Fusion neutron irradiation of Ni(Si) alloys at high temperature

International Nuclear Information System (INIS)

Huang, J.S.; Guinan, M.W.; Hahn, P.A.

1987-09-01

Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab

Fusion neutron irradiation of Ni(Si) alloys at high temperature

Energy Technology Data Exchange (ETDEWEB)

Huang, J.S.; Guinan, M.W.; Hahn, P.A.

1987-09-01

Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab.

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

International Nuclear Information System (INIS)

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

1981-01-01

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

Irradiation tests of critical components for remote handling in gamma radiation environment

International Nuclear Information System (INIS)

Obara, Henjiro; Kakudate, Satoshi; Oka, Kiyoshi

1994-08-01

Since the fusion power core of a D-T fusion reactor will be highly activated once it starts operation, personnel access will be prohibited so that assembly and maintenance of the components in the reactor core will have to be totally conducted by remote handling technology. Fusion experimental reactors such as ITER require unprecedented remote handling equipments which are tolerable under gamma radiation of more than 10 6 R/h. For this purpose, the Japan Atomic Energy Research Institute (JAERI) has been developing radiation hard components for remote handling purpose and a number of key components have been tested over 10 9 rad at a radiation dose rate of around 10 6 R/h, using Gamma Ray Radiation Test Facility in JAERI-Takasaki Establishment. This report summarizes the irradiation test results and the latest status of AC servo motor, potentiometer, optical elements, lubricant, sensors and cables, which are key elements of the remote handling system. (author)

Effects of non-steady irradiation conditions on fusion materials performance

International Nuclear Information System (INIS)

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

2001-01-01

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

Fusion neutron irradiation of Ni-Si alloys at high temperature*1

Science.gov (United States)

Huang, J. S.; Guinan, M. W.; Hahn, P. A.

1988-07-01

Two Ni-4% Si alloys, with different cold work levels, have been irradiated with 14-MeV fusion neutrons at 623 K, and their Curie temperatures have been monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2-MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14-MeV fusion neutrons is only 6-7% of that for an identical alloy irradiated by 2-MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6-7% for the fusion neutron irradiated sample.

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.)

Effect of irradiation-induced defects on fusion reactor ceramics

International Nuclear Information System (INIS)

Clinard, F.W. Jr.

1986-01-01

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

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)

Calculation of damage function of Al{sub 2}O{sub 3} in irradiation facilities for fusion reactor applications

Energy Technology Data Exchange (ETDEWEB)

Mota, F., E-mail: fernando.mota@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Ortiz, C.J., E-mail: christophe.ortiz@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Vila, R., E-mail: rafael.vila@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Casal, N., E-mail: natalia.casal@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); García, A., E-mail: angela.garcia@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Ibarra, A., E-mail: Angel.ibarra@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain)

2013-11-15

A rigorous material testing program is essential for the development of the nuclear fusion world program. In particular, it is very important to predict the generation of the displacement damage in materials, because the irradiation intensity expected in fusion conditions is such that the performance of materials and components under these extreme conditions is unknown. To study the damage produced by neutrons in materials of interest for fusion, a specific computational methodology was developed. Neutron fluxes expected in different irradiation facilities (International Fusion Materials Irradiation Facility [IFMIF] and DEMO-HCLL) and in different irradiation spots were obtained with particles transport codes (McDeLicious, MCNP). The energy differential cross sections of primary knock-on atoms were calculated using the NJOY code. Resulting data were input into the Monte Carlo code MARLOWE to calculate the corresponding displacements (i.e., interstitials (I) and vacancies (V)). However, the number of Frenkel pairs created during irradiation strongly depends on the recombination radius between interstitials and vacancies. This parameter corresponds to the minimum distance below which instantaneous recombination occurs. Mainly, the influence of such parameter on the damage function in Al{sub 2}O{sub 3} was assessed in this report. The displacements per atom values calculated as a function of the recombination radius considered are compared to experimental data to determine the most appropriate capture radius. In addition, the damage function and damage dose generated at different experimental irradiation facilities are compared with those expected in DEMO. The conclusion is that both IFMIF and TechnoFusión (future triple beam ion accelerator to emulate fusion neutron irradiation effects in materials) facilities are suited to perform relevant irradiation experiments for the design of DEMO.

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

Tritium in fusion reactor components

International Nuclear Information System (INIS)

Watson, J.S.; Fisher, P.W.; Talbot, J.B.

1980-01-01

When tritium is used in a fusion energy experiment or reactor, several implications affect and usually restrict the design and operation of the system and involve questions of containment, inventory, and radiation damage. Containment is expected to be particularly important both for high-temperature components and for those components that are prone to require frequent maintenance. Inventory is currently of major significance in cases where safety and environmental considerations limit the experiments to very low levels of tritium. Fewer inventory restrictions are expected as fusion experiments are placed in more-remote locations and as the fusion community gains experience with the use of tritium. However, the advent of power-producing experiments with high-duty cycle will again lead to serious difficulties based principally on tritium availability; cyclic operations with significant regeneration times are the principal problems

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

Interspecific transfer of only part of genome by fusion between non-irradiated protoplasts of Nicotiana glauca and X-ray irradiated protoplasts of N. Langsdorffii

International Nuclear Information System (INIS)

Itoh, K.; Futsuhara, Y.

1983-01-01

To transfer only part of genome, X-ray irradiated suspension cell protoplasts of N. langsdorffii were fused with suspension cell protoplasts of N. glauca by polyethylene glycol. Somatic hybrid calli were selected by the growth in the hormone-free medium. Some of somatic hybrid calli from fusion with irradiated protoplasts indicated the loss of small subunit polypeptide of fraction 1 protein which was coded by N. langsdorffii nuclear DNA. Cytological analysis provided an information on significant decrease of chromosomes in somatic hybrid calli from fusion with irradiated protoplasts, compared with the somatic hybrid calli from fusion with non-irradiated protoplasts. In addition, isozyme analysis revealed that somatic hybrid calli from fusion with irradiated protoplasts lost particular bands of N. langsdorffli. These results demonstrate the tranfer of only part of genome from N, langsdorffii to N, glauca by fusion with X-ray irradiated protoplasts

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

Microstructural origins of yield-strength changes in AISI 316 during fission or fusion irradiation

International Nuclear Information System (INIS)

Garner, F.A.; Hamilton, M.L.; Panayotou, N.F.; Johnson, G.D.

1981-08-01

The changes in yield strength of AISI 316 irradiated in breeder reactors have been successfully modeled in terms of concurrent changes in microstructural components. Two new insights involving the strength contributions of voids and Frank loops have been incorporated into the hardening models. Both the radiation-induced microstructure and the yield strength exhibit transients which are then followed by saturation at a level dependent on the irradiation temperature. Extrapolation to anticipated fusion behavior based on microstructural comparisons leads to the conclusion that the primary influence of transmutational differences is only to alter the transient behavior and not the saturation level of yield strength

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

International Nuclear Information System (INIS)

Vladimirov, P.; Moeslang, A.

2004-01-01

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

Material synergism fusion-fission

International Nuclear Information System (INIS)

Sankara Rao, K.B.; Raj, B.; Cook, I.; Kohyama, A.; Dudarev, S.

2007-01-01

In fission and fusion reactors the common features such as operating temperatures and neutron exposures will have the greatest impact on materials performance and component lifetimes. Developing fast neutron irradiation resisting materials is a common issue for both fission and fusion reactors. The high neutron flux levels in both these systems lead to unique materials problems like void swelling, irradiation creep and helium embitterment. Both fission and fusion rely on ferritic-martensitic steels based on 9%Cr compositions for achieving the highest swelling resistance but their creep strength sharply decreases above ∝ 823K. The use of oxide dispersion strengthened (ODS) alloys is envisaged to increase the operating temperature of blanket systems in the fusion reactors and fuel clad tubes in fast breeder reactors. In view of high operating temperatures, cyclic and steady load conditions and the long service life, properties like creep, low cycle fatigue,fracture toughness and creepfatigue interaction are major considerations in the selection of structural materials and design of components for fission and fusion reactors. Currently, materials selection for fusion systems has to be based upon incomplete experimental database on mechanical properties. The usage of fairly well developed databases, in fission programmes on similar materials, is of great help in the initial design of fusion reactor components. Significant opportunities exist for sharing information on technology of irradiation testing, specimen miniaturization, advanced methods of property measurement, safe windows for metal forming, and development of common materials property data base system. Both fusion and fission programs are being directed to development of clean steels with very low trace and tramp elements, characterization of microstructure and phase stability under irradiation, assessment of irradiation creep and swelling behaviour, studies on compatibility with helium and developing

Precipitation response of austenitic stainless steel to simulated fusion irradiation

International Nuclear Information System (INIS)

Maziasz, P.J.

1978-01-01

The precipitation response of annealed type 316 stainless steel irradiated in HFIR is studied and compared to previously observed thermal aging and fast reactor irradiation responses. Irradiation in HFIR simultaneously produces high levels of helium and displacement damage and partially simulates a fusion environment. Samples have been irradiated at temperatures from 550 to 680 0 C to fluences producing up to 3300 appm He and 47 dpa

Development of radiation fusion technology with food technology by the application of high dose irradiation

International Nuclear Information System (INIS)

Lee, Juwoon; Kim, Jaehun; Choi, Jongil

2012-04-01

This study was performed to achieve stable food supply and food safety with radiation fusion technology as a preparation for food weaponization. Results at current stage are following: First, for the development of radiation and food engineering fusion technology using high dose irradiation, the effects of high dose irradiation on food components were evaluated. The combination treatment of irradiation with food engineering was developed. Irradiation condition to destroy radiation resistant foodborne bacteria were determined. Second, for the development of E-beam irradiation technology, the effects of radiation sources on food compounds, processing conditions, and food quality of final products were compared. Food processing conditions for agricultural/aquatic products with different radiation sources was developed and the domination of E-beam irradiation foods were determined. The physical marker for E-beam irradiated foods or not was developed. Third, for the fundamental researches to develop purposed foods to extreme environmental, ready-to-eat foods were developed using high dose irradiation. Food processing for export strategy foods such as process ginseng were developed. Food processing with irradiation to destroy mycotoxin and to inhibit production of mycotoxin was developed. Mathematical models to predict necessary irradiation doses and radiation sources were developed and validated. Through the fundamental researches, the legislation for irradiation approval on meat products, sea foods and dried sea foods, and use of E-beam was introduced. Results from this research project, the followings are expected. Improvement of customer acceptance and activation of irradiation technology by the use of various irradiation rays. Increase of indirect food productivity, and decrease of SOC and improvement of public health by prevention of foodborne outbreaks. Build of SPS/TBT system against imported products and acceleration of domestic product export. Systemized

Development of Radiation Fusion Technology with Food Technology by the Application of High Dose Irradiation

International Nuclear Information System (INIS)

Kim, Ju Won; Kim, Jae Hun; Choi, Jong Il

2010-04-01

This study was studied to achieve stable food supply and food safety with radiation fusion technology as a preparation for food weaponization. Results at current stage are following: First, for the development of radiation and food engineering fusion technology using high dose irradiation, the effects of high dose irradiation on food components were evaluated. The combination treatment of irradiation with food engineering were developed. Irradiation condition to destroy radiation resistant food borne bacteria were determined. Second, for the development of E-beam irradiation technology, the effects of radiation sources on food compounds, processing conditions, and food quality of final products were compared. Food processing conditions for agricultural/aquatic products with different radiation sources were developed and the domination of E-beam irradiation foods were determined. The physical marker for E-beam irradiated foods or not were developed. Third, for the fundamental researches to develop purposed foods to extreme environmental, ready-to-eat foods were developed using high dose irradiation. Food processing for export strategy foods such as process ginseng were developed. Food processing with irradiation to destroy mycotoxin and to inhibit production of mycotoxin were developed. Mathematical models to predict necessary irradiation doses and radiation sources were developed and validated. Through the fundamental researches, the legislation for irradiation approval on meat products, sea foods and dried sea foods, and use of E-beam were introduced. Results from this research project, the followings are expected. (1) Improvement of customer acceptance and activation of irradiation technology by the use of various irradiation rays. (2) Increase of indirect food productivity, and decrease of SOC and improvement of public health by prevention of food borne outbreaks. (3) Build of SPS/TBT system against imported products and acceleration of domestic product export

Development of Radiation Fusion Technology with Food Technology by the Application of High Dose Irradiation

Energy Technology Data Exchange (ETDEWEB)

Kim, Ju Won; Kim, Jae Hun; Choi, Jong Il

2010-04-15

This study was studied to achieve stable food supply and food safety with radiation fusion technology as a preparation for food weaponization. Results at current stage are following: First, for the development of radiation and food engineering fusion technology using high dose irradiation, the effects of high dose irradiation on food components were evaluated. The combination treatment of irradiation with food engineering were developed. Irradiation condition to destroy radiation resistant food borne bacteria were determined. Second, for the development of E-beam irradiation technology, the effects of radiation sources on food compounds, processing conditions, and food quality of final products were compared. Food processing conditions for agricultural/aquatic products with different radiation sources were developed and the domination of E-beam irradiation foods were determined. The physical marker for E-beam irradiated foods or not were developed. Third, for the fundamental researches to develop purposed foods to extreme environmental, ready-to-eat foods were developed using high dose irradiation. Food processing for export strategy foods such as process ginseng were developed. Food processing with irradiation to destroy mycotoxin and to inhibit production of mycotoxin were developed. Mathematical models to predict necessary irradiation doses and radiation sources were developed and validated. Through the fundamental researches, the legislation for irradiation approval on meat products, sea foods and dried sea foods, and use of E-beam were introduced. Results from this research project, the followings are expected. (1) Improvement of customer acceptance and activation of irradiation technology by the use of various irradiation rays. (2) Increase of indirect food productivity, and decrease of SOC and improvement of public health by prevention of food borne outbreaks. (3) Build of SPS/TBT system against imported products and acceleration of domestic product export

Development of radiation fusion technology with food technology by the application of high dose irradiation

Energy Technology Data Exchange (ETDEWEB)

Lee, Juwoon; Kim, Jaehun; Choi, Jongil; and others

2012-04-15

This study was performed to achieve stable food supply and food safety with radiation fusion technology as a preparation for food weaponization. Results at current stage are following: First, for the development of radiation and food engineering fusion technology using high dose irradiation, the effects of high dose irradiation on food components were evaluated. The combination treatment of irradiation with food engineering was developed. Irradiation condition to destroy radiation resistant foodborne bacteria were determined. Second, for the development of E-beam irradiation technology, the effects of radiation sources on food compounds, processing conditions, and food quality of final products were compared. Food processing conditions for agricultural/aquatic products with different radiation sources was developed and the domination of E-beam irradiation foods were determined. The physical marker for E-beam irradiated foods or not was developed. Third, for the fundamental researches to develop purposed foods to extreme environmental, ready-to-eat foods were developed using high dose irradiation. Food processing for export strategy foods such as process ginseng were developed. Food processing with irradiation to destroy mycotoxin and to inhibit production of mycotoxin was developed. Mathematical models to predict necessary irradiation doses and radiation sources were developed and validated. Through the fundamental researches, the legislation for irradiation approval on meat products, sea foods and dried sea foods, and use of E-beam was introduced. Results from this research project, the followings are expected. Improvement of customer acceptance and activation of irradiation technology by the use of various irradiation rays. Increase of indirect food productivity, and decrease of SOC and improvement of public health by prevention of foodborne outbreaks. Build of SPS/TBT system against imported products and acceleration of domestic product export. Systemized

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

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

Method for mounting laser fusion targets for irradiation

Science.gov (United States)

Fries, R. Jay; Farnum, Eugene H.; McCall, Gene H.

1977-07-26

Methods for preparing laser fusion targets of the ball-and-disk type are disclosed. Such targets are suitable for irradiation with one or two laser beams to produce the requisite uniform compression of the fuel material.

Performance of Shiva as a laser fusion irradiation facility

International Nuclear Information System (INIS)

Speck, D.R.; Bliss, E.S.; Glaze, J.A.; Johnson, B.C.; Manes, K.R.; Ozarski, R.G.; Rupert, P.R.; Simmons, W.W.; Swift, C.D.; Thompson, C.E.

1979-01-01

Shiva is a 20 beam Nd:Glass Laser and Target Irradiation Facility at the Lawrence Livermore Laboratory. The laser system and integrated target facility evolved during the last year from a large, untested, experimental laser system to a target irradiation facility which has provided significant laser driven inertial confinement fusion data. The operation of the facility is discussed

Recommendations for the use of irradiated components

International Nuclear Information System (INIS)

2007-01-01

The disease 'graft-versus-host' associated with the transfusion (EIVH TA) is an adverse reaction rare but fatal, linked to the proliferation of T cells that are found in cellular components and reacting against the receptor's tissues). Gamma irradiation of cellular components is used as a prevention method because it deactivates the lymphocytes T by reducing its survival and by restraining its proliferation without producing alterations in others cells function. Recommendations for the use of gamma irradiation along with clinical indications for pediatric patients, patients with acquired immunosuppression and immunocompetent patients are given in this study. A brief description of operative aspects of irradiation procedures such as components to be irradiated, irradiation method, irradiation dose and viability of irradiated components is given [es

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2002-01-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

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

Inertial fusion energy; L'energie de fusion inertielle

Energy Technology Data Exchange (ETDEWEB)

Decroisette, M.; Andre, M.; Bayer, C.; Juraszek, D. [CEA Bruyeres-le-Chatel, Dir. des Systemes d' Information (CEA/DIF), 91 (France); Le Garrec, B. [CEA Centre d' Etudes Scientifiques et Techniques d' Aquitaine, 33 - Le Barp (France); Deutsch, C. [Paris-11 Univ., 91 - Orsay (France); Migus, A. [Institut d' Optique Centre scientifique, 91 - Orsay (France)

2005-07-01

We first recall the scientific basis of inertial fusion and then describe a generic fusion reactor with the different components: the driver, the fusion chamber, the material treatment unit, the target factory and the turbines. We analyse the options proposed at the present time for the driver and for target irradiation scheme giving the state of art for each approach. We conclude by the presentation of LMJ (laser Megajoule) and NIF (national ignition facility) projects. These facilities aim to demonstrate the feasibility of laboratory DT ignition, first step toward Inertial Fusion Energy. (authors)

BCR-ABL fusion genes are inducible by X-irradiation in vitro

International Nuclear Information System (INIS)

Ito, Takashi; Seyama, Toshio; Mizuno, Terumi; Hayashi, Tomonori; Nakamura, Nori; Akiyama, Mitoshi; Dohi, Kiyohiko.

1992-01-01

The Philadelphia chromosome consists of a reciprocal translocation between the ABL oncogene at chromosome 9q34 and the BCR gene at chromosome 22q resulting in the expression of chimeric BCR-ABL mRNAs specific to chronic myelogenous leukemia (CML). The presence of the fusion genes can be detected with high specificity and sensitivity by means of reverse transcription and polymerase chain reaction. Using this assay, it was possible to detect BCR-ABL fusion genes induced among HL60 cells after 100 Gy of X-irradiation in vitro. A total of five fusion gene transcripts were obtained. These fusion genes contained not only CML-specific BCR-ABL rearrangements, but also other forms of BCR-ABL fusions. These latter genes had junctions of BCR exon 4/ABL exon 2 intervened by a segment of DNA of unknown origin, BCR exon 5/ABL exon 2, and BCR exon 4/ABL exon 2. The results appear to be the first evidence for the induction of the BCR-ABL fusion gene by X-irradiation. In terms of leukemogenesis, it is suggested that only those cells bearing certain CML-related BCR-ABL fusion genes are positively selected by virtue of a growth advantage in vivo. (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

Ion beam irradiation of ceramics at fusion relevant conditions

International Nuclear Information System (INIS)

Zinkle, S.J.

1991-01-01

Ceramic materials are required at a variety of locations in proposed fusion reactors where significant ionizing and displacive fields may be present. Energetic ion beams are a useful tool for probing the effects of irradiation on the structure and electrical properties of ceramics over a wide range of experimental conditions. The advantages and disadvantages of using ion beams to provide information on anticipated ceramic radiation effects in a fusion reactor environment are discussed. In this paper particular emphasis is placed on microstructural changes and how the high helium generation rates associated with DT fusion neutrons affect cavity swelling

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.

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2002-04-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

Behavior of Type 316 stainless steel under simulated fusion reactor irradiation

International Nuclear Information System (INIS)

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

1978-05-01

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

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

Inertial fusion energy

International Nuclear Information System (INIS)

Decroisette, M.; Andre, M.; Bayer, C.; Juraszek, D.; Le Garrec, B.; Deutsch, C.; Migus, A.

2005-01-01

We first recall the scientific basis of inertial fusion and then describe a generic fusion reactor with the different components: the driver, the fusion chamber, the material treatment unit, the target factory and the turbines. We analyse the options proposed at the present time for the driver and for target irradiation scheme giving the state of art for each approach. We conclude by the presentation of LMJ (laser Megajoule) and NIF (national ignition facility) projects. These facilities aim to demonstrate the feasibility of laboratory DT ignition, first step toward Inertial Fusion Energy. (authors)

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

A comparison of microstructures in copper irradiated with fission, fusion, and spallation neutrons

International Nuclear Information System (INIS)

Muroga, T.; Heinisch, H.L.; Sommer, W.F.; Ferguson, P.D.

1992-01-01

The objective of this work is to investigate the effects of the neutron energy spectrum in low dose irradiations on the microstructure and mechanical properties of metals. The microstructures of pure copper irradiated to low doses at 36-90 C with spallation neutrons, fusion neutrons and fission neutrons are compared. The defect cluster densities for the spallation and fusion neutrons are very similar when compared on the basis of displacements per atom (dpa). In both cases, the density increases in proportion to the square root of the dpa. The difference in defect density between fusion neutrons and fission neutrons corresponds with differences observed in data on yield stress changes

Materials for fusion reactors

International Nuclear Information System (INIS)

Ehrlich, K.; Kaletta, D.

1978-03-01

The following report describes five papers which were given during the IMF seminar series summer 1977. The purpose of this series was to discuss especially the irradiation behaviour of materials intended for the first wall of future fusion reactors. The first paper deals with the basic understanding of plasma physics relating to the fusion reactor and presents the current state of art of fusion technology. The next two talks discuss the metals intended for the first wall and structural components of a fusion reactor. Since 14 MeV neutrons play an important part in the process of irradiation damage their role is discussed in detail. The question which machines are presently available to simulate irradiation damage under conditions similar to the ones found in a fusion reactor are investigated in the fourth talk which also presents the limitations of the different methods of simulation. In this context also discussed is the importance future intensive neutron sources and materials test reactors will have for this problem area. The closing paper has as a theme the review of the present status of research of metallic and non-metallic materials in view of the quite different requirements for different fusion systems; a closing topic is the world supply on rare materials required for fusion reactors. (orig) [de

Vacuum Plasma Spraying W-coated Reduced Activation Structural Steels for Fusion Plasma Facing Components

Energy Technology Data Exchange (ETDEWEB)

Noh, Sanghoon; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Tungsten (W) and its alloys are considered as candidate materials for plasma facing materials of the first wall and diverter components in fusion reactor systems because of high sputtering resistance and low tritium retention in a fusion environment. Therefore, it is considered that the joining between W and reduced activation structural steels, and its evaluation, are critical issues for the development of fusion reactors. However, the joining between these materials is a very challenging process because of significant differences in their physical properties, particularly the mismatch of coefficients of thermal expansion (CTE). For instance, the CTE of pure W is known to be about 4.3Χ10{sup -6}K{sup -1}; however, that of martensitic steels reaches over three times, about 12-14Χ10{sup -6}K{sup -1} at room temperature even up to 373K. Nevertheless, several joining techniques have been developed for joining between W and structural steels, such as a vapor deposition method, brazing and diffusion bonding. Meanwhile, vacuum plasma spraying (VPS) is supposed to be one of the prospective methods to fabricate a sufficient W layer on the steel substrates because of the coating of a large area with a relatively high fabricating rate. In this study, the VPS method of W powders on reduced activation steels was employed, and its microstructure and hardness distribution were investigated. ODS ferritic steels and F82H steel were coated by VPS-W, and the microstructure and hardness distribution were investigated. A microstructure analysis revealed that pure W was successfully coated on steel substrates by the VPS process without an intermediate layer, in spite of a mismatch of the CTE between dissimilar materials. After neutron irradiation, irradiation hardening significantly occurred in the VPSW. However, the hardening of VPS-W was lesser than that of bulk W irradiated HFIR at 773K. Substrate materials, ODS ferritic steels, and F82H steel, did not show irradiation hardening

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

Numerical analysis of magnetoelastic coupled buckling of fusion reactor components

International Nuclear Information System (INIS)

Demachi, K.; Yoshida, Y.; Miya, K.

