 60 ) films subject to 320 keV Xe ion irradiation are investigated as a function of ion dose. From a comparison of this dependence with similar data on other Xe irradiated C containing insulating materials and with data on C implanted fused quartz, it is concluded that upon ion impact C 60 clusters completely disintegrate. This disintegration releases about 60 C atoms which disperse amongst the remaining intact C 60 spheres giving rise to hopping conductivity between isolated C atoms. 16 refs., 3 figs

Insulation interlaminar shear strength testing with compression and irradiation

International Nuclear Information System (INIS)

McManamy, T.J.; Brasier, J.E.; Snook, P.

1989-01-01

The Compact Ignition Tokamak (CIT) project identified the need for research and development for the insulation to be used in the toroidal field coils. The requirements included tolerance to a combination of high compression and shear and a high radiation dose. Samples of laminate-type sheet material were obtained from commercial vendors. The materials included various combinations of epoxy, polyimide, E-glass, S-glass, and T-glass. The T-glass was in the form of a three-dimensional weave. The first tests were with 50 x 25 x 1 mm samples. These materials were loaded in compression and then to failure in shear. At 345-MPa compression, the interlaminar shear strength was generally in the range of 110 to 140 MPa for the different materials. A smaller sample configuration was developed for irradiation testing. The data before irradiation were similar to those for the larger samples but approximately 10% lower. Limited fatigue testing was also performed by cycling the shear load. No reduction in shear strength was found after 50,000 cycles at 90% of the failure stress. Because of space limitations, only three materials were chosen for irradiation: two polyimide systems and one epoxy system. All used boron-free glass. The small shear/compression samples and some flexure specimens were irradiated to 4 x 10 9 and 2 x 10 10 rad in the Advanced Technology Reactor at Idaho National Engineering Laboratory. A lead shield was used to ensure that the majority of the dose was from neutrons. The shear strength with compression before and after irradiation at the lower dose was determined. Flexure strength and the results from irradiation at the higher dose level will be available in the near future. 7 refs., 7 figs., 2 tabs

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)

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

Estimation of irradiation-induced material damage measure of FCM fuel in LWR core

International Nuclear Information System (INIS)

Lee, Kyung-Hoon; Lee, Chungchan; Park, Sang-Yoon; Cho, Jin-Young; Chang, Jonghwa; Lee, Won Jae

2014-01-01

An irradiation-induced material damage measure on tri-isotropic (TRISO) multi-coating layers of fully ceramic micro-encapsulated (FCM) fuel to replace conventional uranium dioxide (UO 2 ) fuel for existing light water reactors (LWRs) has been estimated using a displacement per atom (DPA) cross section for a FCM fuel performance analysis. The DPA cross sections in 47 and 190 energy groups for both silicon carbide (SiC) and graphite are generated based on the molecular dynamics simulation by SRIM/TRIM. For the selected FCM fuel assembly design with FeCrAl cladding, a core depletion analysis was carried out using the DeCART2D/MASTER code system with the prepared DPA cross sections to evaluate the irradiation effect in the Korean OPR-1000. The DPA of the SiC and IPyC coating layers is estimated by comparing the discharge burnup obtained from the MASTER calculation with the burnup-dependent DPA for each coating layer calculated using DeCART2D. The results show that low uranium loading and hardened neutron spectrum compared to that of high temperature gas-cooled reactor (HTGR) result in high discharge burnup and high fast neutron fluence. In conclusion, it can be seen that the irradiation-induced material damage measure is noticeably increased under LWR operating conditions compared to HTGRs. (author)

NRI experimental facility for the testing of irradiation assisted stress corrosion cracking

International Nuclear Information System (INIS)

Ruscak, M.; Chvatal, P.; Zamboch, M.