1994-01-01

For a tokamak fusion reactor, it is one of the most important subjects to establish the structural design in which its components can stand for strong magnetic force induced by plasma disruption. A number of magnetostructural analysis of the fusion reactor components were done recently. However, in these researches the structural behavior was calculated based on the small deformation theory where the nonlinearity was neglected. But it is known that some kinds of structures easily exceed the geometrical nonlinearity. In this paper, the deflection and the magnetoelastic buckling load of fusion reactor components during plasma disruption were calculated

The impact of pulsed irradiation upon neutron activation calculations for inertial and magnetic fusion energy power plants

International Nuclear Information System (INIS)

Latkowski, J.F.; Sanz, J.; Vujic, J.L.

1996-01-01

Inertial fusion energy (IFE) and magnetic fusion energy (MFE) power plants will probably operate in a pulsed mode. The two different schemes, however, will have quite different time periods. Typical repetition rates for IFE power plants will be 1-5 Hz. MFE power plants will ramp up in current for about 1 hour, shut down for several minutes, and repeat the process. Traditionally, activation calculations for IFE and MFE power plants have assumed continuous operation and used either the ''steady state'' (SS) or ''equivalent steady state'' (ESS) approximations. It has been suggested recently that the SS and ESS methods may not yield accurate results for all radionuclides of interest. The present work expands that of Sisolak, et al. by applying their formulae to conditions which might be experienced in typical IFE and MFE power plants. In addition, complicated, multi-step reaction/decay chains are analyzed using an upgraded version of the ACAB radionuclide generation/depletion code. Our results indicate that the SS method is suitable for application to MFE power plant conditions. We also find that the ESS method generates acceptable results for radionuclides with half-lives more than a factor of three greater than the time between pulses. For components that are subject to 0.05 Hz (or more frequent) irradiation (such as coolant), use of the ESS method is recommended. For components or materials that are subject to less frequent irradiation (such as high-Z target materials), pulsed irradiation calculations should be used

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.

Ion irradiated graphite exposed to fusion-relevant deuterium plasma

International Nuclear Information System (INIS)

Deslandes, Alec; Guenette, Mathew C.; Corr, Cormac S.; Karatchevtseva, Inna; Thomsen, Lars; Ionescu, Mihail; Lumpkin, Gregory R.; Riley, Daniel P.

2014-01-01

Graphite samples were irradiated with 5 MeV carbon ions to simulate the damage caused by collision cascades from neutron irradiation in a fusion environment. The ion irradiated graphite samples were then exposed to a deuterium plasma in the linear plasma device, MAGPIE, for a total ion fluence of ∼1 × 10 24 ions m −2 . Raman and near edge X-ray absorption fine structure (NEXAFS) spectroscopy were used to characterize modifications to the graphitic structure. Ion irradiation was observed to decrease the graphitic content and induce disorder in the graphite. Subsequent plasma exposure decreased the graphitic content further. Structural and surface chemistry changes were observed to be greatest for the sample irradiated with the greatest fluence of MeV ions. D retention was measured using elastic recoil detection analysis and showed that ion irradiation increased the amount of retained deuterium in graphite by a factor of four

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.

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

The PIREX proton irradiation facility

Energy Technology Data Exchange (ETDEWEB)

Victoria, M. [Association EURATOM, Villigen (Switzerland)

1995-10-01

The proton Irradiation Experiment (PIREX) is a materials irradiation facility installed in a beam line of the 590 MeV proton accelerator at the Paul Scherrer Institute. Its main purpose is the testing of candidate materials for fusion reactor components. Protons of this energy produce simultaneously displacement damage and spallation products, amongst them helium and can therefore simulate any possible synergistic effects of damage and helium, that would be produced by the fusion neutrons.

The PIREX proton irradiation facility

International Nuclear Information System (INIS)

Victoria, M.

1995-01-01

The proton Irradiation Experiment (PIREX) is a materials irradiation facility installed in a beam line of the 590 MeV proton accelerator at the Paul Scherrer Institute. Its main purpose is the testing of candidate materials for fusion reactor components. Protons of this energy produce simultaneously displacement damage and spallation products, amongst them helium and can therefore simulate any possible synergistic effects of damage and helium, that would be produced by the fusion neutrons

Irradiation of electronic components and circuits at the Portuguese Research Reactor: Lessons learned

Energy Technology Data Exchange (ETDEWEB)

Marques, J.G.; Ramos, A.R.; Fernandes, A.C.; Santos, J.P. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade de Lisboa, Estrada Nacional 10, 2695-066 Bobadela LRS (Portugal)

2015-07-01

The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, since the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor

High temperature superconductors for fusion magnets -influence of neutron irradiation

International Nuclear Information System (INIS)

Chudy, M.; Eisterer, M.; Weber, H. W.

2010-01-01

In this work authors present the results of study of influence of neutron irradiation of high temperature superconductors for fusion magnets. High temperature superconductors (type of YBCO (Yttrium-Barium-Copper-Oxygen)) are strong candidates to be applied in the next step of fusion devices. Defects induced by fast neutrons are effective pinning centres, which can significantly improve critical current densities and reduce J c anisotropy. Due to induced lattice disorder, T c is reduced. Requirements for ITER (DEMO) are partially achieved at 64 K.

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

International Nuclear Information System (INIS)

Yagi, Toshiaki; Morita, Yousuke; Seguchi, Tadao

1995-03-01

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

The effect of low energy helium ion irradiation on tungsten-tantalum (W-Ta) alloys under fusion relevant conditions

Science.gov (United States)

Gonderman, S.; Tripathi, J. K.; Novakowski, T. J.; Sizyuk, T.; Hassanein, A.

2017-08-01

Currently, tungsten remains the best candidate for plasma-facing components (PFCs) for future fusion devices because of its high melting point, low erosion, and strong mechanical properties. However, continued investigation has shown tungsten to undergo severe morphology changes under fusion-like conditions. These results motivate the study of innovative PFC materials which are resistant to surface morphology evolution. The goal of this work is to examine tungsten-tantalum (W-Ta) alloys, a potential PFC material, and their response to low energy helium ion irradiation. Specifically, W-Ta samples are exposed to 100 eV helium irradiations with a flux of 1.15 × 1021 ions m-2 s-1, at 873 K, 1023 K, and 1173 K for 1 h duration. Scanning electron microscopy (SEM) reveals significant changes in surface deterioration due to helium ion irradiation as a function of both temperature and tantalum concentration in W-Ta samples. X-Ray Diffraction (XRD) studies show a slight lattice parameter expansion in W-Ta alloy samples compared to pure W samples. The observed lattice parameter expansion in W-Ta alloy samples (proportional to increasing Ta wt.% concentrations) reflect significant differences observed in the evolution of surface morphology, i.e., fuzz development processes for both increasing Ta wt.% concentration and target temperature. These results suggest a correlation between the observed morphology differences and the induced crystal structure change caused by the presence of tantalum. Shifts in the XRD peaks before and after 100 eV helium irradiation with a flux of 1.15 × 1021 ions m-2 s-1, 1023 K, for 1 h showed a significant difference in the magnitude of the shift. This has suggested a possible link between the atomic spacing of the material and the accumulated damage. Ongoing research is needed on W-Ta alloys and other innovative materials for their application as irradiation resistant materials in future fusion or irradiation environments.

Positron annihilation lifetime measurements of vanadium alloy and F82H irradiated with fission and fusion neutrons

International Nuclear Information System (INIS)

Sato, K.; Inoue, K.; Yoshiie, T.; Xu, Q.; Wakai, E.; Kutsukake, C.; Ochiai, K.

2009-01-01

V-4Cr-4Ti, F82H, Ni and Cu were irradiated with fission and fusion neutrons at room temperature and 473 K. Defect structures were analyzed and compared using positron annihilation lifetime measurement, and microstructural evolution was discussed. The mean lifetime of positrons (the total amount of residual defects) increased with the irradiation dose. The effect of cascade impact was detected in Ni at room temperature. The size and the number of vacancy clusters were not affected by the displacement rate in the fission neutron irradiation at 473 K for the metals studied. The vacancy clusters were not formed in V-4Cr-4Ti irradiated at 473 K in the range of 10 -6 -10 -3 dpa. In F82H irradiated at 473 K, the defect evolution was prevented by pre-existing defects. The mean lifetime of positrons in fission neutron irradiation was longer than that in fusion neutron irradiation in V-4Cr-4Ti at 473 K. It was interpreted that more closely situated subcascades were formed in the fusion neutron irradiation and subcascades interacted with each other, and consequently the vacancy clusters did not grow larger.

An unexpected response of torulopsis glabrata fusion products to x-irradiation

International Nuclear Information System (INIS)

Galeotti, C.L.; Sriprakash, K.S.; Batum, C.M.; Clark-Walker, G.D.

1981-01-01

Intra-species fusion products of Saccharomyces cerevisiae, Saccharomyces unisporus and Torulopsis glabrata have been isolated following polyethylene glycol-induced fusion of protoplasts and selection for prototrophic colonies. Staining with lomofungin showed that all fusion products were uninucleate. Measurement of DNA content mostly gave values between haploid and diploid levels indicating that the majority of fusion products were aneuploid. Nevertheless fusion products of S. cerevisiae and S. unisporus were, as expected, more resistant to X-irradiation than their haploid parents. By contrast, the X-ray doze-response curve of all T. glabrata fusion products was indistinguishable from their progenitors despite the fact that mitotic segregants could be recovered amongst the survivors to X-rays. A possible explanation for the behaviour towards X-rays of T. glabrata fusion products is that this species lacks a DNA repair pathway involving recombination between homologous chromosomes. We conclude from this study that the shape of the X-ray dose-response curve should not be taken to indicate the ploidy of new yeast isolates without supporting data. (orig.)

Detection of irradiated strawberries by identifying ESR peak of irradiated cellulose component

International Nuclear Information System (INIS)

Goto, Michiko; Tanabe, Hiroko

2002-01-01

The method of detecting low-dose irradiated strawberries by identifying ESR peak of irradiated cellulose component was studied. Ratio of peak height (S) of high magnetic field cellulose component, and noise width (N) of either irradiated or unirradiated seeds of strawberries were compared. In this study, sample was identified to be irradiated when S/N ratio of ESR spectrum of 4 min. sweep time was above 0.7. In the case of S/N ratio below 0.7, when the S/N ratio of integrated ESR spectrum, obtained from measuring 10 times with 1 min. sweep time was above 1.0, the sample was identified to be irradiated. The result suggests that S/N ratio is a good marker to detect the irradiation. The strawberries irradiated above 0.5kGy was able to be detected after 3 days storage at room temperature, after 21 days refrigeration and after 60 days freezing, respectively. (author)

3D-printed fusion components concepts and validation for the UST-2 stellarator

International Nuclear Information System (INIS)

Queral, V.

2015-01-01

Highlights: • A fabrication method for fusion components is developed and validated. • Synergies obtained from additive manufacturing and non-metal casting. • 3D-printed polyamide hollow truss structure and casting of acrylic resin tested. • UST- 2 stellarator coil frame fabricated to validate the method performance. - Abstract: The geometric complexity and high accuracy simultaneously required in magnetic fusion devices, particularly stellarators and tokamaks, hampers the production of fusion components and devices. Rapid manufacturing construction methods, particularly enhanced for fusion, may contribute to a faster cycle and lower cost production of certain components for tokamaks and stellarators. Casting, cutting, forming, welding and mechanising are conventional production techniques for major fusion components, i.e. coil casings, coil frames, vacuum vessels and blankets. Synergies may emerge by combination of additive manufacturing (3D printing) with conventional manufacturing methods. 3D printing combined with resin moulding is tested by construction of the coil frame and the vacuum vessel of a small stellarator, UST-2. Satisfactory coil frames have been obtained by moulding acrylic resin in a special 3D printed polyamide hollow three-dimensional structure. The conceptual engineering design, construction process and validation of the components are described. The presented manufacturing method might contribute to advance the future 3D printing of larger metallic components for fusion.

3D-printed fusion components concepts and validation for the UST-2 stellarator

Energy Technology Data Exchange (ETDEWEB)

Queral, V., E-mail: vicentemanuel.queral@ciemat.es

2015-10-15

Highlights: • A fabrication method for fusion components is developed and validated. • Synergies obtained from additive manufacturing and non-metal casting. • 3D-printed polyamide hollow truss structure and casting of acrylic resin tested. • UST- 2 stellarator coil frame fabricated to validate the method performance. - Abstract: The geometric complexity and high accuracy simultaneously required in magnetic fusion devices, particularly stellarators and tokamaks, hampers the production of fusion components and devices. Rapid manufacturing construction methods, particularly enhanced for fusion, may contribute to a faster cycle and lower cost production of certain components for tokamaks and stellarators. Casting, cutting, forming, welding and mechanising are conventional production techniques for major fusion components, i.e. coil casings, coil frames, vacuum vessels and blankets. Synergies may emerge by combination of additive manufacturing (3D printing) with conventional manufacturing methods. 3D printing combined with resin moulding is tested by construction of the coil frame and the vacuum vessel of a small stellarator, UST-2. Satisfactory coil frames have been obtained by moulding acrylic resin in a special 3D printed polyamide hollow three-dimensional structure. The conceptual engineering design, construction process and validation of the components are described. The presented manufacturing method might contribute to advance the future 3D printing of larger metallic components for fusion.

Irradiation damage of ferritic/martensitic steels: Fusion program data applied to a spallation neutron source

International Nuclear Information System (INIS)

Klueh, R.L.

1997-01-01

Ferritic/martensitic steels were chosen as candidates for future fusion power plants because of their superior swelling resistance and better thermal properties than austenitic stainless steels. For the same reasons, these steels are being considered for the target structure of a spallation neutron source, where the structural materials will experience even more extreme irradiation conditions than expected in a fusion power plant first wall (i.e., high-energy neutrons that produce large amounts of displacement damage and transmutation helium). Extensive studies on the effects of neutron irradiation on the mechanical properties of ferritic/martensitic steels indicate that the major problem involves the effect of irradiation on fracture, as determined by a Charpy impact test. There are indications that helium can affect the impact behavior. Even more helium will be produced in a spallation neutron target material than in the first wall of a fusion power plant, making helium effects a prime concern for both applications. 39 refs., 10 figs

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.)

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

International Nuclear Information System (INIS)

Gelles, D.S.

1986-07-01

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

Activation analysis of tritium breeder lithium lead irradiated by fusion neutrons in FDS-II

International Nuclear Information System (INIS)

Mingliang Chen

2006-01-01

R-and-D of fusion materials, especially their activation characteristics, is one of the key issues for fusion research in the world. Research on activation characteristics for low activation materials, such as reduced activation ferritic/martensitic steels, vanadium alloys and SiCf/SiC composites, is being done throughout the world to ensure the attractiveness of fusion power regarding safety and environmental aspects. However, there is less research on the activation characteristics of the other important fusion materials, such as tritium breeder etc.. Lithium lead (Li 17 Pb 83 ) is presently considered as a primary candidate tritium breeder for fusion power reactors because of its attractive characteristics. It can serve as a tritium breeder, neutron multiplier and coolant in the blanket at the same time. The radioactivity of Li 17 Pb 83 by D-T fusion neutrons in FDS-II has been calculated and analyzed. FDS-II is a concept design of fusion power reactor, which consists of fusion core with advanced plasma parameters extrapolated from the ITER (International Thermonuclear Experimental Reactor) and two candidates of liquid lithium breeder blankets (named SLL and DLL blankets). The neutron transport and activation calculation are carried out based on the one-dimensional model for FDS-II with the home-developed multi-functional code system VisualBUS and the multi-group data library HENDL1.0/MG and European Activation File EAF-99. The effects of irradiation time on the activation characteristics of Li 17 Pb 83 were analyzed and it concludes that the irradiation time has an important effect on the activation level of Li 17 Pb 83 . Furthermore, the results were compared with the activation levels of other tritium breeders, such as Li 4 SiO 4 , Li 2 TiO 3 , Li 2 O and Li etc., under the same irradiation conditions. The dominant nuclides to dose rate and activity of Li 17 Pb 83 were analyzed as well. Tritium generated by Li has a great contribution to the afterheat and

Fiscal year 1976 DT fusion neutron irradiations and dosimetry at the LLL rotating target neutron source

International Nuclear Information System (INIS)

MacLean, S.C.

1977-01-01

The DT fusion neutron irradiation of 319 samples during 19 irradiation periods (beam-on time of more than 1026 hours) is described. Experiments from 24 individuals representing 11 institutions are summarized. The numbers of the UCID dosimetry reports detailing each of the irradiations are given

Common and distinct components in data fusion

DEFF Research Database (Denmark)

Smilde, Age Klaas; Mage, Ingrid; Næs, Tormod

2016-01-01

and understanding their relative merits. This paper provides a unifying framework for this subfield of data fusion by using rigorous arguments from linear algebra. The most frequently used methods for distinguishing common and distinct components are explained in this framework and some practical examples are given...

ANITA-2000, Isotope Inventories from Neutron Irradiation, for Fusion Applications

International Nuclear Information System (INIS)

Cepraga, Dan-Gabriel

2000-01-01

1 - Description of program or function: ANITA-2000 is a code package for the activation characterisation of materials exposed to neutrons in fusion machines. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation, continuous or stepwise. It provides activity, atomic density, decay heat, biological hazard, clearance index and gamma-ray source spectra at shut down and for different cooling times. An interactive utility module, MODBIN, to produce the neutron activation cross sections libraries in the required binary ANITA-4M Format, is also included. The GRANITA interactive module may plot activation parameters as a function of the cooling time. The main improvements include: -the number of irradiation time intervals has been increased to 2000; -different neutron wall loading can be used for each burn time interval; -the photon source calculation in the 18 energy group Scale structure has been added; -the clearance index can be calculated. In addition the code language has been standardized to Fortran '95 - by maintaining the backward compatibility (except for the time/date routines) - so as the same code package can be compiled and run on Unix environment and on PC, both under DOS-Windows and under Linux. 2 - Methods: The mathematical solution of the problem is given in analytical form using recurrence relations. Double precision arithmetic is used

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.

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)

Irradiation properties of T0 chopper components

Energy Technology Data Exchange (ETDEWEB)

Itoh, Shinichi, E-mail: shinichi.itoh@kek.jp [Neutron Science Division, Institute of Materials Structure Science, High Energy Accelerator Research Organization, Tsukuba 305-0801 (Japan); Ueno, Kenji; Ohkubo, Ryuji [Mechanical Engineering Center, Applied Research Laboratory, High Energy Accelerator Research Organization, Tsukuba 305-0801 (Japan); Sagehashi, Hidenori [Neutron Science Division, Institute of Materials Structure Science, High Energy Accelerator Research Organization, Tsukuba 305-0801 (Japan); Funahashi, Yoshisato [Mechanical Engineering Center, Applied Research Laboratory, High Energy Accelerator Research Organization, Tsukuba 305-0801 (Japan); Yokoo, Tetsuya [Neutron Science Division, Institute of Materials Structure Science, High Energy Accelerator Research Organization, Tsukuba 305-0801 (Japan)

2011-10-21

We investigated the irradiation properties of the components of a T0 chopper. The organic materials in the rotor bearing grease, the magnetic fluids in seals, and the rubber in the timing belt, as well as the semiconductor materials in the rotation sensor and motor encoder, were all irradiated with high-energy {gamma}-rays up to 100 kGy. No significant damage that shortens the lifetime of a T0 chopper was observed for the mechanical components. However, the semiconductor components were damaged by the irradiation. For the rotation sensor system detecting the rotor phase, the signal from a marker on the rotor shaft was transmitted outside the shielding by an optical fiber with radiation-proofing and the electrical circuits were removed from the beamline shielding. The lifetime of the motor encoder possibly meets the requirement for the maintenance period of the T0 chopper.

Effects of irradiation on the components of implantable pacemakers

CERN Document Server

Kawamura, S; Kuga, N; Shiba, T; Hirose, T; Fujimoto, H; Toyoshima, T; Hyodo, K; Matoba, M

2003-01-01

The purpose of this study was to examine the effects of irradiation on implantable pacemaker components. The pacemaker was divided into three components: lead wire and electrode, battery, and electrical circuit, and each component was irradiated by X-ray and electron beams, respectively. The pacemaker parameters were measured by both telemetry data of the programmer and directly measured data from the output terminal. The following results were obtained. For the lead wire and electrode, there was no effect on the pacemaker function due to irradiation by X-ray and electron beams. In the case of battery irradiation, there was no change in battery voltage or current up to 236 Gy X-ray dose. In the electrical circuit, the pacemaker reverted to the regular beating rate (fixed-rate mode) immediately after the start of X-ray irradiation, and it continued in this mode during irradiation. In patients with their own heartbeat rhythm, changing to the fixed-rate mode may cause dangerous conditions such as ventricular fib...

Miniature tensile test specimens for fusion reactor irradiation studies

International Nuclear Information System (INIS)

Klueh, R.L.

1985-01-01

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

Development of Fusion Nuclear Technologies and the role of MTR's

International Nuclear Information System (INIS)

Laan, J.G. van der; Schaaf, B. van der

2006-01-01

Fusion power plant operation will strongly depend on the economy and reliability of crucial components, such as first wall modules, tritium breeding blankets and divertors. Their operating temperature shall be high to accomplish high plant efficiency. The materials properties and component fabrication routes shall also assure long reliable operation to minimize plant outage. The components must be fabricated in large quantities based on demonstrations with a limited amount of test beds. Mock-ups and test loops will, through iteration processes, demonstrate the reliable operation under reference thermal-hydraulic conditions. Although 14 MeV neutrons dominate the nuclear conditions near the first wall, neutron transport analyses have shown that large portions of the components near the plasma have to cope with a neutron spectrum resembling a fission core. Present Materials Test Reactors, MTR's, offer fluxes relevant for large parts of the fusion major components. The mixed and fast fission spectra though is not representative for all fusion conditions. The strong point of MTR's is their ability to generate sufficient displacement damage in the materials in a relatively short time. The cores of MTR's provide sufficient space for irradiation of representative cut-outs of components to allow integrated functional and materials tests in a high flux neutron field. The MTR's are the primary test bed for structural and functional fusion relevant materials. The MTR space and dose rates provide a valuable base line for the developments and demonstrations of fusion key components in a neutron field. In recent years the pebble bed assembly, PBA, irradiated in the HFR, Petten, has shown the feasibility of the helium-cooled concept with lithium ceramics and beryllium multiplier pebble beds. The irradiations produce a wealth of process parameters for the control of the tritium release of the pebbles. The PBA packaging, cooling and tritium purging arrangements closely resemble the

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

On use of beryllium in fusion reactors: Resources, impurities and necessity of detritiation after irradiation

Energy Technology Data Exchange (ETDEWEB)

Kolbasov, B.N., E-mail: b.kolbasov@yandex.ru; Khripunov, V.I., E-mail: Khripunov_VI@nrcki.ru; Biryukov, A.Yu.

2016-11-01

Highlights: • Potential needs in Be for fusion power engineering may exceed Be resources. • Be recycling after its operation in a fusion power plant (FPP) seems inevitable. • U impurity in Be seriously impairs environmental properties of fusion power plants. • Upon burial of irradiated Be the main problems are caused by U and {sup 3}H impurities. • Clearance of Be extracted from a FPP is impossible due to U impurity. - Abstract: Worldwide identified resources of beryllium somewhat exceed 80 000 t. Beryllium production in all the countries of the world in 2012 was about 230 t. At the same time, some conceptual designs of fusion power reactors envisage utilization of several hundred tons of this metal. Therefore return of beryllium into the production cycle (recycling) will be necessary. The beryllium ore from some main deposits has uranium content inadmissible for fusion reactors. This fact raises a question on the need to develop and apply an economically acceptable technology for beryllium purification from the uranium. Practically any technological procedure with beryllium used in fusion reactors requires its detritiation. A study of tritium and helium release from irradiated beryllium at different temperatures and rates of temperature increase was performed at Kurchatov Institute.

Ultrafine tungsten as a plasma-facing component in fusion devices: effect of high flux, high fluence low energy helium irradiation

International Nuclear Information System (INIS)

El-Atwani, O.; Gonderman, Sean; Allain, J.P.; Efe, Mert; Klenosky, Daniel; Qiu, Tian; De Temmerman, Gregory; Morgan, Thomas; Bystrov, Kirill

2014-01-01

This work discusses the response of ultrafine-grained tungsten materials to high-flux, high-fluence, low energy pure He irradiation. Ultrafine-grained tungsten samples were exposed in the Pilot-PSI (Westerhout et al 2007 Phys. Scr. T128 18) linear plasma device at the Dutch Institute for Fundamental Energy Research (DIFFER) in Nieuwegein, the Netherlands. The He flux on the tungsten samples ranged from 1.0 × 10 23 –2.0 × 10 24  ions m −2  s −1 , the sample bias ranged from a negative (20–65) V, and the sample temperatures ranged from 600–1500 °C. SEM analysis of the exposed samples clearly shows that ultrafine-grained tungsten materials have a greater fluence threshold to the formation of fuzz by an order or magnitude or more, supporting the conjecture that grain boundaries play a major role in the mechanisms of radiation damage. Pre-fuzz damage analysis is addressed, as in the role of grain orientation on structure formation. Grains of (1 1 0) and (1 1 1) orientation showed only pore formation, while (0 0 1) oriented grains showed ripples (higher structures) decorated with pores. Blistering at the grain boundaries is also observed in this case. In situ TEM analysis during irradiation revealed facetted bubble formation at the grain boundaries likely responsible for blistering at this location. The results could have significant implications for future plasma-burning fusion devices given the He-induced damage could lead to macroscopic dust emission into the fusion plasma. (paper)

Selected component failure rate values from fusion safety assessment tasks

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, L.C.

1998-09-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers.

Selected Component Failure Rate Values from Fusion Safety Assessment Tasks

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, Lee Charles

1998-09-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers.

Selected component failure rate values from fusion safety assessment tasks

International Nuclear Information System (INIS)

Cadwallader, L.C.

1998-01-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers

Fusion Materials Irradiation Test (FMIT) facility lithium system: a design and development status

International Nuclear Information System (INIS)

Brackenbury, P.J.; Bazinet, G.D.; Miller, W.C.

1983-01-01

The design and development of the Fusion Materials Irradiation Test (FMIT) Facility lithium system is outlined. This unique liquid lithium recirculating system, the largest of its kind in the world, is described with emphasis on the liquid lithium target assembly and other important components necessary to provide lithium flow to the target. The operational status and role of the Experimental Lithium System (ELS) in the design of the FMIT lithium system are discussed. Safety aspects of operating the FMIT lithium system in a highly radioactive condition are described. Potential spillage of the lithium is controlled by cell liners, by argon flood systems and by remote maintenance features. Lithium chemistry is monitored and controlled by a side-stream loop, where impurities measured by instruments are collected by hot and cold traps

Fusion Materials Irradiation Test (FMIT) facility lithium system: a design and development status

Energy Technology Data Exchange (ETDEWEB)

Brackenbury, P.J.; Bazinet, G.D.; Miller, W.C.

1983-01-01

The design and development of the Fusion Materials Irradiation Test (FMIT) Facility lithium system is outlined. This unique liquid lithium recirculating system, the largest of its kind in the world, is described with emphasis on the liquid lithium target assembly and other important components necessary to provide lithium flow to the target. The operational status and role of the Experimental Lithium System (ELS) in the design of the FMIT lithium system are discussed. Safety aspects of operating the FMIT lithium system in a highly radioactive condition are described. Potential spillage of the lithium is controlled by cell liners, by argon flood systems and by remote maintenance features. Lithium chemistry is monitored and controlled by a side-stream loop, where impurities measured by instruments are collected by hot and cold traps.

Near term, low cost, 14 MeV fusion neutron irradiation facility for testing the viability of fusion structural materials

Energy Technology Data Exchange (ETDEWEB)

Kulcinski, Gerald L., E-mail: glkulcin@wisc.edu [University of Wisconsin-Madison, Madison, WI (United States); Radel, Ross F. [Phoenix Nuclear Labs LLC, Monona, WI (United States); Davis, Andrew [University of Wisconsin-Madison, Madison, WI (United States)

2016-11-01

For over 50 years, engineers have been looking for an irradiation facility that can provide a fusion reactor appropriate neutron spectrum over a significant volume to test fusion reactor materials that is relatively inexpensive and can be built in a minimum of time. The 14 MeV neutron irradiation facility described here can nearly exactly duplicate the neutron spectrum typical of a DT fusion reactor first wall at damage rates of ≈4 displacements per atom and 40 appm He generated over a 2 l volume per full power year of operation. The projected cost of this multi-beam facility is estimated at ≈$20 million and it can be built in <4 years. A single-beam prototype, funded by the U.S. Department of Energy, is already being built to produce medical isotopes. The neutrons are produced by a 300 keV deuterium beam accelerated into 4 kPa (30 Torr) tritium target. The total tritium inventory is <2 g and <0.1 g of T{sub 2} is consumed per year. The core technology proposed has already been fully demonstrated, and no new plasma physics or materials innovations will be required for the test facility to become operational.

Tensile property changes of metals and irradiated to low doses with fission, fusion and spallation neutrons

International Nuclear Information System (INIS)

Heinisch, H.L.; Hamilton, M.L.; Sommer, W.F.; Ferguson, P.D.

1992-01-01

The objective of this work is to investigate the effects of the neutron energy spectrum in low dose irradiations on the microstructures and mechanical properties of metals. Radiation effects due to low doses of spallation neutrons are compared directly to those produced by fission and fusion neutrons. Yield stress changes of pure Cu, alumina-dispersion-strengthened Cu and AISI 316 stainless steel irradiated at 36-55 C in the Los Alamos Spallation Radiation Effects Facility (LASREF) are compared with earlier results of irradiations at 90 C using 14 MeV D-T fusion neutrons at the Rotating Target Neutron Source and fission reactor neutrons in the Omega West Reactor. At doses up to 0.04 displacements per atom (dpa), the yield stress changes due to the three quite different neutron spectra correlate well on the basis of dpa in the stainless steel and the Cu alloy. However, in pure Cu, the measured yield stress changes due to spallation neutrons were anomalously small and should be verified by additional irradiations. With the exception of pure Cu, the low dose, low temperature experiments reveal no fundamental differences in radiation hardening by fission, fusion or spallation neutrons when compared on the basis of dpa

Pixel-level multisensor image fusion based on matrix completion and robust principal component analysis

Science.gov (United States)

Wang, Zhuozheng; Deller, J. R.; Fleet, Blair D.

2016-01-01

Acquired digital images are often corrupted by a lack of camera focus, faulty illumination, or missing data. An algorithm is presented for fusion of multiple corrupted images of a scene using the lifting wavelet transform. The method employs adaptive fusion arithmetic based on matrix completion and self-adaptive regional variance estimation. Characteristics of the wavelet coefficients are used to adaptively select fusion rules. Robust principal component analysis is applied to low-frequency image components, and regional variance estimation is applied to high-frequency components. Experiments reveal that the method is effective for multifocus, visible-light, and infrared image fusion. Compared with traditional algorithms, the new algorithm not only increases the amount of preserved information and clarity but also improves robustness.

Radiation effects on superconducting fusion magnet components

International Nuclear Information System (INIS)

Weber, H.W.

2011-01-01

Nuclear fusion devices based on the magnetic confinement principle heavily rely on the existence and performance of superconducting magnets and have always significantly contributed to advancing superconductor and magnet technology to their limits. In view of the presently ongoing construction of the tokamak device ITER and the stellerator device Wendelstein 7X and their record breaking parameters concerning size, complexity of design, stored energy, amperage, mechanical and magnetic forces, critical current densities and stability requirements, it is deemed timely to review another critical parameter that is practically unique to these devices, namely the radiation response of all magnet components to the lifetime fluence of fast neutrons and gamma rays produced by the fusion reactions of deuterium and tritium. I will review these radiation effects in turn for the currently employed standard "technical" low temperature superconductors NbTi and Nb 3 Sn, the stabilizing material (Cu) as well as the magnet insulation materials and conclude by discussing the potential of high temperature superconducting materials for future generations of fusion devices, such as DEMO. (author)

Fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1984-01-01

This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

Magnetic fusion energy. Progress report, January--June 1976

International Nuclear Information System (INIS)

Doran, D.G.; Yoshikawa, H.H.

1976-01-01

Brief descriptions are given of progress in the Irradiation Effects Analysis and Mechanical Performance of Magnetic Fusion Energy (MFE) Materials programs and in related programs. The objective of the Irradiation Effects Analysis program is the correlation of effects produced in neutron and charged particle irradiations in order to apply them to fusion reactor environments. Low energy displacement cascades--of intrinsic interest and the least understood component of high energy cascades--are being simulated by computer codes of the dynamical (D), quasi-dynamical (Q-D), and binary collision (BC) types. Fair agreement has been found between D and Q-D for low index focused replacement sequences; substantial differences appeared for a 250 eV high index event. The objective of the Mechanical Performance of MFE Materials program is to establish the effects of fusion reactor irradiation environments on the mechanical properties of candidate first wall materials. A Precision Torsional Creep Apparatus is being developed to permit accelerator studies of irradiation creep and behavior under cyclic conditions. This apparatus has demonstrated the required strain sensitivity, stress control, and thermal stability for long term thermal testing, and that it can be used for cyclic testing

Influence of rhTPO/GM-CSF fusion protein on hemopoiesis in mice irradiated with 60Co γ-rays

International Nuclear Information System (INIS)

Cao Hua; Ge Zhongliang; Zhang Qunwei; Liu Xiuzhen

1999-01-01

Objective: To find a new biological therapy for secondary hematopoietic failure including anemia, infection and hemorrhage after administration of chemotherapeutic drugs etc. Methods: hGM-CSF gene was ligated with hTPO gene isolated from human fetal liver mRNA and a new fusion protein rh TPO/GM-CSF obtained. Results: The new fusion protein could promote recovery of peripheral WBC and PLT of 5.0 Gy irradiated mice. BFU-E, CFU-Meg and CFU-GM in bone marrow of mice after irradiation recovered significantly by treatment with rhTPO/GM-CSF fusion protein for 10 days. Conclusion: These results suggest that the new fusion protein has the biological activity of both hTPO and hGM-CSF simultaneously and can stimulate the proliferation of megakaryocytes and granulocyte progenitors

Ceramics for fusion devices

International Nuclear Information System (INIS)

Clinard, F.W. Jr.

1984-01-01

Ceramics are required for a number of applications in fusion devices, among the most critical of which are magnetic coil insulators, windows for RF heating systems, and structural uses. Radiation effects dominate consideration of candidate materials, although good pre-irradiation properties are a requisite. Materials and components can be optimized by careful control of chemical and microstructural content, and application of brittle material design and testing techniques. Future directions for research and development should include further extension of the data base in the areas of electrical, structural, and thermal properties; establishment of a fission neutron/fusion neutron correlation including transmutation gas effects; and development of new materials tailored to meet the specific needs of fusion reactors

Irradiation effects on plasma diagnostic components

International Nuclear Information System (INIS)

Nishitani, Takeo; Iida, Toshiyuki; Ikeda, Yujiro

1998-10-01

One of the most important issues to develop the diagnostics for the experimental thermonuclear reactor such as ITER is the irradiation effects on the diagnostics components. Typical neutron flux and fluence on the first wall are 1 MW/m 2 and 1 MWa/m 2 , respectively for ITER. In such radiation condition, most of the present diagnostics could not survive so that those will be planed to be installed far from the vacuum vessel. However, some diagnostics sensors such as bolometers and magnetic probes still have to be install inside vessel. And many transmission components for lights, wave and electric signals are inevitable even inside vessel. As a part of this R and D program of the ITER Engineering Design Activities (EDA), we carried out the irradiation tests on the basic materials of the transmission components and in-vessel diagnostics sensors in order to identify radiation hardened materials that can be used for diagnostic systems. (J.P.N.)

Irradiation effects on plasma diagnostic components

Energy Technology Data Exchange (ETDEWEB)

Nishitani, Takeo [ed.] [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Iida, Toshiyuki; Ikeda, Yujiro [and others

1998-10-01

One of the most important issues to develop the diagnostics for the experimental thermonuclear reactor such as ITER is the irradiation effects on the diagnostics components. Typical neutron flux and fluence on the first wall are 1 MW/m{sup 2} and 1 MWa/m{sup 2}, respectively for ITER. In such radiation condition, most of the present diagnostics could not survive so that those will be planed to be installed far from the vacuum vessel. However, some diagnostics sensors such as bolometers and magnetic probes still have to be install inside vessel. And many transmission components for lights, wave and electric signals are inevitable even inside vessel. As a part of this R and D program of the ITER Engineering Design Activities (EDA), we carried out the irradiation tests on the basic materials of the transmission components and in-vessel diagnostics sensors in order to identify radiation hardened materials that can be used for diagnostic systems. (J.P.N.)

Tritium-doping enhancement of polystyrene by ultraviolet laser and hydrogen plasma irradiation for laser fusion experiments

Energy Technology Data Exchange (ETDEWEB)

Iwasa, Yuki, E-mail: iwasa-y@ile.osaka-u.ac.jp [Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871 (Japan); Yamanoi, Kohei; Iwano, Keisuke; Empizo, Melvin John F.; Arikawa, Yasunobu; Fujioka, Shinsuke; Sarukura, Nobuhiko; Shiraga, Hiroyuki; Takagi, Masaru; Norimatsu, Takayoshi; Azechi, Hiroshi [Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871 (Japan); Noborio, Kazuyuki; Hara, Masanori; Matsuyama, Masao [Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 3190 Gofuku, Toyama 930-8555 (Japan)

2016-11-15

Highlights: • Tritium-doped polystyrene films are fabricated by the Wilzbach method with UV laser and hydrogen plasma irradiation. • The 266-nm laser-irradiated, 355-nm laser-irradiated, and hydrogen plasma-irradiated polystyrene films exhibit higher PSL intensities and specific radioactivities than the non-irradiated sample. • Tritium doping by UV laser irradiation can be largely affected by the laser wavelength because of polystyrene’s absorption. • Hydrogen plasma irradiation results to a more uniform doping concentration even at low partial pressure and short irradiation time. • UV laser and plasma irradiations can be utilized to fabricate tritium-doped polystyrene shell targets for future laser fusion experiments. - Abstract: We investigate the tritium-doping enhancement of polystyrene by ultraviolet (UV) laser and hydrogen plasma irradiation. Tritium-doped polystyrene films are fabricated by the Wilzbach method with UV laser and hydrogen plasma. The 266-nm laser-irradiated, 355-nm laser-irradiated, and hydrogen plasma-irradiated polystyrene films exhibit higher PSL intensities and specific radioactivities than the non-irradiated sample. Tritium doping by UV laser irradiation can be largely affected by the laser wavelength because of polystyrene’s absorption. In addition, UV laser irradiation is more localized and concentrated at the spot of laser irradiation, while hydrogen plasma irradiation results to a more uniform doping concentration even at low partial pressure and short irradiation time. Both UV laser and plasma irradiations can nevertheless be utilized to fabricate tritium-doped polystyrene targets for future laser fusion experiments. With a high doping rate and efficiency, a 1% tritium-doped polystyrene shell target having 7.6 × 10{sup 11} Bq g{sup −1} specific radioactivity can be obtained at a short period of time thereby decreasing tritium consumption and safety management costs.

IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

International Nuclear Information System (INIS)

2003-03-01

The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m 2 , 20 dpa/y in Fe, in a volume of 500 cm 3 and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

IFMIF (International Fusion Materials Irradiation Facility) key element technology phase interim report

International Nuclear Information System (INIS)

Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki

2002-03-01

Activities of International Fusion Materials Irradiation Facility (IFMIF) have been performed under an IEA collaboration since 1995. IFMIF is an accelerator-based deuteron (D + )-lithium (Li) neutron source designed to produce an intense neutron field (2 MW/m 2 , 20 dpa/year for Fe) in a volume of 500 cm 3 for testing candidate fusion materials. In 2000, a 3 year Key Element technology Phase (KEP) of IFMIF was started to reduce the key technology risk factors. This interim report summarizes the KEP activities until mid 2001 in the major project work-breakdown areas of accelerator, target, test facilities and design integration. (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

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)

Thermal-hydraulic limitations on water-cooled fusion reactor components

International Nuclear Information System (INIS)

Cha, Y.S.; Misra, B.

1986-01-01

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein are concerned primarily with thermal protection of the reactor components, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue

Development of Radiation Fusion Biotechnology

Energy Technology Data Exchange (ETDEWEB)

Jo, Sung Kee; Lee, Ju Woon; Park, Sang Hyun

2010-04-15

{center_dot} Development of Radiation Fusion Technology with Food Technology by the Application of High Dose Irradiation - To develop fundamental technology using high dose irradiation, effects of high dose irradiation on food components, combined effects of irradiation with food engineering, irradiation condition to destroy radiation resistant foodborne bacteria were studied. - To develop E-beam irradiation technology, irradiation conditions for E-beam and domination effects of E-beam irradiation were determined. The physical marker for E-beam irradiated foods or not was developed. - To develop purposed foods to extreme environmental, ready-to-eat foods and low toxic animal feeds were developed. Through the fundamental researches, the legislation for new irradiated foods and application of E-beam was introduced. {center_dot} Development of modulators against degenerative aging using radiation fusion technology - Confirmation of similarity of radiation-induced aging and normal aging by comparative analysis study - Selection of degenerative aging biomarkers related to immune/hematopoiesis, oxidative damage, molecular signaling, lipid metabolism - Establishment of optimal radiation application conditions for aging modeling - Validation of biomarkers and models using substances {center_dot} Development of biochips and kits using RI detection technology for life science - Establishment of kinase-substrate interaction analysis using RI detection technique (More than 30 times detection sensitivity compared to conventional fluorescence detection techniques). - The RI detection technique reduces the overall experiment time, as the use of blocking agent can be avoided, offer minimum non-specific binding, and facilitates a rapid data analysis with a simplify the process of chip manufacturing

Development of Radiation Fusion Biotechnology

International Nuclear Information System (INIS)

Jo, Sung Kee; Lee, Ju Woon; Park, Sang Hyun

2010-04-01

· Development of Radiation Fusion Technology with Food Technology by the Application of High Dose Irradiation - To develop fundamental technology using high dose irradiation, effects of high dose irradiation on food components, combined effects of irradiation with food engineering, irradiation condition to destroy radiation resistant foodborne bacteria were studied. - To develop E-beam irradiation technology, irradiation conditions for E-beam and domination effects of E-beam irradiation were determined. The physical marker for E-beam irradiated foods or not was developed. - To develop purposed foods to extreme environmental, ready-to-eat foods and low toxic animal feeds were developed. Through the fundamental researches, the legislation for new irradiated foods and application of E-beam was introduced. · Development of modulators against degenerative aging using radiation fusion technology - Confirmation of similarity of radiation-induced aging and normal aging by comparative analysis study - Selection of degenerative aging biomarkers related to immune/hematopoiesis, oxidative damage, molecular signaling, lipid metabolism - Establishment of optimal radiation application conditions for aging modeling - Validation of biomarkers and models using substances · Development of biochips and kits using RI detection technology for life science - Establishment of kinase-substrate interaction analysis using RI detection technique (More than 30 times detection sensitivity compared to conventional fluorescence detection techniques). - The RI detection technique reduces the overall experiment time, as the use of blocking agent can be avoided, offer minimum non-specific binding, and facilitates a rapid data analysis with a simplify the process of chip manufacturing

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

XHM-1 alloy as a promising structural material for water-cooled fusion reactor components

International Nuclear Information System (INIS)

Solonin, M.I.; Alekseev, A.B.; Kazennov, Yu.I.; Khramtsov, V.F.; Kondrat'ev, V.P.; Krasina, T.A.; Rechitsky, V.N.; Stepankov, V.N.; Votinov, S.N.

1996-01-01

Experience gained in utilizing austenitic stainless steel components in water-cooled power reactors indicates that the main cause of their failure is the steel's propensity for corrosion cracking. In search of a material immune to this type of corrosion, different types of austenitic steels and chromium-nickel alloys were investigated and tested at VNIINM. This paper presents the results of studying physical and mechanical properties, irradiation and corrosion resistance in a water coolant at <350 C of the alloy XHM-1 as compared with austenitic stainless steels 00Cr16Ni15Mo3Nb, 00Cr20Ni25Nb and alloy 00Cr20Ni40Mo5Nb. Analysis of the results shows that, as distinct from the stainless steels studied, the XHM-1 alloy is completely immune to corrosion cracking (CC). Not a single induced damage was encountered within 50 to 350 C in water containing different amounts of chlorides and oxygen under tensile stresses up to the yield strength of the material. One more distinctive feature of the alloy compared to steels is that no change in the strength or total elongation is encountered in the alloy specimens irradiated to 32 dpa at 350 C. The XHM-1 alloy has adequate fabricability and high weldability characteristics. As far as its properties are concerned, the XHM-1 alloy is very promising as a material for water-cooled fusion reactor components. (orig.)

Hazard evaluation of The International Fusion Materials Irradiation Facility

Energy Technology Data Exchange (ETDEWEB)

Burgazzi, Luciano [ENEA-Centro Ricerche ' Ezio Clementel' , Advanced Physics Technology Division, Via Martiri di Monte Sole, 4, 40129 Bologna (Italy)]. E-mail: burgazzi@bologna.enea.it

2005-01-15

The International Fusion Materials Irradiation Facility (IFMIF) is aimed to provide an intense neutron source by a high current deuteron linear accelerator and a high-speed lithium flow target, for testing candidate materials for fusion. Liquid lithium is being circulated through a loop and is kept at a temperature above its freezing point. In the frame of the design phase called Key Element technology Phase (KEP), jointly performed by an international team to verify the most important risk factors, safety assessment of the whole plant has been required in order to identify the hazards associated with the plant operation. This paper discusses the safety assessments that were performed and their outcome: Failure Mode and Effect Analysis (FMEA) approach has been adopted in order to accomplish the task. Main conclusions of the study is that, on account of the safety and preventive measures adopted, potential plant related hazards are confined within the IFMIF security boundaries and great care must be exercised to protect workers and site personnel from operating the plant. The analysis has provided as a result a set of Postulated Initiating Events (PIEs), that is off-normal events, that could result in hazardous consequences for the plant, together with the total frequency and the list of component failures which could induce the PIE: this assures the exhaustive list of major initiating events of accident sequences, helpful to the further accident sequence analysis phase. Finally, for each one of the individuated PIEs, the evaluation of the accident evolution, in terms of effects on the plant and relative countermeasures, has allowed to verify that adequate measures are being taken both to prevent the accident occurrence and to cope with the accident consequences, thus assuring the fulfilment of the safety requirements.

Sulfur-containing components of gamma-irradiated garlic bulbs

International Nuclear Information System (INIS)

Joongho Kwon; Jonguck Choi; Hyungsik Yoon

1989-01-01

Sulfur-containing components associated with garlic flavors were investigated to determine the effect of γ-irradiation at 0.1Gy on the quality of garlic bulbs (Allium sativum L.) during storage at 3±1 0 C and 80±5% RH for 10 months. Irradiation treatment had no influence on the amount of total sulfur and thiosulfinate of stored garlic for 10 months, while the storage period brought about a significant reduction (P < 0.05) in the content of both components after the 6-8th month of storage compared with that at the beginning of storage period. The identity of irradiated alliin (S-allyl-L-cysteine sulfoxide) at sprout-inhibition dose was confirmed according to thin-layer chromatography, i.r. and NMR spectroscopy data. (author)

Sulfur-containing components of gamma-irradiated garlic bulbs

Science.gov (United States)

Kwon, Joong-Ho; Choi, Jong-Uck; Yoon, Hyung-Sik

Sulfur-containing components associated with garlic flavors were investigated to determine the effect of γ-irradiation at 0.1 kGy on the quality of garlic bulbs ( Allium sativum L.) during storage at 3±1°C and 80±5% RH for 10 months. Irradiation treatment had no influence on the amount of total sulfur and thiosulfinate of stored garlic for 10 months, while the storage period brought about a significant reduction ( P<0.05) in the content of both components after the 6-8th month of storage compared with that at the beginning of storage period. The identity of irradiated alliin ( S-allyl- L-cysteine sulfoxide) at sprout-inhibition dose was confirmed according to thin-layer chromatography, i.r. and NMR spectroscopy data.