1998-01-01

IASCC influencing reactor internals of both BWR and PWR reactors is a complex phenomenon covering influences of material structure, neutron fluence, neutron flux, chemistry of environment, gamma radiation and mechanical stress. To evaluate such degradation, tests should be performed under conditions similar to those in real structure. Nuclear Research Institute has built several experimental facilities in order to be able to test IASCC degradation of materials. Basically, reactor water loops, both PWR and BWR, could be used to model environmental conditions including gamma and neutron irradiation. Pre-irradiation can be done in irradiation channels under well controlled temperature conditions. During the experiment, in-pile conditions can be compared with those out of pile. It enables to clarify pure influence of irradiation. For testing of irradiated specimens, hot cell facility has been developed for slow strain rate tests. The paper will show all above mentioned facilities as well as some of the results observed with them. (author)

Neutron-irradiation facilities at the Intense Pulsed Neutron Source-I for fusion magnet materials studies

International Nuclear Information System (INIS)

Brown, B.S.; Blewitt, T.H.

1982-01-01

The decommissioning of reactor-based neutron sources in the USA has led to the development of a new generation of neutron sources that employ high-energy accelerators. Among the accelerator-based neutron sources presently in operation, the highest-flux source is the Intense Pulsed Neutron Source (IPNS), a user facility at Argonne National Laboratory. Neutrons in this source are produced by the interaction of 400 to 500 MeV protons with either of two 238 U target systems. In the Radiation Effects Facility (REF), the 238 U target is surrounded by Pb for neutron generatjion and reflection. The REF has three separate irradiation thimbles. Two thimbles provide irradiation temperatures between that of liquid He and several hundred degrees centigrade. The third thimble operates at ambient temperature. The large irradiation volume, the neutron spectrum and flux, the ability to transfer samples without warm up, and the dedication of the facilities during the irradiation make this ideally suited for radiation damage studies on components for superconducting fusion magnets. Possible experiments for fusion magnet materials are discussed on cyclic irradiation and annealing of stabilizers in a high magnetic field, mechanical tests on organic insulation irradiated at 4 K, and superconductors measured in high fields after irradiation

Study of dielectric materials irradiated with electron beam by using the Pulsed Electro-Acoustic (PEA) method

International Nuclear Information System (INIS)

Nguyen, Xuan Truong

2014-01-01

Dielectric materials are frequently used as electrical insulators in spatial applications. Due to their dielectric nature, these dielectrics are likely to accumulate electric charges during their service. Under certain critical conditions, these internal or surface space charges can lead to an electrostatic surface discharge. To understand these phenomena, an experimental device has been developed in the laboratory. This device allows us to simulate the electronic irradiation conditions encountered in space. The aim of our study is to characterize the electrical behavior of insulating materials irradiated by electron beam, to investigate charge storage and transport phenomena and anticipate electrostatic discharges. In this work, the device based on the Pulsed Electro-Acoustic (PEA) technique has been chosen. It has been implanted in the irradiation chamber. It allows us to obtain the spatial distribution of charges injected between two periods of irradiation and during relaxation. However the PEA method offers a limited resolution and does not allow the detection of injected charges when they are too close to the surface. First, we performed a parameters signal processing analysis that we will call the spreading factor and the resolution factor. The preliminary study post-irradiation in air of experimental measurements showed that the resolution factor choice is important for the analysis and interpretation of the signal when the space charge is localized near the surface. Then, a comparison to the spreading parameter used in some deconvolution technique was established. In the second time, space charge distribution measurements in vacuum have been carried out on Poly Tetra Fluoro Ethylene (PTFE) films irradiated by an electron beam in the range [10-100] keV. Results from irradiation periods with increasing energies [10 keV → 100 keV] of the electron beam have been compared with results from irradiation periods with decreasing energies [100 keV → 10 keV]. In

Low dose irradiation performance of SiC interphase SiC/SiC composites

International Nuclear Information System (INIS)

Snead, L.L.; Lowden, R.A.; Strizak, J.; More, K.L.; Eatherly, W.S.; Bailey, J.; Williams, A.M.; Osborne, M.C.; Shinavski, R.J.