Material Challenges For Plasma Facing Components in Future Fusion Reactors

International Nuclear Information System (INIS)

Linke, J; Pintsuk, G.; Rödig, M.

2013-01-01

. Here a considerable fraction of the plasma energy is deposited on a localized surface area in the divertor strike zone; the time scale of these events is typically in the order of 1 ms. As a consequence, thermal shock induced crack formation, vaporization, surface melting and droplet ejection as well as particle emission induced by brittle destruction processes will limit the lifetime of the components. This is also valid for instabilities in the plasma positioning (vertical displacement events) which cause irreversible damage to plasma facing components, particularly to the metallic wall armour. Moreover, dust particles (neutron activated or toxic metals or tritium enriched carbon) are a serious concern form a safety point of view. In order to investigate the thermally induced plasma wall interaction under fusion specific thermal loads, high heat flux simulation tests are performed routinely in electron or ion beam test facilities as well as in quasi stationary plasma devices. These experiments cover thermal fatigue loads and/or thermal shock tests with relevant operational loading conditions. Furthermore, the wall bombardment with 14 MeV neutrons in D-T-burning plasma devices and the resulting material damage are another critical issue, both, from a safety point of view, but also under the aspect of the component lifetime. While the integrated neutron fluence in ITER will be only in the order of 1 dpa (displacements per atom), future devices such as DEMO or commercial fusion reactors will experience integrated neutron wall loads of 80 to 150 dpa. Therefore the development of new radiation resistant materials and their testing under realistic conditions is required. Due to the lack of an intense 14 MeV neutron source, complex neutron irradiation experiments are performed in material test reactors to quantify the neutron-induced material damage. These tests provide a valuable data base on the degradation of thermal and mechanical parameters. (author)

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

Understanding susceptibility of in-core components to irradiation-assisted stress corrosion cracking

International Nuclear Information System (INIS)

Chung, H.M.; Ruther, W.E.; Sanecki, J.E.; Kassner, T.F.

1991-03-01

As nuclear plants age and accumulated fluences of core structural components increase, susceptibility of the components to irradiation-assisted stress corrosion cracking (IASCC) is also expected to increase. Irradiation-induced sensitization, commonly associated with an IASCC failure, was investigated in this study to provide a better understanding of long-term structural integrity of safety-significant in-core components. Irradiation-induced sensitization of high- and commercial-purity Type 304 stainless steels irradiated in BWRs was analyzed. 7 refs., 8 figs

Effects of irradiation on the components of implantable pacemakers

International Nuclear Information System (INIS)

Kawamura, Shinji; Ono, Seiji; Kuga, Noriyuki; Shiba, Tooru; Hirose, Tetsuo; Matoba, Masaru

2003-01-01

The purpose of this study was to examine the effects of irradiation on implantable pacemaker components. The pacemaker was divided into three components: lead wire and electrode, battery, and electrical circuit, and each component was irradiated by X-ray and electron beams, respectively. The pacemaker parameters were measured by both telemetry data of the programmer and directly measured data from the output terminal. The following results were obtained. For the lead wire and electrode, there was no effect on the pacemaker function due to irradiation by X-ray and electron beams. In the case of battery irradiation, there was no change in battery voltage or current up to 236 Gy X-ray dose. In the electrical circuit, the pacemaker reverted to the regular beating rate (fixed-rate mode) immediately after the start of X-ray irradiation, and it continued in this mode during irradiation. In patients with their own heartbeat rhythm, changing to the fixed-rate mode may cause dangerous conditions such as ventricular fibrillation. When the accumulated irradiation dose is increased, another failure can be seen in the output voltage of the pacemaker. The pacing output voltage dropped rapidly by about 40% at 30-88 Gy. Decreasing the output voltage results in pacing disorders, and heart failure may occur. In the telemetry data of the programmer, no change in output voltage could be detected, highlighting the difference between telemetry data and actual pacing data. (author)

IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-03-01

The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m{sup 2}, 20 dpa/y in Fe, in a volume of 500 cm{sup 3} and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

Computer simulation of defect behavior under fusion irradiation environments

International Nuclear Information System (INIS)

Muroga, T.; Ishino, S.

1983-01-01

To simulate defect behavior under irradiation, three kinds of cascade-annealing calculations have been carried out in alpha-iron using the codes MARLOWE, DAIQUIRI and their modifications. They are (1) cascade-annealing calculation with different masses of projectile, (2) defect drifting near dislocations after cascade production and (3) cascade-overlap calculation. The defect survival ratio is found to increase as decreasing mass of the projectile both after athermal close-pair recombination and after thermal annealing. It is shown that at moderate temperatures vacancy clustering is enhanced near dislocations. Cascade-overlap is found to decrease the defect survivability. In addition, the role of helium in vacancy clustering has been calculated in aluminium lattices and its effect is found to depend strongly on temperature, interstitials and the mobility of small clusters. These results correspond well to the experimental data and will be helpful for correlating between fusion and simulation irradiations. (orig.)

IFMIF (International Fusion Materials Irradiation Facility) key element technology phase interim report

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-03-01

Activities of International Fusion Materials Irradiation Facility (IFMIF) have been performed under an IEA collaboration since 1995. IFMIF is an accelerator-based deuteron (D{sup +})-lithium (Li) neutron source designed to produce an intense neutron field (2 MW/m{sup 2}, 20 dpa/year for Fe) in a volume of 500 cm{sup 3} for testing candidate fusion materials. In 2000, a 3 year Key Element technology Phase (KEP) of IFMIF was started to reduce the key technology risk factors. This interim report summarizes the KEP activities until mid 2001 in the major project work-breakdown areas of accelerator, target, test facilities and design integration. (author)

Sulfur-containing components of gamma-irradiated garlic bulbs

Energy Technology Data Exchange (ETDEWEB)

Joongho Kwon (Korea Advanced Energy Research Inst., Daeduk (Republic of Korea)); Jonguck Choi; Hyungsik Yoon (Kyungpook National Univ., Taegu (Republic of Korea))

1989-01-01

Sulfur-containing components associated with garlic flavors were investigated to determine the effect of {gamma}-irradiation at 0.1Gy on the quality of garlic bulbs (Allium sativum L.) during storage at 3{plus minus}1{sup 0}C and 80{plus minus}5% RH for 10 months. Irradiation treatment had no influence on the amount of total sulfur and thiosulfinate of stored garlic for 10 months, while the storage period brought about a significant reduction (P < 0.05) in the content of both components after the 6-8th month of storage compared with that at the beginning of storage period. The identity of irradiated alliin (S-allyl-L-cysteine sulfoxide) at sprout-inhibition dose was confirmed according to thin-layer chromatography, i.r. and NMR spectroscopy data. (author).

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...

Conceptual design of a fission-based integrated test facility for fusion reactor components

International Nuclear Information System (INIS)

Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

1982-01-01

The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

Reliability, availability, and quality assurance considerations for fusion components

International Nuclear Information System (INIS)

Buende, R.

1995-01-01

The complexity of magnetic confinement machines has been a matter of concern in developing fusion power plants as electricity generating stations because it might reduce plant availability. A comprehensive reliability and availability (R and A) programme to determine the availability of a next step fusion machine was performed during definition and conceptual design of the Next European Torus. In addition to giving an overview of the expected contributions to unavailability of the various components, this activity identified the basic approach to be taken to specify and to achieve necessary improvements. This paper, after giving some basic definitions, describes the essentials of the R and A programme, its results, and the guidelines derived for further work towards a sufficiently reliable fusion plant. These guidelines refer to improvement of the reliability database and the quality assurance to be performed at the design stage of a next step machine. (orig.)

Fiscal year 1976T (add-on quarter) DT fusion neutron irradiations and dosimetry at the LLL rotating target neutron source

International Nuclear Information System (INIS)

MacLean, S.C.

1977-01-01

The DT fusion neutron irradiation of more than 90 samples during seven irradiation periods (beam-on time of more than 430.9 hours) is described. Experiments from 15 individuals representing six institutions are summarized. The numbers of UCID dosimetry reports detailing each of the irradiations is given

Effectiveness of thermoluminescence analysis to detect low quantity of gamma-irradiated component in non-irradiated mushroom powders

International Nuclear Information System (INIS)

Akram, Kashif; Ahn, Jae-Jun; Shahbaz, Hafiz Muhammad; Jo, Deokjo; Kwon, Joong-Ho

2013-01-01

Gamma-irradiated (0–10 kGy) dried mushrooms (Lentinus edodes) powders were mixed at different ratios (1–10%) in the non-irradiated samples and investigated using photostimulated-luminescence (PSL), electron spin resonance (ESR) and thermoluminescence (TL) techniques. The PSL results were negative for all samples at 1% mixing ratio, whereas intermediate results were observed for the samples containing 5% or 10% irradiated component with the exception (positive) of 10% mixing of 10 kGy-irradiated sample. The ESR analysis showed the presence of crystalline sugar radicals in the irradiated samples but the radiation-specific spectral features were absent in the mixed samples. TL analysis showed the radiation-specific TL glow curves; however, the complicated results were observed at 1% mixing of 2 and 5 kGy-irradiated samples, which required careful evaluations to draw the final conclusion about the irradiation status of the samples. TL ratios could only confirm the results of samples with 5% and 10% mixing of 10 kGy, and 10% mixing of 5 kGy-irradiated components. SEM-EDX analysis showed that feldspar and quartz were major contaminating minerals, responsible for the radiation-specific luminescence characteristics. -- Highlights: ► Detection of irradiated food is important to enforce the applied regulations. ► The effectiveness of TL analysis was investigated to detect irradiated component. ► The TL results were compared with those from PSL and ESR analysis. ► TL analysis was most effective to characterize the irradiation status of samples. ► SEM-EDX analysis showed feldspar and quartz as the main source of TL properties

Effectiveness of thermoluminescence analysis to detect low quantity of gamma-irradiated component in non-irradiated mushroom powders

Energy Technology Data Exchange (ETDEWEB)

Akram, Kashif [School of Food Science and Biotechnology, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Institute of Food Science and Nutrition, University of Sargodha, Sargodha 40100 (Pakistan); Ahn, Jae-Jun [School of Food Science and Biotechnology, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Shahbaz, Hafiz Muhammad [School of Food Science and Biotechnology, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Institute of Food Science and Nutrition, University of Sargodha, Sargodha 40100 (Pakistan); Jo, Deokjo [School of Food Science and Biotechnology, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kwon, Joong-Ho, E-mail: jhkwon@knu.ac.kr [School of Food Science and Biotechnology, Kyungpook National University, Daegu 702-701 (Korea, Republic of)

2013-04-15

Gamma-irradiated (0–10 kGy) dried mushrooms (Lentinus edodes) powders were mixed at different ratios (1–10%) in the non-irradiated samples and investigated using photostimulated-luminescence (PSL), electron spin resonance (ESR) and thermoluminescence (TL) techniques. The PSL results were negative for all samples at 1% mixing ratio, whereas intermediate results were observed for the samples containing 5% or 10% irradiated component with the exception (positive) of 10% mixing of 10 kGy-irradiated sample. The ESR analysis showed the presence of crystalline sugar radicals in the irradiated samples but the radiation-specific spectral features were absent in the mixed samples. TL analysis showed the radiation-specific TL glow curves; however, the complicated results were observed at 1% mixing of 2 and 5 kGy-irradiated samples, which required careful evaluations to draw the final conclusion about the irradiation status of the samples. TL ratios could only confirm the results of samples with 5% and 10% mixing of 10 kGy, and 10% mixing of 5 kGy-irradiated components. SEM-EDX analysis showed that feldspar and quartz were major contaminating minerals, responsible for the radiation-specific luminescence characteristics. -- Highlights: ► Detection of irradiated food is important to enforce the applied regulations. ► The effectiveness of TL analysis was investigated to detect irradiated component. ► The TL results were compared with those from PSL and ESR analysis. ► TL analysis was most effective to characterize the irradiation status of samples. ► SEM-EDX analysis showed feldspar and quartz as the main source of TL properties.

Materials for heat flux components of the first wall in fusion reactors

International Nuclear Information System (INIS)

Hoven, H.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.

1985-08-01

Materials of the First Wall in near-fusion plasma machines are subjected to a complex load system resulting from the plasma-wall interaction. The materials for their part also influence the plasma. Suitable materials must be available in order to ensure that the wall components achieve a sufficiently long dwell time and that their effects on the plasma remain small and controllable. The present report discusses relations between the plasma-wall interaction, the reactions of the materials and testing and examination methods for specific problems in developing and selecting suitable materials for highly stressed components on the First Wall of fusion reactors. (orig.)

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)

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

DT fusion neutron irradiation of ORNL magnesium oxide crystals and BNL--LASL superconductor wires

International Nuclear Information System (INIS)

MacLean, S.C.

1978-01-01

The DT fusion neutron irradiation of two ORNL magnesium oxide crystals and eleven BNL-LASL superconductor wires is described. The sample position and neutron dose record are given. The maximum neutron fluence on any sample was 2.16 x 10 16 neutrons/cm 2

Design of a high-flux test assembly for the Fusion Materials Irradiation Test Facility

International Nuclear Information System (INIS)

Opperman, E.K.; Vogel, M.A.

1982-01-01

The Fusion Material Test Facility (FMIT) will provide a high flux fusion-like neutron environment in which a variety of structural and non-structural materials irradiations can be conducted. The FMIT experiments, called test assemblies, that are subjected to the highest neutron flux magnitudes and associated heating rates will require forced convection liquid metal cooling systems to remove the neutron deposited power and maintain test specimens at uniform temperatures. A brief description of the FMIT facility and experimental areas is given with emphasis on the design, capabilities and handling of the high flux test assembly

Superconducting magnet radiation limit considerations for fusion reactors

International Nuclear Information System (INIS)

Sawan, M.E.; Walstrom, P.L.

1986-01-01

The radiation limits for fusion reactor magnets have a direct impact on the cost of electricity. For example, reducing the inboard shield by 1 cm saves up to $3 million in the Tokamak Fusion Core Experiment cost. The magnet components most sensitive to radiation damage are the superconductor, stabilizer, and insulators. Nuclear heating in the magnet affects the design and also impacts the economic performance of the reactor through increased refrigeration costs. The radiation effects in the different components of the magnet are related, as all of them are determined by the flux level in the magnet. Hence, in efforts to push radiation limits, these effects should be considered simultaneously. Furthermore, the levels of radiation effects that correspond to the optimum nuclear heating determined from economic trade-off analysis will be useful in specifying the fluence, dose, and stabilization limit goals for the magnet development program. In this paper, we review the available irradiation data and assess the need for achieving higher irradiation levels

Reduced cost design of liquid lithium target for international fusion material irradiation facility (IFMIF)

International Nuclear Information System (INIS)

Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki

2001-01-01

The International Fusion Materials Irradiation Facility (IFMIF) is being jointly planned to provide an accelerator-based D-Li neutron source to produce intense high energy neutrons (2 MW/m 2 ) up to 200 dpa and a sufficient irradiation volume (500 cm 3 ) for testing the candidate materials and components up to about a full lifetime of their anticipated use in ITER and DEMO. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid lithium flow with a speed of 20 m/s. Following Conceptual Design Activity (1995-1998), a design study with focus on cost reduction without changing its original mission has been done in 1999. The following major changes to the CAD target design have been considered in the study and included in the new design: i) number of the Li target has been changed from 2 to 1, ii) spare of impurity traps of the Li loop was removed although the spare will be stored in a laboratory for quick exchange, iii) building volume was reduced via design changes in lithium loop length. This paper describes the reduced cost design of the lithium target system and recent status of Key Element Technology activities. (author)

Accelerators for Fusion Materials Testing

Science.gov (United States)

Knaster, Juan; Okumura, Yoshikazu

Fusion materials research is a worldwide endeavor as old as the parallel one working toward the long term stable confinement of ignited plasma. In a fusion reactor, the preservation of the required minimum thermomechanical properties of the in-vessel components exposed to the severe irradiation and heat flux conditions is an indispensable factor for safe operation; it is also an essential goal for the economic viability of fusion. Energy from fusion power will be extracted from the 14 MeV neutron freed as a product of the deuterium-tritium fusion reactions; thus, this kinetic energy must be absorbed and efficiently evacuated and electricity eventually generated by the conventional methods of a thermal power plant. Worldwide technological efforts to understand the degradation of materials exposed to 14 MeV neutron fluxes >1018 m-2s-1, as expected in future fusion power plants, have been intense over the last four decades. Existing neutron sources can reach suitable dpa (“displacement-per-atom”, the figure of merit to assess materials degradation from being exposed to neutron irradiation), but the differences in the neutron spectrum of fission reactors and spallation sources do not allow one to unravel the physics and to anticipate the degradation of materials exposed to fusion neutrons. Fusion irradiation conditions can be achieved through Li (d, xn) nuclear reactions with suitable deuteron beam current and energy, and an adequate flowing lithium screen. This idea triggered in the late 1970s at Los Alamos National Laboratory (LANL) a campaign working toward the feasibility of continuous wave (CW) high current linacs framed by the Fusion Materials Irradiation Test (FMIT) project. These efforts continued with the Low Energy Demonstrating Accelerator (LEDA) (a validating prototype of the canceled Accelerator Production of Tritium (APT) project), which was proposed in 2002 to the fusion community as a 6.7MeV, 100mA CW beam injector for a Li (d, xn) source to bridge

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

Radiation-induced organogenesis: effects of irradiated medium and its components on tobacco tissue culture

International Nuclear Information System (INIS)

Degani, N.

1975-01-01

Gamma irradiated medium induces the formation of buds in non-irradiated dark growth tobacco callus (Nicotiana tabacum Var. Wisconsin No.38). Experiments were conducted to determine the component(s) of the medium that is effective in this radiation-induced organogenesis. Fraction of medium were irradiated singly and in combination, then combined with non-irradiated fractions to form the complete growth medium. The results showed that irradiated indoleacetic acid (IAA) was not the effective component in the induction of organogensis. Omission of IAA from the medium resulted in the formation of buds, as expected. Irradiated myo-inositol induced organogenesis more consistently than the other irradiated components. The age of the inoculum tissue and its passage number from the tobacco stem affected the potency of the tobacco callus to organise. (author)

Fusion reactor materials

International Nuclear Information System (INIS)

Sethi, V.K.; Scholz, R.; Nolfi, F.V. Jr.; Turner, A.P.L.

1980-01-01

Data are given for each of the following areas: (1) effects of irradiation on fusion reactor materials, (2) hydrogen permeation and materials behavior in alloys, (3) carbon coatings for fusion applications, (4) surface damage of TiB 2 coatings under energetic D + and 4 He + irradiations, and (5) neutron dosimetry

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Detection of irradiated components in flavour blends composed of non-irradiated species herbs and vegetable seasonings by thermoluminescence method

International Nuclear Information System (INIS)

Malec-Czechowska, K.; Stachowicz, W.

2003-01-01

The results of experiments on the detection of irradiated component in commercial flavour blends composed of a mixture of non-irradiated spices, herbs and seasonings are presented. A method based on the thermoluminescence measurements on silicate materials isolated from blends has been adapted. It has been proved that by applying of this technique it is possible to detect 0.95% by weight of paprika, irradiated with a dose of 7 kGy, which was a minor component of non-irradiated flavour blends. (author)

Infrared and visible image fusion based on robust principal component analysis and compressed sensing

Science.gov (United States)

Li, Jun; Song, Minghui; Peng, Yuanxi

2018-03-01

Current infrared and visible image fusion methods do not achieve adequate information extraction, i.e., they cannot extract the target information from infrared images while retaining the background information from visible images. Moreover, most of them have high complexity and are time-consuming. This paper proposes an efficient image fusion framework for infrared and visible images on the basis of robust principal component analysis (RPCA) and compressed sensing (CS). The novel framework consists of three phases. First, RPCA decomposition is applied to the infrared and visible images to obtain their sparse and low-rank components, which represent the salient features and background information of the images, respectively. Second, the sparse and low-rank coefficients are fused by different strategies. On the one hand, the measurements of the sparse coefficients are obtained by the random Gaussian matrix, and they are then fused by the standard deviation (SD) based fusion rule. Next, the fused sparse component is obtained by reconstructing the result of the fused measurement using the fast continuous linearized augmented Lagrangian algorithm (FCLALM). On the other hand, the low-rank coefficients are fused using the max-absolute rule. Subsequently, the fused image is superposed by the fused sparse and low-rank components. For comparison, several popular fusion algorithms are tested experimentally. By comparing the fused results subjectively and objectively, we find that the proposed framework can extract the infrared targets while retaining the background information in the visible images. Thus, it exhibits state-of-the-art performance in terms of both fusion effects and timeliness.

Measurements of integrated components' parameters versus irradiation doses gamma radiation (60Co) dosimetry-methodology-tests

International Nuclear Information System (INIS)

Fuan, J.

1991-01-01

This paper describes the methodology used for the irradiation of the integrated components and the measurements of their parameters, using Quality Insurance of dosimetry: - Measurement of the integrated dose using the competences of the Laboratoire Central des Industries Electriques (LCIE): - Measurement of irradiation dose versus source/component distance, using a calibrated equipment. - Use of ALANINE dosimeters, placed on the support of the irradiated components. - Assembly and polarization of components during the irradiations. Selection of the irradiator. - Measurement of the irradiated components's parameters, using the competences of the societies: - GenRad: GR130 tests equipement placed in the DEIN/SIR-CEN SACLAY. - Laboratoire Central des Industries Electriques (LCIE): GR125 tests equipment and this associated programmes test [fr

Development of lithium target system in engineering validation and engineering design activity of the international fusion materials irradiation facility (IFMIF/EVEDA)

International Nuclear Information System (INIS)

Wakai, Eiichi; Kondo, Hiroo; Sugimoto, Masayoshi; Ida, Mizuho; Kanemura, Takuji; Watanabe, Kazuyoshi; Fujishiro, Kouji; Edao, Yuuki; Niitsuma, Shigeto; Kimura, Haruyuki; Fukada, Satoshi; Hiromoto, Tetsushi; Shigeharu, Satoshi; Yagi, Jyuro; Furukawa, Tomohiro; Hirakawa, Yasushi; Suzuki, Akihiro; Terai, Takayuki; Horiike, Hiroshi; Hoashi, Eiji; Suzuki, Sachiko; Yamaoka, Nobuo; Serizawa, Hisashi; Kawahito, Yosuke; Tsuji, Yoshiyuki; Furuya, Kazuyuki; Takeo, Fumio

2012-01-01

Engineering validation and engineering design activity (EVEDA) for the international fusion materials irradiation facility (IFMIF) has been conducted since 2007. Research and development of the Lithium target facility is an important part of this activity. We constructed a world largest liquid Lithium test loop with a capacity of 5000 L in 2010 and successfully completed the first stage validation tests (functional tests of components and Lithium flow test (flow velocity 15 m/s at the target). In the present article, recent results of the EVEDA activity for the Lithium target facility and related technologies on liquid Lithium are reviewed. (author)

Structural materials for fusion reactors

International Nuclear Information System (INIS)

Victoria, M.; Baluc, N.; Spaetig, P.

2001-01-01

In order to preserve the condition of an environmentally safe machine, present selection of materials for structural components of a fusion reactor is made not only on the basis of adequate mechanical properties, behavior under irradiation and compatibility with other materials and cooling media, but also on their radiological properties, i.e. activity, decay heat, radiotoxicity. These conditions strongly limit the number of materials available to a few families of alloys, generically known as low activation materials. We discuss the criteria for deciding on such materials, the alloys resulting from the application of the concept and the main issues and problems of their use in a fusion environment. (author)

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)

High-power corrugates waveguide components for mm-wave fusion heating systems

International Nuclear Information System (INIS)

Olstad, R.A.; Doane, J.L.; Moeller, C.P.; O'Neill, R.C.; Di Martino, M.

1996-10-01

Considerable progress has been made over the last year in the U.S., Japan, Russia, and Europe in developing high power long pulse gyrotrons for fusion plasma heating and current drive. These advanced gyrotrons typically operate at a frequency in the range 82 GHz to 170 GHz at nearly megawatt power levels for pulse lengths up to 5 s. To take advantage of these new microwave sources for fusion research, new and improved transmission line components are needed to reliably transmit microwave power to plasmas with minimal losses. Over the last year, General Atomics and collaborating companies (Spinner GmbH in Europe and Toshiba Corporation in Japan) have developed a wide variety of new components which meet the demanding power, pulse length, frequency, and vacuum requirements for effective utilization of the new generation of gyrotrons. These components include low-loss straight corrugated waveguides, miter bends, miter bend polarizers, power monitors, waveguide bellows, de breaks, waveguide switches, dummy loads, and distributed windows. These components have been developed with several different waveguide diameters (32, 64, and 89 mm) and frequency ranges (82 GHz to 170 GHz). This paper describes the design requirements of selected components and their calculated and measured performance characteristics

Parameter studies for a two-component fusion experiment

International Nuclear Information System (INIS)

Towner, H.H.

1975-01-01

The sensitivity of the energy multiplication of a two-component fusion experiment is examined relative to the following parameters: energy confinement time (tau/sub E/), particle confinement time (tau/sub p/), effective Z of the plasma (Z/sub eff/), injection rate (j/sub I/) and injection energy (E/sub I/). The Energy Research and Development Administration recently approved funding for such a fusion device (the Toroidal Fusion Test Reactor or TFTR) which will be built at the Princeton Plasma Physics Laboratory. Hence, such a parameter study seems both timely and necessary. This work also serves as an independent check on the design values proposed for the TFTR to enable it to achieve energy breakeven (F = 1). Using the nominal TFTR design parameters and a self-consistent ion-electron power balance, the maximum F-value is found to be approximately 1.2 which occurs at an injection energy of approximately 210 KeV. The injector operation, i.e. its current and energy capability are shown to be a very critical factor in the TFTR performance. However, if the injectors meet the design objectives, there appears to be sufficient latitude in the other parameters to offer reasonable assurance that energy breakeven can be achieved. (U.S.)

Structural materials requirements for in-vessel components of fusion power plants

International Nuclear Information System (INIS)

Schaaf, B. van der

2000-01-01

The economic production of fusion energy is determined by principal choices such as using magnetic plasma confinement or generating inertial fusion energy. The first generation power plants will use deuterium and tritium mixtures as fuel, producing large amounts of highly energetic neutrons resulting in radiation damage in materials. In the far future the advanced fuels, 3 He or 11 B, determine power plant designs with less radiation damage than in the first generation. The first generation power plants design must anticipate radiation damage. Solid sacrificing armour or liquid layers could limit component replacements costs to economic levels. There is more than radiation damage resistance to determine the successful application of structural materials. High endurance against cyclic loading is a prominent requirement, both for magnetic and inertial fusion energy power plants. For high efficiency and compactness of the plant, elevated temperature behaviour should be attractive. Safety and environmental requirements demand that materials have low activation potential and little toxic effects under both normal and accident conditions. The long-term contenders for fusion power plant components near the plasma are materials in the range from innovative steels, such as reduced activation ferritic martensitic steels, to highly advanced ceramic composites based on silicon carbide, and chromium alloys. The steels follow an evolutionary path to basic plant efficiencies. The competition on the energy market in the middle of the next century might necessitate the riskier but more rewarding development of SiCSiC composites or chromium alloys

Reducing risk and accelerating delivery of a neutron source for fusion materials research

Energy Technology Data Exchange (ETDEWEB)

Surrey, E., E-mail: Elizabeth.Surrey@ccfe.ac.uk [EURATOM/CCFE, Abingdon OX14 3DB (United Kingdom); Porton, M. [EURATOM/CCFE, Abingdon OX14 3DB (United Kingdom); Davenne, T.; Findlay, D.; Letchford, A.; Thomason, J. [STFC Rutherford Appleton Laboratory, Harwell OX11 0QX (United Kingdom); Roberts, S.G.; Marrow, J.; Seryi, A. [University of Oxford, Oxford OX1 3DP (United Kingdom); Connolly, B. [University of Birmingham, Edgbaston B15 2TT (United Kingdom); Owen, H. [University of Manchester, Manchester M13 9PL (United Kingdom)

2014-04-15

Highlights: • Proposed neutron source for fusion materials – FAFNIR – n(d,C) stripping source. • Near term technology, reduces risk compared with IFMIF, timely data production. • Technical, economic and programme needs assessed, compatible with EU Roadmap proposals. • Safety case impacts regulatory role for source, now mainly stakeholder insurance. - Abstract: The materials engineering database relevant to fusion irradiation is poorly populated and it has long been recognized that a fusion spectrum neutron source will be required, the facility IFMIF being the present proposal. Re-evaluation of the regulatory approach for the EU proposed DEMO device shows that the purpose of the source can be changed from lifetime equivalent irradiation exposure to data generation at lower levels of exposure by adopting a defence in depth strategy and regular component surveillance. This reduces the specification of the source with respect to IFMIF allowing lower risk technology solutions to be considered. A description of such a source, the Facility for Fusion Neutron Irradiation Research, FAFNIR, is presented here along with project timescales and costs.

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

Effects of low-temperature fusion neutron irradiation on critical properties of a monofilament niobium-tin superconductor

International Nuclear Information System (INIS)

Guinan, M.W.; Van Konynenburg, R.A.; Mitchell, J.B.

1984-01-01

The objective of this work was to irradiate a Nb 3 Sn superconductor with 14.8 MeV neutrons at 4 K and measure critical current in transverse fields of up to 12 T, irradiating up to a fluence sufficient to decrease the critical current to below its initial value. Critical temperatures were also to be measured. The samples were to be kept near 4 K between the irradiation and the measurement of critical properties. This work is directed toward establishing an engineering design fluence limit for Nb 3 Sn when used in fusion reactor superconducting magnets

Effects of low-temperature fusion neutron irradiation on critical properties of a monofilament niobium-tin superconductor

Energy Technology Data Exchange (ETDEWEB)

Guinan, M.W.; Van Konynenburg, R.A.; Mitchell, J.B.

1984-03-22

The objective of this work was to irradiate a Nb/sub 3/Sn superconductor with 14.8 MeV neutrons at 4 K and measure critical current in transverse fields of up to 12 T, irradiating up to a fluence sufficient to decrease the critical current to below its initial value. Critical temperatures were also to be measured. The samples were to be kept near 4 K between the irradiation and the measurement of critical properties. This work is directed toward establishing an engineering design fluence limit for Nb/sub 3/Sn when used in fusion reactor superconducting magnets.

Development of large components for the fusion reactor vacuum circuits

International Nuclear Information System (INIS)

Perinic, D.; Lorrain, C.

1986-06-01

The Commission of the European Communities appointed in mid-1983 the Centre d'Etudes Nucleaires de Saclay and the Kernforschungszentrum Karlsruhe GmbH to investigate whether large vacuum components for use in the fusion machine can be built. The following individual targets have been defined for studies under this project: - Elaboration of technical specifications for large components. - Investigation of the feasibility. - Specification of the tests required and planning of a testing facility. The plasma chamber pumping system is essentially concerned

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.

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

RCC-MRx: Design and construction rules for mechanical components in high-temperature structures, experimental reactors and fusion reactors

International Nuclear Information System (INIS)

2015-01-01

The RCC-MRx code was developed for sodium-cooled fast reactors (SFR), research reactors (RR) and fusion reactors (FR-ITER). It provides the rules for designing and building mechanical components involved in areas subject to significant creep and/or significant irradiation. In particular, it incorporates an extensive range of materials (aluminum and zirconium alloys in response to the need for transparency to neutrons), sizing rules for thin shells and box structures, and new modern welding processes: electron beam, laser beam, diffusion and brazing. The RCC-MR code was used to design and build the prototype Fast Breeder Reactor (PFBR) developed by IGCAR in India and the ITER Vacuum Vessel. The RCC-Mx code is being used in the current construction of the RJH experimental reactor (Jules Horowitz reactor). The RCC-MRx code is serving as a reference for the design of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), for the design of the primary circuit in MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) and the design of the target station of the ESS project (European Spallation Source). Contents of the 2015 edition of the RCC-MRx code: Section I General provisions; Section II Additional requirements and special provisions; Section III Rules for nuclear installation mechanical components: Volume I: Design and construction rules: Volume A (RA): General provisions and entrance keys, Volume B (RB): Class 1 components and supports, Volume C (RC): Class 2 components and supports, Volume D (RD): Class 3 components and supports, Volume K (RK): Examination, handling or drive mechanisms, Volume L (RL): Irradiation devices, Volume Z (Ai): Technical appendices; Volume II: Materials; Volume III: Examinations methods; Volume IV: Welding; Volume V: Manufacturing operations; Volume VI: Probationary phase rules

Irradiation effects on plasma diagnostic components (2)

International Nuclear Information System (INIS)

Nishitani, Takeo; Sugie, Tatsuo

2002-03-01

Irradiation tests on a number of diagnostic components under fission neutrons, gamma-rays and 14 MeV neutrons have been carried out as a part of the ITER technology R and D program. UV range transmission losses of a KU-1 quartz were measured during 14 MeV neutron and 60 Co gamma-ray irradiation. Significant transmission losses were observed in the wavelength of 200-300 nm. Five kinds of ITER round robin fibers were irradiated in JMTR and the 60 Co gamma-ray irradiation facility. KS-4V, KU-H2G and F-doped fibers have a rather good radiation hardness, which might be available just outside of the vacuum vessel in ITER. Mica substrate bolometer was irradiated in JMTR up to 0.1 dpa. During the cool down phase of the first cycle all connections went open circuit. The use of gold meanders in the bolometer might be problematic in ITER. The magnetic probes were irradiated in JMTR. Drift of 10 - 40 mVs for 1000s was observed with a digital longterm integrator, however, which might be induced not only by RIEMF but also drift inside the integrator itself. ITER-relevant magnetic coil could be made with MI-cables, whose electric drift for 1000-s integration is less than 0.5 mVs. (author)

The role of radiation damage analysis in the fusion program

International Nuclear Information System (INIS)

Doran, D.G.

1983-01-01

The objective of radiation damage analysis is the prediction of the performance of facility components exposed to a radiation environment. The US Magnetic Fusion Energy materials program includes an explicit damage analysis activity within the Damage Analysis and Fundamental Studies (DAFS) Program. Many of the papers in these Proceedings report work done directly or indirectly in support of the DAFS program. The emphasis of this program is on developing procedures, based on an understanding of damage mechanisms, for applying data obtained in diverse radiation environments to the prediction of component behavior in fusion devices. It is assumed that the Fusion Materials Irradiation Test Facility will be available in the late 1980s to test (and calibrate where necessary) correlation procedures to the high fluences expected in commercial reactors. (orig.)

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

Materials research and development for fusion energy applications

International Nuclear Information System (INIS)

Zinkle, S.J.; Snead, L.L.

1998-01-01

Some of the critical issues associated with materials selection for proposed magnetic fusion reactors are reviewed, with a brief overview of refractory alloys (vanadium, tantalum, molybdenum, tungsten) and primary emphasis on ceramic materials. SiC/SiC composites are under consideration for the first wall and blanket structure, and dielectric insulators will be used for the heating, control and diagnostic measurement of the fusion plasma. Key issues for SiC/SiC composites include radiation-induced degradation in the strength and thermal conductivity. Recent work has focused on the development of radiation-resistant fibers and fiber/matrix interfaces (porous SiC, SiC multilayers) which would also produce improved SiC/SiC performance for applications such as heat engines and aerospace components. The key physical parameters for dielectrics include electrical conductivity, dielectric loss tangent and thermal conductivity. Ionizing radiation can increase the electrical conductivity of insulators by many orders of magnitude, and surface leakage currents can compromise the performance of some fusion energy components. Irradiation can cause a pronounced degradation in the loss tangent and thermal conductivity. Fundamental physical parameter measurements on ceramics which are of interest for both fusion and non-fusion applications are discussed

DT fusion neutron irradiation of BPNL niobium nickel and 316 stainless steel at 1750C

International Nuclear Information System (INIS)

MacLean, S.C.

1977-01-01

The DT fusion neutron irradiation at 175 0 C of 17 niobium wires, one niobium foil, 14 316 stainless steel wires, one 316 stainless steel foil, nine nickel wires, and two nickel foils from BPNL is described. The sample position, beam-on time, neutron dose record, and neutron fluence are given

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

CERN Document Server

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

2014-01-01

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

Stored energy in irradiated silicon carbide

Energy Technology Data Exchange (ETDEWEB)

Snead, L.L.; Burchell, T.D. [Oak Ridge National Lab., TN (United States)

1997-04-01

This report presents a short review of the phenomenon of Wigner stored energy release from irradiated graphite and discusses it in relation to neutron irradiation of silicon carbide. A single published work in the area of stored energy release in SiC is reviewed and the results are discussed. It appears from this previous work that because the combination of the comparatively high specific heat of SiC and distribution in activation energies for recombining defects, the stored energy release of SiC should only be a problem at temperatures lower than those considered for fusion devices. The conclusion of this preliminary review is that the stored energy release in SiC will not be sufficient to cause catastrophic heating in fusion reactor components, though further study would be desirable.

Accelerator conceptual design of the international fusion materials irradiation facility

International Nuclear Information System (INIS)

Sugimoto, M.; Kinsho, M.; Teplyakov, V.; Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J.; Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K.; Miyahara, A.; Olivier, M.; Piechowiak, E.; Tanabe, Y.

1998-01-01

The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.)

Analysis of displacement damage in materials in nuclear fusion facilities (DEMO, IFMIF and TechnoFusion)

International Nuclear Information System (INIS)

Mota, F.; Vila, R.; Ortiz, C.; Garcia, A.; Casal, N.; Ibarra, A.; Rapisarda, D.; Queral, V.

2011-01-01

Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al 2 O 3 , SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.

Analysis of displacement damage in materials in nuclear fusion facilities (DEMO, IFMIF and TechnoFusion)

Energy Technology Data Exchange (ETDEWEB)

Mota, F., E-mail: fernando.mota@ciemat.es [Laboratorio Nacional de Fusion por Confinamiento Magnetico-CIEMAT, 28040 Madrid (Spain); Vila, R.; Ortiz, C.; Garcia, A.; Casal, N.; Ibarra, A.; Rapisarda, D.; Queral, V. [Laboratorio Nacional de Fusion por Confinamiento Magnetico-CIEMAT, 28040 Madrid (Spain)

2011-10-15

Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al{sub 2}O{sub 3}, SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.

Use of virtual reality for optimizing the life cycle of a fusion component

Energy Technology Data Exchange (ETDEWEB)

Keller, D., E-mail: delphine.keller@cea.fr [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); Doceul, L.; Ferlay, F.; Louison, C.; Pilia, A.; Pavy, K. [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); Chodorge, L.; Andriot, C. [CEA Saclay – DIGITEO Moulon, DRT/LIST/DIASI/LSI, F-91191 Gif Sur Yvette (France)

2015-12-15

Efficient development of a complex system such as a fusion component needs a stringent integration of standard and new constraints. For example, compared to the previous fusion experimental devices, remote handling (RH) and safety requirements are in ITER key parameters which must be integrated since the earliest design. For optimizing such integration studies, CEA, IRFM decided in 2010 to implement the use of virtual reality (VR) tools during the life cycle (from design to operation) of a fusion component. This paper describes a first feedback of such use for fusion engineering purposes. After a short overview of the CEA, IRFM VR platform capabilities, three main uses will be described: design review, simulation of remote handling and hands-on operations, with man in the loop. The Design review mode was intensively used within the framework of a fruitful collaboration with ITER design Integration Team. This mode, fully compatible with CAD software, enables scale one data visualization with stereoscopic rendering. It improves the efficiency in detecting inconsistencies inside models and machine sub-system design optimization needs. Several accessibility cases of major Safety Important Components (SIC-1) were studied giving important requirements to the design at an early stage. CEA, IRFM, in close collaboration with expertise of CEA, LIST for VR simulation software, applies VR technologies for designing RH maintenance scenario for ITER Test Blanket System (TBS) and Ion cyclotron Resonance Heating (ICRH) Port Plugs. RH compatibility studies using VR pointed out major design drivers while helping to propose credible solution. VR platform is intensively used in the design of WEST (Tungsten (W) Environment Steady-state Tokamak) components and assembly studies, providing important information about the feasibility of assembly processes, optimization of physical mock-ups and ergonomic posture and gestures of operator. Finally, new perspectives, as the integration of

Use of virtual reality for optimizing the life cycle of a fusion component

International Nuclear Information System (INIS)

Keller, D.; Doceul, L.; Ferlay, F.; Louison, C.; Pilia, A.; Pavy, K.; Chodorge, L.; Andriot, C.

2015-01-01

Efficient development of a complex system such as a fusion component needs a stringent integration of standard and new constraints. For example, compared to the previous fusion experimental devices, remote handling (RH) and safety requirements are in ITER key parameters which must be integrated since the earliest design. For optimizing such integration studies, CEA, IRFM decided in 2010 to implement the use of virtual reality (VR) tools during the life cycle (from design to operation) of a fusion component. This paper describes a first feedback of such use for fusion engineering purposes. After a short overview of the CEA, IRFM VR platform capabilities, three main uses will be described: design review, simulation of remote handling and hands-on operations, with man in the loop. The Design review mode was intensively used within the framework of a fruitful collaboration with ITER design Integration Team. This mode, fully compatible with CAD software, enables scale one data visualization with stereoscopic rendering. It improves the efficiency in detecting inconsistencies inside models and machine sub-system design optimization needs. Several accessibility cases of major Safety Important Components (SIC-1) were studied giving important requirements to the design at an early stage. CEA, IRFM, in close collaboration with expertise of CEA, LIST for VR simulation software, applies VR technologies for designing RH maintenance scenario for ITER Test Blanket System (TBS) and Ion cyclotron Resonance Heating (ICRH) Port Plugs. RH compatibility studies using VR pointed out major design drivers while helping to propose credible solution. VR platform is intensively used in the design of WEST (Tungsten (W) Environment Steady-state Tokamak) components and assembly studies, providing important information about the feasibility of assembly processes, optimization of physical mock-ups and ergonomic posture and gestures of operator. Finally, new perspectives, as the integration of

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2000-01-01

SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2000-07-01

SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

Identification by irradiation, in vitro, of two components of erythroprotein action

International Nuclear Information System (INIS)

Barcos, M.

1978-01-01

The effect of ionizing radiation on the response of normal cultured rat marrow cells to erythropoietin yielded two-component inactivation curves for induced iron uptake and hemoglobin synthesis. The radioresistant component of the induced hemoglobin response (1) was detected earlier, at 6 to 20 hr after irradiation, (2) had a DO 0 > or = to 900 R, (3) gave a nonlinear erythropoietin dose--response plot at 600 R, (4) disappeared when marrow from erythremic rats was used, and (5) showed maximal inactivation by 500 R when irradiation preceded hormone addition by 1.5 to 2.5 hr. The radiosensitive component (1) was observed without any contribution from the radioresistant component when the time of assay of normal marrow was postponed from 6 to 20 hr to 20 to 44 hr of culture, (2) had a D 0 = 63 R, (3) gave linear erythropoietin dose--response curves at 15 to 60 R, and (4) showed enhanced inhibition by 60 R if irradiation either preceded or followed hormone addition by 3 hr or more

Development of radiation fusion biotechnology

International Nuclear Information System (INIS)

Jung, Uhee; Lee, Ju Woon; Park, Sang Hyun

2012-04-01

Development of Radiation Fusion Technology with Food Technology by the Application of High Dose Irradiation - To develop fundamental technology using high dose irradiation, effects of high dose irradiation on food components, combined effects of irradiation with food engineering, irradiation condition to destroy radiation resistant foodborne bacteria were studied. - To develop E-beam irradiation technology, irradiation conditions for E-beam and domination effects of E-beam irradiation were determined. The physical marker for E beam irradiated foods or not was developed. - To develop purposed foods to extreme environmental, ready to eat foods and low toxic animal feeds were developed. Through the fundamental researches, the legislation for new irradiated foods and application of E-beam was introduced. Development of modulators against degenerative aging using radiation fusion technology - Selection of 20 kinds of degenerative aging biomarkers related to immune/hematopoiesis, oxidative damage, molecular signaling, lipid metabolism - Establishment of optimal radiation application conditions for aging modeling (fractionated irradiation of total 5Gy, a lapse of 4 months or more - Selection of effective aging modulating substances by screening of 800 natural substances - Development of 1 multi-functional and high-efficacy aging modulator by combination of effective substances and evaluation by in vivo models Development of biochips and kits using RI detection technology for life science - Establishment of kinase substrate interaction analysis using RI detection technique (More than 100 times detection sensitivity compared to conventional fluorescence detection techniques). - The RI detection technique reduces the overall experiment time, as the use of blocking agent can be avoided, offer minimum non specific binding, and facilitates a rapid data analysis with a simplify the process of chip manufacturing. - Establishment of multi-channel type Lab on a chip (LOC) using

Development of radiation fusion biotechnology

Energy Technology Data Exchange (ETDEWEB)

Jung, Uhee; Lee, Ju Woon; Park, Sang Hyun

2012-04-15

Development of Radiation Fusion Technology with Food Technology by the Application of High Dose Irradiation - To develop fundamental technology using high dose irradiation, effects of high dose irradiation on food components, combined effects of irradiation with food engineering, irradiation condition to destroy radiation resistant foodborne bacteria were studied. - To develop E-beam irradiation technology, irradiation conditions for E-beam and domination effects of E-beam irradiation were determined. The physical marker for E beam irradiated foods or not was developed. - To develop purposed foods to extreme environmental, ready to eat foods and low toxic animal feeds were developed. Through the fundamental researches, the legislation for new irradiated foods and application of E-beam was introduced. Development of modulators against degenerative aging using radiation fusion technology - Selection of 20 kinds of degenerative aging biomarkers related to immune/hematopoiesis, oxidative damage, molecular signaling, lipid metabolism - Establishment of optimal radiation application conditions for aging modeling (fractionated irradiation of total 5Gy, a lapse of 4 months or more - Selection of effective aging modulating substances by screening of 800 natural substances - Development of 1 multi-functional and high-efficacy aging modulator by combination of effective substances and evaluation by in vivo models Development of biochips and kits using RI detection technology for life science - Establishment of kinase substrate interaction analysis using RI detection technique (More than 100 times detection sensitivity compared to conventional fluorescence detection techniques). - The RI detection technique reduces the overall experiment time, as the use of blocking agent can be avoided, offer minimum non specific binding, and facilitates a rapid data analysis with a simplify the process of chip manufacturing. - Establishment of multi-channel type Lab on a chip (LOC) using

IFMIF suitability for evaluation of fusion functional materials

International Nuclear Information System (INIS)

Casal, N.; Sordo, F.; Mota, F.; Jordanova, J.; Garcia, A.; Ibarra, A.; Vila, R.; Rapisarda, D.; Queral, V.; Perlado, M.

2011-01-01

The International Fusion Materials Irradiation Facility (IFMIF) is a future neutron source based on the D-Li stripping reaction, planned to test candidate fusion materials at relevant fusion irradiation conditions. During the design of IFMIF special attention was paid to the structural materials for the blanket and first wall, because they will be exposed to the most severe irradiation conditions in a fusion reactor. Also the irradiation of candidate materials for solid breeder blankets is planned in the IFMIF reference design. This paper focuses on the assessment of the suitability of IFMIF irradiation conditions for testing functional materials to be used in liquid blankets and diagnostics systems, since they are been also considered within IFMIF objectives. The study has been based on the analysis and comparison of the main expected irradiation parameters in IFMIF and DEMO reactor.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-10-01

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

Modeling of cascade and sub-cascade formation at high pka energies in irradiated fusion structural materials

International Nuclear Information System (INIS)

Ryazanov, A.; Metelkin, E.V.; Semenov, E.A.