1998-01-01

Reduced oxygen Hi-Nicalon fiber reinforced composite SiC materials were densified with a chemically vapor infiltrated (CVI) silicon carbide (SiC) matrix and interphases of either 'porous' SiC or multilayer SiC and irradiated to a neutron fluence of 1.1 x 10 25 n m -2 (E>0.1 MeV) in the temperature range of 260 to 1060 C. The unirradiated properties of these composites are superior to previously studied ceramic grade Nicalon fiber reinforced/carbon interphase materials. Negligible reduction in the macroscopic matrix microcracking stress was observed after irradiation for the multilayer SiC interphase material and a slight reduction in matrix microcracking stress was observed for the composite with porous SiC interphase. The reduction in strength for the porous SiC interfacial material is greatest for the highest irradiation temperature. The ultimate fracture stress (in four point bending) following irradiation for the multilayer SiC and porous SiC interphase materials was reduced by 15% and 30%, respectively, which is an improvement over the 40% reduction suffered by irradiated ceramic grade Nicalon fiber materials fabricated in a similar fashion, though with a carbon interphase. The degradation of the mechanical properties of these composites is analyzed by comparison with the irradiation behavior of bare Hi-Nicalon fiber and Morton chemically vapor deposited (CVD) SiC. It is concluded that the degradation of these composites, as with the previous generation ceramic grade Nicalon fiber materials, is dominated by interfacial effects, though the overall degradation of fiber and hence composite is reduced for the newer low-oxygen fiber. (orig.)

Deuterium ion irradiation damage and deuterium trapping mechanism in candidate stainless steel material (JPCA2) for fusion reactor

International Nuclear Information System (INIS)

Ashizuka, Norihiro; Kurita, Takaaki; Yoshida, Naoaki; Fujiwara, Tadashi; Muroga, Takeo

1987-01-01

An improved austenitic stainless steel (JPCA), a candidate material for fusion reactor, is irradiated at room temperature with deuterium ion beams. Desorption spectra of deuterium gas is measured at various increased temperatures and defects formed under irradiation are observed by transmission electron microscopy to determine the mechanism of the thermal release of deuteriums and the characteristics of irradiation-induced defects involved in the process. In the deuterium deportion spectra observed, five release stages are found to exist at 90 deg C, 160 deg C, 220 deg C, 300 deg C and 400 deg C, referred to as Stage I, II, III, IV and V, respectively. Stage I is interpreted as representing the release of deuteriums trapped in point defects (presumably vacancies) formed under irradiation. The energy of desorption from the trapping sites is estimated at 0.8 eV. Stage II is concluded to be associated with the release of deuteriums trapped in a certain kind of existing defects. Stage III involves the release of deuteriums that are trapped in dislocations, dislocation loops or dislocated portions of stacking fault tetrahedra. This release occurs significantly in processed materials and other materials irradiated with high energy ion beams that may cause cascade damage. Stage IV is interpreted in terms of thermal decomposition of small deuterium clusters. Stage V is associated with the decomposition of rather large deuterium clusters grown on the {111} plane. (Nogami, K.)

ESR (Electronic Spin Resonance Spectroscopy) study of irradiated paper for biomedical material wrapping

International Nuclear Information System (INIS)

Huarte, Monica; Rubin de Celis, Emilio; Kairiyama, Eulogia; Zapata, Miguel; Santoro, Natalia; Magnavacca, Cecilia

2009-01-01

Ionising radiation treatments are used for sterilization, microbiological decontamination, disinfection, insect disinfestation and food preservation. This ionising radiation generates free radicals (FR) in matter, which can be detected by Electronic Spin Resonance Spectroscopy (ESR). For this work it had analysed different kind of irradiated package papers of syringes, surgical gloves and dressings by ESR. These were irradiated with doses between 20 and 35 kGy of gamma radiation (Cobalt 60). The processed samples were measured in a Bruker ECS 106 spectrometer. The obtained results were: 1-) The irritated samples showed a central peak and two satellites induced by the applied radiation; 2-) The non-irradiated samples did not show the characteristic satellite peaks of the irritated ones; 3-) A linear relationship between the signal heights per unit mass and the applied doses was found; and 4-) The signals were highly stable, with half-time values between 240 and 370 days for 20 and 30 kGy, permitting more than one year of monitoring proceedings. In conclusion, the ESR allows the detection, quantification and time monitoring processes of this kind of irradiated materials. (author) [es

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.