2007-01-01

Full text of publication follows: A new theoretical model is developed for the investigations of cascade and sub-cascade formation in fusion structural materials under fast neutron irradiation at high primary knock atom (PKA) energies. Under 14 MeV neutron irradiation especially of light fusion structural materials such as Be, C, SiC materials PKA will have the energies up to 1 MeV. At such high energies it is very difficult to use the Monte Carlo or molecular dynamic simulations. The developed model is based on the analytical consideration of elastic collisions between displaced moving atoms into atomic cascades produced by a PKAs with the some kinetic energy obtained from fast neutrons. The Tomas-Fermy interaction potential is used for the describing of elastic collisions between moving atoms. The suggested model takes into account also the electronic losses for moving atoms between elastic collisions. The self consistent criterion for sub-cascade formation is suggested here which is based on the comparison of mean distance between two consequent PKA collisions and size of sub-cascade produced by PKA. The analytical relations for the most important characteristics of cascades and sub-cascade are determined including the average number of sub-cascades per one PKA in the dependence on PKA energy, the distance between sub-cascades and the average cascade and sub-cascade sizes as a function of PKA energy. The developed model allows determining the total numbers, distribution functions of cascades and sub-cascades in dependence on their sizes and generation rate of cascades and sub-cascades for different fusion neutron energy spectra. Based on the developed model the numerical calculations for main characteristics of cascades and sub-cascades in different fusion structural materials are performed using the neutron flux and PKA energy spectra for fusion reactors: ITER and DEMO. The main characteristics for cascade and sub-cascade formation are calculated here for the

The manufacture of carbon armoured plasma-facing components for fusion devices

International Nuclear Information System (INIS)

Schedler, B.; Huber, T.; Zabernig, A.; Rainer, F.; Scheiber, K.H.; Schedle, D.

2001-01-01

Within the last decade Plansee has been active in the development and manufacture of different plasma-facing-components for nuclear fusion experiments consisting in a tungsten or CFC-armor joined onto metallic substrates like TZM, stainless steel or copper-alloys. The manufacture of these components requires unique joining technologies in order to obtain reliable thermo mechanical stable joints able to withstand highest heat fluxes without any deterioration of the joint. In an overview the different techniques will be presented by some examples of components already manufactured and successfully tested under high heat flux conditions. Furthermore an overview will be given on the manufacture of different high heat flux components for TORE SUPRA, Wendelstein 7-X and ITER. (author)

Optimization of tungsten-steel joints for plasma facing components in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Heuer, Simon; Linsmeier, Christian [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik, Juelich (Germany); Weber, Thomas; Linke, Jochen [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Werkstoffstruktur und -eigenschaften, Juelich (Germany); Matejicek, Jiri [Institute of Plasma Physics, Academy of Sciences of the Czech Republic, Prague (Czech Republic)

2015-07-01

Tungsten, joint to a martensitic-ferritic EUROFER97 structure, is a promising plasma facing material composite for fusion reactors. Due to the effect of mismatch in thermo-mechanical properties direct bonding is not feasible. Current research is therefore ongoing on interlayer systems. While the adhesion was already improved by the utilization of a discrete Cu, Ti or V interlayer, that is able to relax stresses by plastic deformation, joints still do not resist the expected load cycles in a fusion reactor. Therefore, alternatives for the interface are needed. This contribution presents research on functionally graded materials (FGM). The particular microstructure of a graded interlayer allows re-distributing macro stresses from a discrete interface to a greater volume while avoiding in particular Cu which tends to swell under neutron irradiation. A parameter study on the basis of finite element analysis will be presented as well as first results of several processing routes for FGM that shall be evaluated and benchmarked by mechanical as well as thermal testing.

Development of materials for the fusion nuclear energy system

International Nuclear Information System (INIS)

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

2005-01-01

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

Analysis of electron spin resonance spectra of irradiated gingers: Organic radical components derived from carbohydrates

International Nuclear Information System (INIS)

Yamaoki, Rumi; Kimura, Shojiro; Ohta, Masatoshi

2010-01-01

Electron spin resonance (ESR) spectral characterizations of gingers irradiated with electron beam were studied. Complex asymmetrical spectra (near g=2.005) with major spectral components (line width=2.4 mT) and minor signals (at 6 mT apart) were observed in irradiated gingers. The spectral intensity decreased considerably 30 days after irradiation, and continued to decrease steadily thereafter. The spectra simulated on the basis of characteristics of free radical components derived from carbohydrates in gingers are in good agreement with the observed spectra. Analysis showed that shortly after irradiation the major radical components of gingers were composed of radical species derived from amylose and cellulose, and the amylose radicals subsequently decreased considerably. At 30 days after irradiation, the major radical components of gingers were composed of radical species derived from cellulose, glucose, fructose or sucrose.

Accelerator conceptual design of the international fusion materials irradiation facility

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, M.; Kinsho, M. [Japan Atomic Energy Res. Inst., Tokai, Ibaraki (Japan). Intense Neutron Source Lab.; Jameson, R.A.; Blind, B. [Los Alamos National Lab., NM (United States); Teplyakov, V. [Institute for High Energy Physics, Moscow (Russian Federation); Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J. [Northrop Grumman Corp., Bethpage, NY (United States); Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K. [Johann Wolfgang Goethe Univ., Frankfurt (Germany). Inst. fur Angewandte Phys.; Ferdinand, R.; Lagniel, J.-M. [CEA Saclay LNS, Gif-sur-Yvette (France); Miyahara, A. [Teikyo Univ., Tokyo (Japan); Olivier, M. [CEA DSM, Saclay, Gif-sur-Yvette (France); Piechowiak, E. [Northrop Grumman Corp., Baltimore, MD (United States); Tanabe, Y. [Toshiba Corp., Tsurumi-ku, Yokohama (Japan)

1998-10-01

The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.) 8 refs.

Neutron irradiation results for the LHCb silicon tracker data readout system components

CERN Document Server

Vollhardt, A

2003-01-01

This note reports irradiation data for different components of the LHCb Silicon Tracker data readout system, which will be exposed to neutron fluences due to their location in the readout link service box on the tracking station frame. The components were part of a neutron irradiation campaign in April 2003 at the Prospero reactor at CEA Valduc (France) and were exposed to fluences 5 to 100 times higher than the expected fluences at the experiment. For all tested components, minor or no influence on device performance was measured. We therefore consider the tested components to be compatible with the expected neutron fluences at the foreseen locations in the LHCb experiment.

Electron spin resonance characterization of radical components in irradiated black pepper skin and core

International Nuclear Information System (INIS)

Yamaoki, Rumi; Kimura, Shojiro; Ohta, Masatoshi

2011-01-01

Characteristics of free radical components of irradiated black pepper fruit (skin) and the pepper seed (core) were analyzed using electron spin resonance. A weak signal near g=2.005 was observed in black pepper before irradiation. Complex spectra near g=2.005 with three lines (the skin) or seven lines (the core) were observed in irradiated black pepper (both end line width; ca. 6.8 mT). The spectral intensities decreased considerably at 30 days after irradiation, and continued to decrease steadily thereafter. The spectra simulated on the basis of the content and the stability of radical components derived from plant constituents, including fiber, starch, polyphenol, mono- and disaccharide, were in good agreement with the observed spectra. Analysis showed that the signal intensities derived from fiber in the skin for an absorbed dose were higher, and the rates of decrease were lower, than that in the core. In particular, the cellulose radical component in the skin was highly stable. - Highlights: → We identified the radical components in irradiated black pepper skin and core. → The ESR spectra near g=2.005 with 3-7 lines were emerged after irradiation. → Spectra simulated basing on the content and the stability of radical from the plant constituents. → Cellulose radical component in black pepper skin was highly stable. → Single signal near g=2.005 was the most stable in black pepper core.

High quality actively cooled plasma-facing components for fusion

International Nuclear Information System (INIS)

Nygren, R.E.

1995-01-01

This paper interweaves some suggestions for developing actively cooled plasma-facing components (PFCs) for future fusion devices, with supporting examples taken from the design, fabrication and operation of Tore Supra's Phase III outboard pump limiter (OPL). This actively cooled midplane limiter, designed for heat and particle removal during long-pulse operation, has been operated under essentially thermally steady state conditions. Testing to identify braze flaws, analysis of the impact of joining flaws on the thermal-hydraulic performance of the OPL, and the extensive calorimetry and IR thermography used to confirm and update safe operating limits for power handling of the OPL are reviewed. This experience suggests that, for PFCs in future fusion devices, flaw-tolerant designs are possible; analyses of the impacts of flaws on performance can provide criteria for quality assurance; and validating appropriate methods of inspection for such flaws early in the design development of PFCs is prudent. The need for in-service monitoring is also discussed. (orig.)

Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

International Nuclear Information System (INIS)

Lund, A.L.

1997-11-01

In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path

PFMC14. 14th international conference on plasma-facing materials and components for fusion applications. Book of abstracts

International Nuclear Information System (INIS)

2013-01-01

The performance of fusion devices and of a future fusion power plant critically depends on the plasma facing materials and components. Resistance to local heat and particle loads, thermo-mechanical properties, as well as the response to neutron damage of the selected materials are critical parameters which need to be understood and tailored from atomistic to component levels. The 14th International Conference on Plasma-Facing Materials and Components for Fusion Applications addresses these issues. Among the topics of the joint conference recent developments and research results in the following fields are addressed: - Tungsten and tungsten alloys - Low-Z materials - Mixed materials - Erosion, redeposition and fuel retention - Materials under extreme thermal loads - Technology and testing of plasma-facing components - Neutron effects in plasma-facing materials - Advanced characterization of materials and components. Selected international speakers present overview lectures and treat detailed aspects of the given topics. Contributed papers to the subjects of the meeting are solicited for oral and poster presentations.

Surface erosion of fusion reactor components due to radiation blistering and neutron sputtering

International Nuclear Information System (INIS)

Das, S.K.; Kaminsky, M.

1975-01-01

Radiation blistering and neutron sputtering can lead to the surface erosion of fusion reactor components exposed to plasma radiations. Recent studies of methods to reduce the surface erosion caused by these processes are discussed

Large aperture components for solid state laser fusion systems

International Nuclear Information System (INIS)

Simmons, W.W.

1978-01-01

Solid state lasers for fusion experiments must reliably deliver maximum power to small (approximately .5 mm) targets from stand-off focal distances of 1 m or more. This requirement places stringent limits upon the optical quality, resistance to damage, and overall performance of the several major components--amplifiers, Faraday isolators, spatial filters--in each amplifier train. Component development centers about achieving (1) highest functional material figure of merit, (2) best optical quality, and (3) maximum resistance to optical damage. Specific examples of the performance of large aperture components will be presented within the context of the Argus and Shiva laser systems, which are presently operational at Lawrence Livermore Laboratory. Shiva comprises twenty amplifiers, each of 20 cm output clear aperture. Terawatt beams from these amplifiers are focused through two opposed, nested clusters of f/6 lenses onto such targets. Design requirements upon the larger aperture Nova laser components, up to 35 cm in clear aperture, will also be discussed; these pose a significant challenge to the optical industry

Present activities for the preparation of a Japanese draft of structural design guidelines for the experimental fusion reactor

International Nuclear Information System (INIS)

Miya, K.; Muto, Y.; Takatsu, H.; Hada, K.; Koizumi, K.; Jitsukawa, S.; Arai, T.; Ohkawa, Y.; Shimakawa, T.; Aoto, K.; Shiraishi, H.; Takagi, T.; Miki, N.; Takahashi, S.; Sato, K.; Takemasa, F.; Kasaba, M.; Kudough, F.; Fujita, J.; Kajiura, S.; Kinoshita, S.

1996-01-01

Since November 1990, systematic research has been carried out in preparation for a Japanese draft of structural design guidelines for the experimental fusion reactor. This report summarizes the major results of the work and the status of these efforts. A classification of components and definition of operating conditions are proposed on the basis of the ITER-CDA design, in the light of the safety characteristics of the fusion reactor and relevant conventions for the existing fission reactor design code. Specific issues regarding the structural design of the experimental fusion reactor are discussed based on the experimental and analytical work. The validity of the existing structural design method is confirmed for the use of irradiated 316 SS, irrespective of the significant reduction in uniform elongation capability caused by heavy neutron irradiation. Further important phenomena are treated such as magnetic damping, magnetic stiffness and fracture due to electromagnetic forces. Finally, the issues concerned with welding and non-destructive examinations are discussed with relevance to component classification. (orig.)

Proposed rf system for the fusion materials irradiation test facility

International Nuclear Information System (INIS)

Fazio, M.V.; Johnson, H.P.; Hoffert, W.J.; Boyd, T.J.

1979-01-01

Preliminary rf system design for the accelerator portion of the Fusion Materials Irradiation Test (FMIT) Facility is in progress. The 35-MeV, 100-mA, cw deuteron beam will require 6.3 MW rf power at 80 MHz. Initial testing indicates the EIMAC 8973 tetrode is the most suitable final amplifier tube for each of a series of 15 amplifier chains operating at 0.5-MW output. To satisfy the beam dynamics requirements for particle acceleration and to minimize beam spill, each amplifier output must be controlled to +-1 0 in phase and the field amplitude in the tanks must be held within a 1% tolerance. These tolerances put stringent demands on the rf phase and amplitude control system

COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

Directory of Open Access Journals (Sweden)

Phani Kumar Domalapally

2015-04-01

Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

Design issues and implications for the structural integrity and lifetime of fusion power plant components

International Nuclear Information System (INIS)

Karditas, P.J.

1996-05-01

This review discusses, with example calculations, the criteria, and imposed constraints and limitations, for the design of fusion components and assesses the implications for successful design and power plant operation. The various loading conditions encountered during the operation of a tokamak lead to structural damage and possible failure by such mechanisms as yielding, thermal creep rupture and fatigue due to thermal cycling, plastic strain cycling (ratcheting), crack growth-propagation and radiation induced swelling and creep. Of all the possible damage mechanisms, fatigue, creep and their combination are the most important in the structural design and lifetime of fusion power plant components operating under steady or load varying conditions. Also, the effect of neutron damage inflicted onto the structural materials and the degradation of key properties is of major concern in the design and lifetime prediction of components. Structures are classified by, and will be restricted by existing or future design codes relevant to medium and high temperature power plant environments. The ways in which existing design codes might be used in present and near future design activities, and the implications, are discussed; the desirability of an early start towards the development of fusion-specific design codes is emphasised. (UK)

Modeling of cascade and sub-cascade formation at high PKA energies in irradiated fusion structural materials

International Nuclear Information System (INIS)

Ryazanov, A.I.; Metelkin, E.V.; Semenov, E.V.

2009-01-01

A new theoretical model is developed for the investigations of cascade and sub-cascade formation in fusion structural materials under fast neutron irradiation at high primary knock-on atom energies. Light fusion structural materials: such as Be, C and SiC under 14 MeV neutron irradiation in fusion reactor will have the primary knock-on atoms with the energies up to 1 MeV. It is very difficult to use at such high energies the Monte-Carlo or molecular dynamic simulations [H.L. Heinisch, B.N. Singh, Philos. Mag. A67 (1993) 407; H.L. Heinisch, B.N. Singh, J. Nucl. Mater. 251 (1997) 77]. The developed model is based on the analytical consideration of elastic collisions between displaced moving atoms produced by primary knock-on atoms with some kinetic energies obtained from fast neutrons and crystal lattice atoms. The Thomas-Fermi interaction potential is used here for the description of these elastic atomic collisions. The suggested model takes into account also the electronic losses for moving atoms between elastic collisions. The self-consistent criterion for sub-cascade formation is suggested here which is based on the comparison of mean distance of primary knock-on atoms between consequent collisions of them with the target atoms and a size of sub-cascade produced by moving secondary knock-on atoms produced in such collisions. The analytical relations for the most important characteristics of cascades and sub-cascades are determined including the average number of sub-cascades per one primary knock-on atom in the dependence on its energy, the distance between sub-cascades and the average cascade and sub-cascade sizes. The developed model allows determining the total numbers, distribution functions of cascades and sub-cascades in dependence on their sizes and generation rate of cascades and sub-cascades for the different fusion neutron energy spectra. On the basis of this developed model the numerical calculations for main characteristics of cascades and sub

Estimates of time-dependent fatigue behavior of Type 316 stainless steel subject to irradiation damage in fast breeder and fusion power reactor systems

International Nuclear Information System (INIS)

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

1978-01-01

Cyclic lives obtained from strain-controlled fatigue tests at 593 0 C of specimens irradiated in the experimental breeder reactor II (EBR-II) to a fluence of 1 to 2.63*10 26 neutrons (n)/m 2 (E>0.1 MeV) were compared with predictions based on the method of strain-range partitioning. It was demonstrated that, when appropriate tensile and creep-rupture ductilities were employed, reasonably good estimates of the influence of hold periods and irradiation damage on the fully reversed fatigue life of Type 316 stainless steel could be made. After applicability of this method was demonstrated, ductility values for 20 percent cold-worked Type 316 stainless steel specimens irradiated in a mixed-spectrum fission reactor were used to estimate fusion reactor first-wall lifetime. The ductility values used were from irradiations that simulate the environment of the first wall of a fusion reactor. Neutron wall loadings ranging from 2 to 5 MW/m 2 were used. 27 refs

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)

Re-weldability tests of irradiated Inconel 625 by TIG welding method

International Nuclear Information System (INIS)

Tsuchiya, K.; Shimizu, M.; Kawamura, H.; Matsuda, F.; Kalinin, G.

1998-01-01

Inconel 625 is one of the possible materials for the vacuum vessel (VV) and for the in-vessel components of fusion reactors where high strength and high electrical resistance are required. In particular, Inconel 625 is used for the VV of JET and for flexible branch pipe lines in the ITER design. One of the most important issues for their applications is its re-weldability between un-irradiated and irradiated materials. This has a large impact on the design of in-vessel components. In this study, re-weldability of un-irradiated and/or irradiated Inconel 625 that has been welded by the tungsten inert gas (TIG) welding process has been examined, and effect of helium generation amount on mechanical properties of the weld joint has been discussed. (authors)

Composites as structural materials in fusion reactors

International Nuclear Information System (INIS)

Megusar, J.

1989-01-01

In fusion reactors, materials are used under extreme conditions of temperature, stress, irradiation, and chemical environment. The absence of adequate materials will seriously impede the development of fusion reactors and might ultimately be one of the major difficulties. Some of the current materials problems can be solved by proper design features. For others, the solution will have to rely on materials development. A parallel and balanced effort between the research in plasma physics and fusion-related technology and in materials research is, therefore, the best strategy to ultimately achieve economic, safe, and environmentally acceptable fusion. The essential steps in developing composites for structural components of fusion reactors include optimization of mechanical properties followed by testing under fusion-reactor-relevant conditions. In optimizing the mechanical behavior of composite materials, a wealth of experience can be drawn from the research on ceramic matrix and metal matrix composite materials sponsored by the Department of Defense. The particular aspects of this research relevant to fusion materials development are methodology of the composite materials design and studies of new processing routes to develop composite materials with specific properties. Most notable examples are the synthesis of fibers, coatings, and ceramic materials in their final shapes form polymeric precursors and the infiltration of fibrous preforms by molten metals

DT fusion neutron irradiation of LLL Nb3Sn and LLL superconductor wires at 4.20K

International Nuclear Information System (INIS)

MacLean, S.C.

1977-01-01

The DT fusion neutron irradiation of one LLL superconductor wire and one LLL Nb 3 Sn foil at 4.2 0 K is described. The sample position, beam-on time, and neutron dose record are given. The results from two ''profile'' dosimetry foils measuring the lateral variation in neutron flux are included

Irradiation-induced segregation in multi-component alloys

International Nuclear Information System (INIS)

Chen, I.W.

1983-01-01

A unified analysis of irradiation-induced segregation in multi-component alloys is developed using the formulation of irreversible thermodynamics. Three distinct mechanisms for segregation, namely the inverse Kirkendall effect, the vacancy-wind effect, and the solute drag of interstitials, are identified. In particular, the inverse Kirkendall effect due to interstitials arises only if a solute-interstitial interaction or a mutual conversion among interstitials via lattice atom intermediaries operates simultaneously. In the limit of fast conversion a para-equilibrium state may be reached between interstitials and lattice atoms, and the interstitial mechanism becomes formally analogous to the vacancy mechanism. Although the past treatment of rate phenomena in this field was apparently limited to the latter case, the importance of the consideration of separate chemical potentials for interstitials of different species, in segregation and other irradiation effects, is emphasized. (orig.)

Material testing facilities and programs for plasma-facing component testing

Science.gov (United States)

Linsmeier, Ch.; Unterberg, B.; Coenen, J. W.; Doerner, R. P.; Greuner, H.; Kreter, A.; Linke, J.; Maier, H.

2017-09-01

Component development for operation in a large-scale fusion device requires thorough testing and qualification for the intended operational conditions. In particular environments are necessary which are comparable to the real operation conditions, allowing at the same time for in situ/in vacuo diagnostics and flexible operation, even beyond design limits during the testing. Various electron and neutral particle devices provide the capabilities for high heat load tests, suited for material samples and components from lab-scale dimensions up to full-size parts, containing toxic materials like beryllium, and being activated by neutron irradiation. To simulate the conditions specific to a fusion plasma both at the first wall and in the divertor of fusion devices, linear plasma devices allow for a test of erosion and hydrogen isotope recycling behavior under well-defined and controlled conditions. Finally, the complex conditions in a fusion device (including the effects caused by magnetic fields) are exploited for component and material tests by exposing test mock-ups or material samples to a fusion plasma by manipulator systems. They allow for easy exchange of test pieces in a tokamak or stellarator device, without opening the vessel. Such a chain of test devices and qualification procedures is required for the development of plasma-facing components which then can be successfully operated in future fusion power devices. The various available as well as newly planned devices and test stands, together with their specific capabilities, are presented in this manuscript. Results from experimental programs on test facilities illustrate their significance for the qualification of plasma-facing materials and components. An extended set of references provides access to the current status of material and component testing capabilities in the international fusion programs.

Beryllium R and D for fusion applications

International Nuclear Information System (INIS)

Scaffidi-Argentina, F.; Longhurst, G.R.; Shestakov, V.; Kawamura, H.

2000-01-01

Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma-structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal

Beryllium R and D for fusion applications

Energy Technology Data Exchange (ETDEWEB)

Scaffidi-Argentina, F. E-mail: francesco.scaffidi@iket.fzk.de; Longhurst, G.R.; Shestakov, V.; Kawamura, H

2000-11-01

Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma-structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal.

FOREWORD: 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications

Science.gov (United States)

Kreter, Arkadi; Linke, Jochen; Rubel, Marek

2009-12-01

The 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications (PFMC-12) was held in Forschungszentrum Jülich (FZJ) in Germany in May 2009. This symposium is the successor to the International Workshop on Carbon Materials for Fusion Applications series. Between 1985 and 2003, 10 'Carbon Workshops' were organized in Jülich, Stockholm and Hohenkammer. After this time, the scope of the symposium was redefined to reflect the new requirements of ITER and the ongoing evolution of the field. The workshop was first organized under its new name in 2006 in Greifswald, Germany. The main objective of this conference series is to provide a discussion forum for experts from research institutions and industry dealing with materials for plasma-facing components in present and future controlled fusion devices. The operation of ASDEX-Upgrade with tungsten-coated wall, the fast progress of the ITER-Like Wall Project at JET, the plans for the EAST tokamak to install tungsten, the start of ITER construction and a discussion about the wall material for DEMO all emphasize the importance of plasma-wall interactions and component behaviour, and give much momentum to the field. In this context, the properties and behaviour of beryllium, carbon and tungsten under plasma impact are research topics of foremost relevance and importance. Our community realizes both the enormous advantages and serious drawbacks of all the candidate materials. As a result, discussion is in progress as to whether to use carbon in ITER during the initial phase of operation or to abandon this element and use only metal components from the start. There is broad knowledge about carbon, both in terms of its excellent power-handling capabilities and the drawbacks related to chemical reactivity with fuel species and, as a consequence, about problems arising from fuel inventory and dust formation. We are learning continuously about beryllium and tungsten under fusion conditions, but our

The effect of gamma irradiation on curcumin component of Curcuma domestica

Science.gov (United States)

Chosdu, R.; Erizal; Iriawan, T.; Hilmy, N.

1995-02-01

The effect of gamma irradiation on curcumin component of Curcuma domestica rhizome were investigated. Pure curcumin, sliced and powdered rhizome with 10% of moisture content were irradiated at 0, 10, 30 and 50 kGy (dose rate of 6 kGy/h). Curcumin content was analysed using HPLC method and ESR spectra. Results show that free radicals are already present in unirradiated rhizome. Gamma irradiation at the doses of 10, 30 and 50 kGy induced the free radicals formation of pure curcumin and Curcuma domestica rhizome. The ESR spectra of irradiated rhizome gave a very similar spectra to the signal of irradiated pure curcumin. The percentage of free radicals intensity from pure curcumin was very stable at room temperature up to 670 hours of storage. However, the percentage intensity of free radicals in the irradiated rhizome were decay during storage. Irradiation treatment and storage time did not give a significant change on curcumin content, water activity, pH and moisture content of rhizome investigated.

The effect of gamma irradiation on curcumin component of Curcuma domestica

Energy Technology Data Exchange (ETDEWEB)

Chosdu, R.E.; Erizal; Iriawan, T.; Hilmy, N. [National Atomic Energy Agency, Jakarta (Indonesia). Center for Applications of Isotopes and Radiation

1995-10-01

The effect of gamma irradiation on curcumin component of Curcuma domestica rhizome were investigated. Pure curcumin, sliced and powdered rhizome with 10% of moisture content were irradiated at 0, 10, 30 and 50 kGy (dose rate of 6 kGy/h). Curcumin content was analysed using HPLC method and ESR spectra. Results show that free radicals are already present in unirradiated rhizome. Gamma irradiation at the doses of 10, 30 and 50 kGy induced the free radicals formation of pure curcumin and curcuma domestica rhizome. The ESR spectra of irradiated rhizome gave a very similar spectra to the signal of irradiated pure curcumin. The percentage of free radicals intensity from pure curcumin was very stable at room temperature up to 670 hours of storage. However, the percentage intensity of free radicals in the irradiated rhizome were decay during storage. Irradiation treatment and storage time did not give a significant change on curcumin content, water activity, pH and moisture content of rhizome investigated. (Author).

The effect of gamma irradiation on curcumin component of Curcuma domestica

International Nuclear Information System (INIS)

Chosdu, R.E.; Erizal; Iriawan, T.; Hilmy, N.

1995-01-01

The effect of gamma irradiation on curcumin component of Curcuma domestica rhizome were investigated. Pure curcumin, sliced and powdered rhizome with 10% of moisture content were irradiated at 0, 10, 30 and 50 kGy (dose rate of 6 kGy/h). Curcumin content was analysed using HPLC method and ESR spectra. Results show that free radicals are already present in unirradiated rhizome. Gamma irradiation at the doses of 10, 30 and 50 kGy induced the free radicals formation of pure curcumin and curcuma domestica rhizome. The ESR spectra of irradiated rhizome gave a very similar spectra to the signal of irradiated pure curcumin. The percentage of free radicals intensity from pure curcumin was very stable at room temperature up to 670 hours of storage. However, the percentage intensity of free radicals in the irradiated rhizome were decay during storage. Irradiation treatment and storage time did not give a significant change on curcumin content, water activity, pH and moisture content of rhizome investigated. (Author)

Design of a tokamak fusion reactor first wall armor against neutral beam impingement

International Nuclear Information System (INIS)

Myers, R.A.

1977-12-01

The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiation damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem

The effect of X-ray irradiation on a red cell component in WB, WRC and LPRC

International Nuclear Information System (INIS)

Tayama, Tatsuya; Toyota, Kuroh; Nagahashi, Hisakata; Masuyama, Tetsuya; Haneda, Kenji; Juji, Takeo.

1990-01-01

In spite of the use of X-ray irradiation on blood products, few data about its effect on components are reported. We need more informations about a quality of irradiated red cell components. This study shows in vitro changes of irradiated red cell component in WB, WRC and LPRC as the minimum dose of 1,500, 3,000, and 5,000 rads. The fact as follows were observed in response to irradiated doses: 1) increased fragility of red cell membrane, 2) increased amount of plasma K and plasma Hb, and 3) decrease of ATP in WB.2,3-DPG, glucose, pH, Ht and Cl. The numbers of RBC, WBC and Platelet were not affected by irradiation with doses between 1,500 and 5,000 rads. According to these results, the followings are recommended: 1) irradiation with 1,500 rads is a proper method for WB, 2) in order to avoid the risk of increased plasma K, WB should be used within 1 week after irradiation, and WRC and LPRC should be used 24 hours after irradiation. (author)

Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

International Nuclear Information System (INIS)

Lee, Ho Jin; Lee, B. S.; Kim, K. B.

2003-09-01

The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si 3 N 4 . Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation

Incomplete momentum transfer components in /sup 16/O + /sup 12/C Fusion at 20 MeV/A

Energy Technology Data Exchange (ETDEWEB)

Menchaca-Rocha, A.; Brandan, M.E. (Grenoble-1 Univ., 38 (France). Inst. des Sciences Nucleaires); Dacal, A.; Galindo, A.; Mahoney, J.; Murphy, M.; Rae, W.D.M. (California Univ., Berkeley (USA). Lawrence Berkeley Lab.)

1983-01-27

The energy spectra of Z = 3-9 particles from reactions induced by 20 MeV/A /sup 16/O on /sup 12/C have been measured. The systematics found from the fusion-like residues clearly deviate from those expected for complete fusion, giving evidence for important incomplete momentum transfer components.

Managing the fusion burn to improve symbiotic system performance

International Nuclear Information System (INIS)

Renier, J.P.; Martin, J.G.

1979-01-01

Symbiotic power systems, in which fissile fuel is produced in fusion-powered factories and burned in thermal reactors characterized by high conversion ratios, constitute an interesting near-term fusion application. It is shown that the economic feasibility of such systems depend on adroit management of the fusion burn. The economics of symbiotes is complex: reprocessing and fabrication of the fusion reactor blankets are important components of the production cost of fissile fuel, but burning fissile material in the breeder blanket raises overall costs and lowers the support ratio. Analyses of factories which assume that the fusion power is constant during an irradiation cycle underestimate their potential. To illustrate the effect of adroit engineering of the fusion burn, this paper analyzes systems based on D-T and semi-catalyzed D-D fusion-powered U-233 breeders. To make the D-T symbiote self-sufficient, tritium is bred in separate lithium blankets designed so as to minimize overall costs. All blankets are assumed to have spherical geometry, with 85% closure. Neutronics depletion calculations were performed with a revised version of the discrete ordinates code XSDRN-PM, using multigroup (100 neutron, 21 gamma-ray groups) coupled cross-section libraries

Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

Energy Technology Data Exchange (ETDEWEB)

Rowcliffe, A.F. [Oak Ridge National Laboratory, TN (United States); Tsai, H.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

1997-08-01

The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

Irradiation behavior of bonded structures: impact of stress-enhanced swelling on irradiation creep and elastic properties

International Nuclear Information System (INIS)

Hassan, M.H.; Blanchard, J.P.; Kulcinski, G.L.

1992-01-01

The objective of this work is to understand the factors that govern the adhesion of coatings on fusion reactor first walls which are subjected to neutron irradiation. Radiation damage will be a major key point in the design of the many duplex components in fusion reactors. There is a substantial amount of available data showing that stress plays a major role in the onset, and possibly the rate, of void growth in austenitic stainless steels. There is also strong support models which predict a coupling of swelling and creep through the stress environment. A parametric study for evidence to stress-enhanced swelling and its connection to creep is conducted for a typical fusion power demonstration reactor. Since microstructural changes are known to affect elastic moduli, the impact of stress enhanced swelling on these moduli are also evaluated

Irradiation tests of critical components for remote handling system in gamma radiation environment

International Nuclear Information System (INIS)

Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi

1996-03-01

This report covers the gamma ray irradiation tests according to the Agreement of ITER R and D Task (T35) in 1994 and describes radiation hardness of the standard components for the ITER remote handling system which are categorized into the robotics (Subtask-1), the viewing system (Subtask-2) and the common components (Subtask-3). The gamma ray irradiation tests have been conducted using No.2 and No.3 cells at the cobalt building of Takasaki Establishment in JAERI. The radiation source is cobalt sixty (Co-60), and the maximum dose rate of No.2 and No.3 cells is about 1x10 6 R/h and 2x10 6 R/h, respectively. The environmental conditions of the irradiation tests are described below and all of components excepting electrical wires have been tested in the No.2 cell. [No.2 cell : Atmosphere and ambient temperature No.3 cell : Nitrogen gas and 250degC] As a whole, many of components have been irradiated up to the rated dose of around 1x10 10 rads and the following main results are obtained. The developed AC servo motor and periscope for radiation use have shown excellent durability with the radiation hardness tolerable for more than 10 9 rads. An electrical connector compatible with remote operation has also shown no degradation of electrical characteristics after the irradiation of 10 10 rads. As for polyimide insulated wires, the mechanical and electrical characteristics are not degradated after the irradiation of 10 9 rads and more radiation hardness can be expected than the anticipation. On the contrary, standard position sensors such as rotary encoder show extremely low radiation hardness and further efforts have to be made for improvements. (J.P.N.)

Fusion Reactor Materials

International Nuclear Information System (INIS)

Moons, F.

1998-01-01

SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

Resonant creep enhancement in austenitic stainless steels due to pulsed irradiation at low doses

International Nuclear Information System (INIS)

Kishimoto, N.; Amekura, H.; Saito, T.

1994-01-01

Steady-state irradiation creep of austenitic stainless steels has been extensively studied as one of the most important design parameters in fusion reactors. The steady-state irradiation creep has been evaluated using in-pile and light-ion experiments. Those creep compliances of various austenitic steels range in the vicinity of ε/Gσ = 10 -6 ∼10 -5 (dpa sm-bullet MPa) -1 , depending on chemical composition etc. The mechanism of steady-state irradiation creep has been elucidated, essentially in terms of stress-induced preferential absorption of point defects into dislocations, and their climb motion. From this standpoint, low doses such as 10 -3 ∼10 -1 dpa would not give rise to any serious creep, and the irradiation creep may not be a critical issue for the low-dose fusion devices including ITER. It is, however, possible that pulsed irradiation causes different creep behaviors from the steady-state one due to dynamic unbalance of interstitials and vacancies. The authors have actually observed anomalous creep enhancement due to pulsed irradiation in austenitic stainless steels. The resonant behavior of creep indicates that pulsed irradiation may cause significant deformation in austenitic steels even at such low doses and slow pulsing rates, especially for the SA-materials. The first-wall materials in plasma operation of ∼10 2 s may suffer from unexpected transient creep, even in the near-term fusion deices, such as ITER. Though this effect might be a transient effect for a relatively short period, it should be taken into account that the pulsed irradiation makes influences on stress relaxation of the fusion components and on the irradiation fatigue. The mechanism and the relevant behaviors of pulse-induced creep will be discussed in terms of a point-defect model based on the resonant interstitial enrichment

Loading nature of the interfacial cracks in a joint component under fusion-relevant thermal loads

International Nuclear Information System (INIS)

You, J.H.

1998-01-01

One of the standard design concepts for divertor components in a fusion reactor is the bonded joint structure. Understanding the loading nature of interfacial cracks are significant for the assessment of structural integrity of divertor joint components. In this paper, the thermomechanical loading nature of interfacial cracks is discussed. A bi-material joint element consisting of the CFC/TZM system is considered. A typical fusion operation condition is simulated assuming a pulsed high heat flux loading. Stress singularities near the interfacial crack tips are characterized quantitatively in terms of the fracture mechanical parameters. The evolution of the stress intensity factors and the energy release rate during the given transient thermal load are determined. The difference in loading characteristics between the edge crack and the center crack is discussed. High heat flux cycling tests are performed on brazed CFC/TZM divertor elements in an electron beam test facility. The microstructures of the damaged interface agree with the predicted fracture modes. The loading nature and possible failure mechanisms are discussed for a fusion-relevant thermal loading. (orig.)

Laser-start-up system for magnetic mirror fusion

International Nuclear Information System (INIS)

Frank, A.M.; Thomas, S.R.; Denhoy, B.S.; Chargin, A.K.

1976-01-01

A CO 2 laser system has been developed at LLL to provide hot start-up plasmas for magnetic mirror fusion experiments. A frozen ammonia pellet is irradiated with a laser power density in excess of 10 13 W/cm 2 in a 50-ns pulse. This system uses commercially available laser systems. Optical components were fabricated both by direct machining and standard techniques. The technologies used in this system are directly applicable to reactor scale systems

Implementation of a design and configuration management platform for fusion components on the Tore Supra WEST Project

Energy Technology Data Exchange (ETDEWEB)

Benoît, Fabrice, E-mail: fabrice-2.benoit@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Allegretti, Ludovic [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Aumeunier, Marie-Hélène [OPTIS, ZE de La Farlède, F-83078 Toulon Cedex 9 (France); Bucalossi, Jérôme; Doceul, Louis; Faïsse, Frederic; Firdaouss, Medhi; Geynet, Michel; Houtte, Didier van; Larroque, Sébastien; Magaud, Philippe; Maini, Patrick; Missirlian, Marc; Parrat, Hélène [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Robert, Julien [SOFYNE, F-69800 Saint Priest (France)

2014-10-15

Highlights: •A design and configuration management platform is under development for managing fusion components lifecycle at CEA. •Design platform ensures an efficient sharing of the data and provides connections between the different software and databases involved in fusion components design. •Design platform rollout on WEST project is ongoing as part of change control and configuration management implementation. -- Abstract: This paper presents the technical solutions and methodologies that are used and under development for managing the design lifecycle of the WEST project (W – for tungsten – Environment in Steady-state Tokamak, upgrade of Tore Supra's with actively cooled tungsten plasma facing components) fusion components and explains the interfaces that are implemented or in construction to connect together the different tools like documents management system, CAD modeler, or simulation codes around the data management backbone. It describes the methodologies used on the WEST project to optimize the design process by managing the engineering data workflow and ensuring the consistency between the different 3D representations for design or analysis as well as the specification or interfaces documents. Finally it explains how this platform contributes to reach the project targets in terms of performance, cost and schedule.

Implementation of a design and configuration management platform for fusion components on the Tore Supra WEST Project

International Nuclear Information System (INIS)

Benoît, Fabrice; Allegretti, Ludovic; Aumeunier, Marie-Hélène; Bucalossi, Jérôme; Doceul, Louis; Faïsse, Frederic; Firdaouss, Medhi; Geynet, Michel; Houtte, Didier van; Larroque, Sébastien; Magaud, Philippe; Maini, Patrick; Missirlian, Marc; Parrat, Hélène; Robert, Julien

2014-01-01

Highlights: •A design and configuration management platform is under development for managing fusion components lifecycle at CEA. •Design platform ensures an efficient sharing of the data and provides connections between the different software and databases involved in fusion components design. •Design platform rollout on WEST project is ongoing as part of change control and configuration management implementation. -- Abstract: This paper presents the technical solutions and methodologies that are used and under development for managing the design lifecycle of the WEST project (W – for tungsten – Environment in Steady-state Tokamak, upgrade of Tore Supra's with actively cooled tungsten plasma facing components) fusion components and explains the interfaces that are implemented or in construction to connect together the different tools like documents management system, CAD modeler, or simulation codes around the data management backbone. It describes the methodologies used on the WEST project to optimize the design process by managing the engineering data workflow and ensuring the consistency between the different 3D representations for design or analysis as well as the specification or interfaces documents. Finally it explains how this platform contributes to reach the project targets in terms of performance, cost and schedule

Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Jin; Lee, B. S.; Kim, K. B

2003-09-01

The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si{sub 3}N{sub 4}. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation.

Activation Inventories after Exposure to DD/DT Neutrons in Safety Analysis of Nuclear Fusion Installations.

Science.gov (United States)

Stankunas, Gediminas; Cufar, Aljaz; Tidikas, Andrius; Batistoni, Paola

2017-11-23

Irradiations with 14 MeV fusion neutrons are planned at Joint European Torus (JET) in DT operations with the objective to validate the calculation of the activation of structural materials in functional materials expected in ITER and fusion plants. This study describes the activation and dose rate calculations performed for materials irradiated throughout the DT plasma operation during which the samples of real fusion materials are exposed to 14 MeV neutrons inside the JET vacuum vessel. Preparatory activities are in progress during the current DD operations with dosimetry foils to measure the local neutron fluence and spectrum at the sample irradiation position. The materials included those used in the manufacturing of the main in-vessel components, such as ITER-grade W, Be, CuCrZr, 316 L(N) and the functional materials used in diagnostics and heating systems. The neutron-induced activities and dose rates at shutdown were calculated by the FISPACT code, using the neutron fluxes and spectra that were provided by the preceding MCNP neutron transport calculations. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Protoplast fusion in Streptomyces: fusions involving ultraviolet-irradiated protoplasts

International Nuclear Information System (INIS)

Hopwood, D.A.; Wright, H.M.

1981-01-01

Protoplasts of Streptomyces coelicolor showed the same ultraviolet killing kinetics as spores. Irradiated protoplasts gave rise to recombinants when they were fused with unirradiated protoplasts of a strain carrying complementary genetic markers. The decline with u.v. fluence in the capacity of irradiated protoplasts to yield recombinants inheriting individual markers was some six times less steep than the survival of unfused protoplasts; thus, for example, protoplasts reduced to only 0.01% survival still yielded 10% as many recombinants as unirradiated protoplasts. Each of six widely separated markers of the irradiated parent was inherited independently of the others, with a frequency falling exponentially with u.v. fluence. (author)

Measurement and modeling of shortwave irradiance components in cloud-free atmospheres

Energy Technology Data Exchange (ETDEWEB)

Halthore, R.N.

1999-08-04

Atmosphere scatters and absorbs incident solar radiation modifying its spectral content and decreasing its intensity at the surface. It is very useful to classify the earth-atmospheric solar radiation into several components--direct solar surface irradiance (E{sub direct}), diffuse-sky downward surface irradiance (E{sub diffuse}), total surface irradiance, and upwelling flux at the surface and at the top-of-the atmosphere. E{sub direct} depends only on the extinction properties of the atmosphere without regard to details of extinction, namely scattering or absorption; furthermore it can be accurately measured to high accuracy (0.3%) with the aid of an active cavity radiometer (ACR). E{sub diffuse} has relatively larger uncertainties both in its measurement using shaded pyranometers and in model estimates, owing to the difficulty in accurately characterizing pyranometers and in measuring model inputs such as surface reflectance, aerosol single scattering albedo, and phase function. Radiative transfer model simulations of the above surface radiation components in cloud-free skies using measured atmospheric properties show that while E{sub direct} estimates are closer to measurements, E{sub diffuse} is overestimated by an amount larger than the combined uncertainties in model inputs and measurements, illustrating a fundamental gap in the understanding of the magnitude of atmospheric absorption in cloud-free skies. The excess continuum type absorption required to reduce the E{sub diffuse} model overestimate ({approximately}3--8% absorptance) would significantly impact climate prediction and remote sensing. It is not clear at present what the source for this continuum absorption is. Here issues related to measurements and modeling of the surface irradiance components are discussed.

Shippingport station decommissioning project irradiated components transfer: Topical report

International Nuclear Information System (INIS)

1988-01-01

This topical report is a synopsis of the transfer of irradiated components into the Shippingport Reactor Pressure Vessel (RPV) performed at the Shippingport Station Decommissioning Project (SSDP). The information is provided as a part of the Technology Transfer Program to document the preparation activities for the decommissioning of a nuclear power reactor to be removed in one piece

Activation and Radiation Damage Behaviour of Russian Structural Materials for Fusion Reactors in the Fission and Fusion Reactors

International Nuclear Information System (INIS)

Blokhin, A.; Demin, N.; Chernov, V.; Leonteva-Smirnova, M.; Potapenko, M.

2006-01-01

Various structural low (reduced) activated materials have been proposed as a candidate for the first walls-blankets of fusion reactors. One of the main problems connected with using these materials - to minimise the production of long-lived radionuclides from nuclear transmutations and to provide with good technological and functional properties. The selection of materials and their metallurgical and fabrication technologies for fusion reactor components is influenced by this factor. Accurate prediction of induced radioactivity is necessary for the development of the fusion reactor materials. Low activated V-Ti-Cr alloys and reduced activated ferritic-martensitic steels are a leading candidate material for fusion first wall and blanket applications. At the present time a range of compositions and an impurity level are still being investigated to better understand the sensitive of various functional and activation properties to small compositional variations and impurity level. For the two types of materials mentioned above (V-Ti-Cr alloys and 9-12 % Cr f/m steels) and manufactured in Russia (Russia technologies) the analysis of induced activity, hydrogen and helium-production as well as the accumulation of such elements as C, N, O, P, S, Zn and Sn as a function of irradiation time was performed. Materials '' were irradiated '' by fission (BN-600, BOR-60) and fusion (Russian DEMO-C Reactor Project) typical neutron spectra with neutron fluency up to 10 22 n/cm 2 and the cooling time up to 1000 years. The calculations of the transmutation of elements and the induced radioactivity were carried out using the FISPACT inventory code, and the different activation cross-section libraries like the ACDAM, FENDL-2/A and the decay data library FENDL-2/D. It was shown that the level of impurities controls a long-term behaviour of induced activity and contact dose rate for materials. From this analysis the concentration limits of impurities were obtained. The generation of gas

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.

A carbon-carbon composite materials development program for fusion energy applications

International Nuclear Information System (INIS)

Burchell, T.D.; Eatherly, W.P.; Engle, G.B.; Hollenberg, G.W.

1992-10-01

Carbon-carbon composites increasingly are being used for plasma-facing component (PFC) applications in magnetic-confinement plasma-fusion devices. They offer substantial advantages such as enhanced physical and mechanical properties and superior thermal shock resistance compared to the previously favored bulk graphite. Next-generation plasma-fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER) and the Burning Plasma Experiment (BPX), will require advanced carbon-carbon composites possessing extremely high thermal conductivity to manage the anticipated extreme thermal heat loads. This report outlines a program that will facilitate the development of advanced carbon-carbon composites specifically tailored to meet the requirements of ITER and BPX. A strategy for developing the necessary associated design data base is described. Materials property needs, i.e., high thermal conductivity, radiation stability, tritium retention, etc., are assessed and prioritized through a systems analysis of the functional, operational, and component requirements for plasma-facing applications. The current Department of Energy (DOE) Office of Fusion Energy Program on carbon-carbon composites is summarized. Realistic property goals are set based upon our current understanding. The architectures of candidate PFC carbon-carbon composite materials are outlined, and architectural features considered desirable for maximum irradiation stability are described. The European and Japanese carbon-carbon composite development and irradiation programs are described. The Working Group conclusions and recommendations are listed. It is recommended that developmental carbon-carbon composite materials from the commercial sector be procured via request for proposal/request for quotation (RFP/RFQ) as soon as possible

Fusion Programme SCK-CEN - Annual Report 2010

Energy Technology Data Exchange (ETDEWEB)

Massaut, V.

2010-10-15

In 2010 SCK-CEN centred his activities in fusion Research and Development around four main poles: 1) the studies of the first wall of future DEMO facility and plasma wall interactions, using the recently refurbished plasmatron VISIONI; 2) the study of the radiation resistance of optical and specific diagnostic components and the development and prototyping of a Fiber Optics Current Sensor (FOCS) for measuring Tokamak plasma current for long plasma pulses and without embarked electronics: 3) the further study, by irradiation and mechanical testing but also by modelling, of the future structural material for a fusion plant, the RAFM Eurofer and the development and characterization of ODS-Eurofer; 4) specific fusion socio-economic studies on fusion, based on the specific developments carried out for fusion and radioactive waste management. SCK-CEN is also strongly involved in the Broader Approach agreement as Designated Institution for the Belgian State, having a coordinating and managing role for all Belgian activities in this agreement. The present report is structured following the work programme 2010 of the Association. Some former activities form the old-EFDA have been grouped with new ones in coherent and collaborative packages. Most activities of SCK-CEN are, and have always been, carried out under EFDA task agreements.

Fusion Programme SCK-CEN - Annual Report 2010

International Nuclear Information System (INIS)

Massaut, V.

2010-01-01

In 2010 SCK-CEN centred his activities in fusion Research and Development around four main poles: 1) the studies of the first wall of future DEMO facility and plasma wall interactions, using the recently refurbished plasmatron VISIONI; 2) the study of the radiation resistance of optical and specific diagnostic components and the development and prototyping of a Fiber Optics Current Sensor (FOCS) for measuring Tokamak plasma current for long plasma pulses and without embarked electronics: 3) the further study, by irradiation and mechanical testing but also by modelling, of the future structural material for a fusion plant, the RAFM Eurofer and the development and characterization of ODS-Eurofer; 4) specific fusion socio-economic studies on fusion, based on the specific developments carried out for fusion and radioactive waste management. SCK-CEN is also strongly involved in the Broader Approach agreement as Designated Institution for the Belgian State, having a coordinating and managing role for all Belgian activities in this agreement. The present report is structured following the work programme 2010 of the Association. Some former activities form the old-EFDA have been grouped with new ones in coherent and collaborative packages. Most activities of SCK-CEN are, and have always been, carried out under EFDA task agreements.

Irradiation creep at temperatures of 400 degrees C and below for application to near-term fusion devices

International Nuclear Information System (INIS)

Grossbeck, M.L.; Gibson, L.T.; Mansur, L.K.

1996-01-01

To study irradiation creep at 400 degrees C and below, a series of six austenitic stainless steels and two ferritic alloys was irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor; and, after an atomic displacement level of 7.4 dpa, the specimens were moved to the High Flux Isotope Reactor for the remainder of the 19 dpa accumulated. Irradiation temperatures of 60, 200, 330, and 400 degrees C were studied with internally pressurized tubes of type 316 stainless steel, PCA, HT 9, and a series of four laboratory heats of: Fe-13.5Cr-15Ni, Fe-13.5Cr-35Ni, Fe-1 3.5Cr-1 W-0.18Ti, and Fe-16Cr. At 330 degrees C, irradiation creep was shown to be linear in fluence and stress. There was little or no effect of cold-work on creep under these conditions at all temperatures investigated. The HT9 demonstrated a large deviation from linearity at high stress levels, and a minimum in irradiation creep with increasing stress was observed in the Fe-Cr-Ni ternary alloys

Inertial Confinement Fusion Target Component Fabrication and Technology Development report

International Nuclear Information System (INIS)

Steinman, D.

1994-03-01

On December 30, 1990, the US Department of Energy entered into a contract with General Atomics (GA) to be the Inertial Confinement Fusion Target Component Fabrication and Technology Development Support contractor. This report documents the technical activities which took place under this contract during the period of October 1, 1992 through September 30, 1993. During this period, GA was assigned 18 tasks in support of the Inertial Confinement Fusion program and its laboratories. These tasks included ''Capabilities Activation'' and ''Capabilities Demonstration'' to enable us to begin production of glass and composite polymer capsules. Capsule delivery tasks included ''Small Glass Shell Deliveries'' and ''Composite Polymer Capsules'' for Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL). We also were asked to provide direct ''Onsite Support'' at LLNL and LANL. We continued planning for the transfer of ''Micromachining Equipment from Rocky Flats'' and established ''Target Component Micromachining and Electroplating Facilities'' at GA. We fabricated over 1100 films and filters of 11 types for Sandia National Laboratory and provided full-time onsite engineering support for target fabrication and characterization. We initiated development of methods to make targets for the Naval Research Laboratory. We investigated spherical interferometry, built an automated capsule sorter, and developed an apparatus for calorimetric measurement of fuel fill for LLNL. We assisted LANL in the ''Characterization of Opaque b-Layered Targets.'' We developed deuterated and UV-opaque polymers for use by the University of Rochester's Laboratory for Laser Energetics (UR/LLE) and devised a triple-orifice droplet generator to demonstrate the controlled-mass nature of the microencapsulation process

Carbon-carbon composite and copper-composite bond damages for high flux component controlled fusion

International Nuclear Information System (INIS)

Chevet, G.

2010-01-01

Plasma facing components constitute the first wall in contact with plasma in fusion machines such as Tore Supra and ITER. These components have to sustain high heat flux and consequently elevated temperatures. They are made up of an armour material, the carbon-carbon composite, a heat sink structure material, the copper chromium zirconium, and a material, the OFHC copper, which is used as a compliant layer between the carbon-carbon composite and the copper chromium zirconium. Using different materials leads to the apparition of strong residual stresses during manufacturing, because of the thermal expansion mismatch between the materials, and compromises the lasting operation of fusion machines as damage which appeared during manufacturing may propagate. The objective of this study is to understand the damage mechanisms of the carbon-carbon composite and the composite-copper bond under solicitations that plasma facing components may suffer during their life. The mechanical behaviours of carbon-carbon composite and composite-copper bond were studied in order to define the most suitable models to describe these behaviours. With these models, thermomechanical calculations were performed on plasma facing components with the finite element code Cast3M. The manufacturing of the components induces high stresses which damage the carbon-carbon composite and the composite-copper bond. The damage propagates during the cooling down to room temperature and not under heat flux. Alternative geometries for the plasma facing components were studied to reduce damage. The relation between the damage of the carbon-carbon composite and its thermal conductivity was also demonstrated. (author) [fr

Current status of research and development on remote maintenance for fusion components

International Nuclear Information System (INIS)

Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi

2008-01-01

There is a growing attention to remote maintenance of nuclear fusion reactors. Remote maintenance is planned in ITER tokamak to keep the health of in-vessel components like blankets and divertors. In this article, current status of the development in the remote maintenance equipments and methods, especially for ITER tokamak are reviewed. The newly developed vehicle type and boom type maintenance devices, manipulator, and transfer cask are illustrated. (J.P.N.)

The international fusion materials irradiation facility

International Nuclear Information System (INIS)

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

1994-01-01

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

Peaceful fusion

Energy Technology Data Exchange (ETDEWEB)

Englert, Matthias [IANUS, TU Darmstadt (Germany)

2014-07-01

Like other intense neutron sources fusion reactors have in principle a potential to be used for military purposes. Although the use of fissile material is usually not considered when thinking of fusion reactors (except in fusion-fission hybrid concepts) quantitative estimates about the possible production potential of future commercial fusion reactor concepts show that significant amounts of weapon grade fissile materials could be produced even with very limited amounts of source materials. In this talk detailed burnup calculations with VESTA and MCMATH using an MCNP model of the PPCS-A will be presented. We compare different irradiation positions and the isotopic vectors of the plutonium bred in different blankets of the reactor wall with the liquid lead-lithium alloy replaced by uranium. The technical, regulatory and policy challenges to manage the proliferation risks of fusion power will be addressed as well. Some of these challenges would benefit if addressed at an early stage of the research and development process. Hence, research on fusion reactor safeguards should start as early as possible and accompany the current research on experimental fusion reactors.

Characterization for fusion first-wall damage studies of using tailored D-T neutron fields

International Nuclear Information System (INIS)

Dierckx, R.; Emigh, C.R.

1979-01-01

The approximation required to apply the Bullough-Haynes results to the present calculations is somewhat crude and may imply that the details of the results contain considerable error. However, when the results for each neutron source are viewed in a relative context, several valid and important observations can be made. The almost identical swelling results obtained for the intense neutron source (INS) with a standard blanket and the fusion first wall are most striking. A further comparison with a fusion reactor shows that even the spatial and energy distributions of the neutron flux are similar. In both the INS with blanket and at the first wall of a fusion reactor, there is a radial source flux component of 14-MeV neutrons and a more or less isotropic flux component of low energy (< 14-MeV) neutrons. One must therefore conclude that from the point-of-view of neutron radiation damage, the INS with a blanket, unlike all other types of neutron sources, is not a simulation environment. It is, in fact, a small scale fusion device, and data obtained from INS irradiation experiments would represent fusion reactor results. Such data could then be used to develop correlative procedures for applying data obtained from other simulation sources to fusion reactor conditions

European Fusion Programme. ITER task T23: Beryllium characterisation. Progress report. Tensile tests on neutron irradiated and reference beryllium

International Nuclear Information System (INIS)

Moons, F.

1996-02-01

As part of the European Technology Fusion Programme, the irradiation embrittlement characteristics of the more ductile and isotopic grades of beryllium manufactured by Brush Wellman has been investigated using modern powder production and consolidation techniques . This study was initiated in support of the development and evaluation of beryllium as a neutron multiplier for the solid breeder blanket design concepts proposed for a DEMO fusion power reactor. Four different species of beryllium: S-200 F (vacuum hot pressed, 1.2 wt% BeO), S-200FH (hot isostatic pressed, 0.9 wt% BeO), S-65 (vacuum hot pressed, 0.6 wt% BeO), S-65H (hot isostatic pressed, 0.5 wt% BeO) have been compared. Three batches of the beryllium have been investigated, a neutron batch, a thermal control batch and a reference batch. Neutron irradiation has been performed at temperatures between 175 and 605 degrees Celsius up to a neutron fluence of 2.1 10 25 n.m -2 (E> 1 MeV) or 750 appm He. The results of the tensile tests are summarized

Irradiation tests of readout chain components of the ATLAS liquid argon calorimeters

International Nuclear Information System (INIS)

Leroy, C.; Cheplakov, A.; Golikov, V.; Golubykh, S.; Kukhtin, V.; Kulagin, E.; Lushchikov, V.; Minashkin, V.; Shalyugin, A.

2000-01-01

Various readout chain components of the ATLAS liquid argon calorimeters have been exposed to high neutron fluences and γ doses at the irradiation test facility of the IBR-2 reactor of JINR, Dubna. Results of the capacitance and impedance measurements of coaxial cables are presented. Results of peeling tests of PC board samples (carton and copper strips) as a measure of the bonding agent irradiation hardness are also reported

Irradiation tests of readout chain components of the ATLAS liquid argon calorimeters

CERN Document Server

Leroy, C; Golikov, V; Golubyh, S M; Kukhtin, V; Kulagin, E; Luschikov, V; Minashkin, V F; Shalyugin, A N

1999-01-01

Various readout chain components of the ATLAS liquid argon calorimeters have been exposed to high neutron fluences and $gamma$-doses at the irradiation test facility of the IBR-2 reactor of JINR, Dubna. Results of the capacitance and impedance measurements of coaxial cables are presented. Results of peeling tests of PC board samples (kapton and copper strips) as a measure of the bonding agent irradiation hardness are also reported.

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

Vibration of fusion reactor components with magnetic damping

Energy Technology Data Exchange (ETDEWEB)

D’Amico, Gabriele; Portone, Alfredo [Fusion for Energy – Torres Diagonal Litoral B3 – c/Josep Plá n.2, Barcelona (Spain); Rubinacci, Guglielmo [Department of Electrical Eng. and Information Technologies, Università di Napoli Federico II, Via Claudio, 21, 80125 Napoli (Italy); Testoni, Pietro, E-mail: pietro.testoni@f4e.europa.eu [Fusion for Energy – Torres Diagonal Litoral B3 – c/Josep Plá n.2, Barcelona (Spain)

2016-11-01

The aim of this paper is to assess the importance of the magnetic damping in the dynamic response of the main plasma facing components of fusion machines, under the strong Lorentz forces due to Vertical Displacement Events. The additional eddy currents due to the vibration of the conducting structures give rise to volume loads acting as damping forces, a kind of viscous damping, being these additional loads proportional to the vibration speed. This effect could play an important role when assessing, for instance, the inertial loads associated to VV movements in case of VDEs. In this paper, we present the results of a novel numerical formulation, in which the field equations are solved by adopting a very effective fully 3D integral formulation, not limited to the analysis of thin shell structures, as already successfully done in several approaches previously published.

Protein components in saliva and plaque fluid from irradiated primates

Energy Technology Data Exchange (ETDEWEB)

Edgar, W.M.; Bowen, W.H.; Cole, M.F. (Caries Prevention and Research Branch, National Caries Program, NIDR, Bethesda, Maryland, USA)

1982-01-01

Irradiation of the major salivary glands of monkeys (Macaca mulatta) fed cariogenic diets leads to caries clinically indistinguishable from radiation caries in man. This study compares the organic compostion of individual samples of plaque fluid and saliva from irradiated and control monkeys receiving the same cariogenic diet. Plaque and saliva were collected from fasting, tranquillised animals. Four irradiated animals were sampled repeatedly as were non-irradiated controls. Total protein, albumin, immunoglobulins A, G, and M, and the third component of complement (C'3) were quantitated in plaque fluid and whole saliva. Salivary amylase and peroxidase activities were also determined. Plaque fluid and saliva samples were also subjected to polyacrylamide gel electrophoresis. The total viable anaerobic count and numbers of Streptococcus mutans were determined in samples of plaque. The results suggest that the major effect of irradiation leading to increased numbers of S. mutans and caries susceptibility is in the amount, and not the composition, of the saliva produced by the residual gland tissue. The scanty flow of saliva may reduce the effectiveness of cleansing, buffering and lubrication mechanisms as well as resulting in a marked reduction in the total amount of specific and non-specific immune factors entering the mouth.

Protein components in saliva and plaque fluid from irradiated primates

International Nuclear Information System (INIS)

Edgar, W.M.; Bowen, W.H.; Cole, M.F.

1982-01-01

Irradiation of the major salivary glands of monkeys (Macaca mulatta) fed cariogenic diets leads to caries clinically indistinguishable from radiation caries in man. This study compares the organic compostion of individual samples of plaque fluid and saliva from irradiated and control monkeys receiving the same cariogenic diet. Plaque and saliva were collected from fasting, tranquillised animals. Four irradiated animals were sampled repeatedly as were non-irradiated controls. Total protein, albumin, immunoglobulins A, G, and M, and the third component of complement (C'3) were quantitated in plaque fluid and whole saliva. Salivary amylase and peroxidase activities were also determined. Plaque fluid and saliva samples were also subjected to polyacrylamide gel electrophoresis. The total viable anaerobic count and numbers of Streptococcus mutans were determined in samples of plaque. The results suggest that the major effect of irradiation leading to increased numbers of S. mutans and caries susceptibility is in the amount, and not the composition, of the saliva produced by the residual gland tissue. The scanty flow of saliva may reduce the effectiveness of cleansing, buffering and lubrication mechanisms as well as resulting in a marked reduction in the total amount of specific and non-specific immune factors entering the mouth. (author)

Protein components in saliva and plaque fluid from irradiated primates

Energy Technology Data Exchange (ETDEWEB)

Edgar, W M; Bowen, W H; Cole, M F [Caries Prevention and Research Branch, National Caries Program, NIDR, Bethesda, Maryland, USA

1982-01-01

Irradiation of the major salivary glands of monkeys (Macaca mulatta) fed cariogenic diets leads to caries clinically indistinguishable from radiation caries in man. This study compares the organic compostion of individual samples of plaque fluid and saliva from irradiated and control monkeys receiving the same cariogenic diet. Plaque and saliva were collected from fasting, tranquillised animals. Four irradiated animals were sampled repeatedly as were non-irradiated controls. Total protein, albumin, immunoglobulins A, G, and M, and the third component of complement (C'3) were quantitated in plaque fluid and whole saliva. Salivary amylase and peroxidase activities were also determined. Plaque fluid and saliva samples were also subjected to polyacrylamide gel electrophoresis. The total viable anaerobic count and numbers of Streptococcus mutans were determined in samples of plaque. The results suggest that the major effect of irradiation leading to increased numbers of S. mutans and caries susceptibility is in the amount, and not the composition, of the saliva produced by the residual gland tissue. The scanty flow of saliva may reduce the effectiveness of cleansing, buffering and lubrication mechanisms as well as resulting in a marked reduction in the total amount of specific and non-specific immune factors entering the mouth.

Fusion technology development: first wall/blanket system and component testing in existing nuclear facilities

International Nuclear Information System (INIS)

Hsu, P.Y.S.; Bohn, T.S.; Deis, G.A.; Judd, J.L.; Longhurst, G.R.; Miller, L.G.; Millsap, D.A.; Scott, A.J.; Wessol, D.E.

1980-12-01

A novel concept to produce a reasonable simulation of a fusion first wall/blanket test environment employing an existing nuclear facility, the Engineering Test Reactor at the Idaho National Engineering Laboratory, is presented. Preliminary results show that an asymmetric, nuclear test environment with surface and volumetric heating rates similar to those expected in a fusion first wall/blanket or divertor chamber surface appears feasible. The proposed concept takes advantage of nuclear reactions within the annulus of an existing test space (15 cm in diameter and approximately 100 cm high) to provide an energy flux to the surface of a test module. The principal reaction considered involves 3 He in the annulus as follows: n + 3 He → p + t + 0.75 MeV. Bulk heating in the test module is accomplished by neutron thermalization, gamma heating, and absorption reactions involving 6 Li in the blanket breeding region. The concept can be extended to modified core configurations that will accommodate test modules of different sizes and types. It makes possible development testing of first wall/blanket systems and other fusion components on a scale and in ways not otherwise available until actual high-power fusion reactors are built

Fusion for Energy: The European joint undertaking for ITER and the development of fusion energy

International Nuclear Information System (INIS)

Diegele, E.

2009-01-01

Materials development in nuclear fusion for in-vessel components, i.e. for breeder blankets and divertors, has a history of more than two decades. It is the specific in-service and loading conditions and the consequentially required properties in combination with safety standards and social-economic demands that create a unique set of specifications. Objectives of Fusion for Energy (F4E) include: 1) To provide Europe's contribution to the ITER international fusion energy project; 2) To implement the Broader Approach agreement between Euratom and Japan; 3) To prepare for the construction and demonstration of fusion reactors (DEMO). Consequently, activities in F4E focus on structural materials for the first generations of breeder blankets, i.e. ITER Test Blanket Modules (TBM) and DEMO, whereas a Fusion Materials Topical Group implemented under EFDA coordinates R and D on physically based modelling of irradiation effects and R and D in the longer term (new and /or higher risk materials). The paper focuses on martensitic-ferritic steels and (i) reviews briefly the challenges and the rationales for the decisions taken in the past, (ii) analyses the status of the main activities of development and qualification, (iii) indicates unresolved issues, and (iv) outlines future strategies and needs and their implications. Due to the exposure to intense high energy neutron flux, the main issue for breeder materials is high radiation resistance. The First Wall of a breeder blanket should survive 3-5 full power years or, respectively in terms of irradiation damage, typically 50-70 dpa for DEMO and double figures for a power plant. Even though the objective is to have the materials and key fabrication technologies needed for DEMO fully developed and qualified within the next two decades, a major part of the task has to be completed much earlier. Tritium breeding test blanket modules will be installed in ITER with the objective to test DEMO relevant technologies in fusion

Coulomb and nuclear components of the breakup, their interference and effect on the fusion process

International Nuclear Information System (INIS)

Gomes, P R S; Lubian, J; Canto, L F; Otomar, D R; Hussein, M S

2015-01-01

We discuss reaction mechanisms involving weakly bound nuclei, at near barrier energies, and the couplings between different reaction channels. This paper may be thought as a brief description of state of the art of this field, particularly on breakup reactions and their influence on the fusion cross section. Recent experimental and theoretical results are presented, including the interference between Coulomb and nuclear components of the breakup and the systematics so far reached on the static effects due to the characteristic of weakly bound nuclei, especially halo-nuclei and the dynamic effects of the breakup coupling on the fusion cross section. (paper)

Effect of rTMP-GH recombinant fusion protein on thrombocytopoiesis in irradiation injured mice

International Nuclear Information System (INIS)

Xu Yang; Wang Junping; Chen Fang; Shen Mingqiang; Chen Mo; Wang Song; Ran Xinze; Su Yongping; Kai Li

2009-01-01

Objective: To investigate the in vivo effects of rTMP-GH recombinant fusion protein on thrombocytopoiesis in mice with thrombopenia inflicted by irradiation. Methods: BALB/C mice weighting around 20 g were irradiated with 5 Gy of 60 Co γ-ray irradiation to generate thrombopenia. The irradiation injured mice were injected with rTMP-GH or rhGH subcutaneously at the dose of 200 (μg ·kg -1 · d -1 for 7 days. From the 6 th day, the platelets in blood samples from vena caudalis were counted routinely, and the pathological changes of bone marrow were determined by morphological observation. Results: From the 10 th day, the levels of blood platelet in rTMP-GH treated mice were much higher than those of rhGH treatment group and normal saline (NS) control group, especially at the nadir (P < 0.01). On the 22 nd day, the platelet count has recovered up to 80% of normal level in rTMP-GH treatment group, while it has just recovered up to 30% in NS control group. Morphological observation showed that there was obvious reconstruction of bone marrow in mice treated with rTMP-GH, compared with NS group.The number of megarkaryoblasts and megakaryocytes in bone marrow of rTMP-GH treated mice (3.07 ± 0.32) was much higher than those of rhGH treatment group (2.20 ± 0.22, P < 0.05) and NS control group (0.87 ± 0.19, P <0.01). Conclusions: rTMP-GH has potent effects on the recovery of blood platelet by promoting megarkaryocytopoiesis in irradiation injuried mice. (authors)

Managing fusion high-level waste-A strategy for burning the long-lived products in fusion devices

International Nuclear Information System (INIS)

El-Guebaly, L.A.

2006-01-01

Fusion devices appear to be a viable option for burning their own high-level waste (HLW). We propose a novel strategy to eliminate (or minimize) the HLW generated by fusion systems. The main source of the fusion HLW includes the structural and recycled materials, refractory metals, and liquid breeders. The basic idea involves recycling and reprocessing the waste, separating the long-lived radionuclides from the bulk low-level waste, and irradiating the limited amount of HLW in a specially designed module to transmute the long-lived products into short-lived radioisotopes or preferably, stable elements. The potential performance of the new concept seems promising. Our analysis indicated moderate to excellent transmutation rates could be achieved in advanced fusion designs. Successive irradiation should burn the majority of the HLW. The figures of merit for the concept relate to the HLW burn-up fraction, neutron economy, and impact on tritium breeding. Hopefully, the added design requirements could be accommodated easily in fusion power plants and the cost of the proposed system would be much less than disposal in a deep geological HLW repository. Overall, this innovative approach offers benefits to fusion systems and helps earn public acceptance for fusion as a HLW-free source of clean nuclear energy

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

Energy Technology Data Exchange (ETDEWEB)

Esposito, B. E-mail: esposito@frascati.enea.it; Bertalot, L.; Maruccia, G.; Petrizzi, L.; Bignan, G.; Blandin, C.; Chauffriat, S.; Lebrun, A.; Recroix, H.; Trapp, J.P.; Kaschuck, Y

2000-11-01

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. The analysis has been carried out using the LSL-M2 code, which optimizes the neutron spectrum by means of a least-squares technique taking into account the variance and covariance files. In the second part of the activity, the possibility of extending to IFMIF the use of existing on-line in-core neutron/gamma monitors (to be located at several positions inside the IFMIF test cell for beam control, safety and diagnostic purposes) has been studied. A feasibility analysis of the modifications required to adapt sub-miniature fission chambers (recently developed by CEA-Cadarache) to the high flux test module of the test cell has been carried out. The verification of this application pertinence and a gross definition of the in-core detector characteristics are described. The option of using self-powered neutron detectors (SPNDs) is also discussed.

Evaluation of microstructure of irradiated fuel channel components of PHWR

International Nuclear Information System (INIS)

Ramadasan, E.

2005-01-01

Performance evaluation and failure analysis of irradiated reactor structural components such as those in-core and PHT circuit components necessitate metallographic evaluation using special metallographic specimen preparation techniques due to the radiation dose and contamination levels involved in handling the specimens. The metallographic specimen preparation techniques that are resorted to involve use of fully automatic and semi automatic machines, shielded metallographic microscope and specialised equipment developed for lead-cell metallography. The techniques used and the results obtained in the metallographic studies on irradiated fuel channel components such as pressure tubes and garter springs of various Indian PHWRs at RAPS, NAPS and MAPS are presented as case studies in the paper. The evaluation of oxidation and hydriding behaviour of zircaloy-2 pressure tubes and garter springs are presented. The paper also gives in detail the microstructural evaluation of hydride blistering seen at the PT-CT contact location of the pressure tubes of RAPS-2. The evaluation revealed that the hydride blisters was small compared to their length, unlike the hydride blisters seen in the CANDU pressure tube G-16 of Pickering-2. This could be attributed to be due to the difference in the annulus conditions between the two types of reactors. The hydride blisters in J-07 pressure tube of RAPS-2 had ductile material adjacent to them. The paper also gives the hydride blistering observation on irradiated Zr-2.5% Nb-0.5% Cu garter springs of RAPS-2. It was seen that there was only negligible hydriding of the garter springs during service through they showed presence of benign hydride blisters in them. The general hydriding observations made on the pressure tubes of Indian PHWRs under different conditions are also presented. (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

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

Sprout inhibition and change in organic components of potato by gamma-irradiation

International Nuclear Information System (INIS)

Rahman, M.S.; Kume, Tamikazu; Ishigaki, Isao.

1989-12-01

Radiation technology for sprout inhibition and change in organic components of potato by irradiation were investigated. Dose distribution in the package filled with potatoes (depth 45cm, density: 0.56g/cm 3 ) was measured using Fricke dosimeter. When the package was irradiated at dose rate of 5 x 10 5 , 1 x 10 5 and 5 x 10 4 rad/hr, the dose uniformities were calculated as 1.79, 1,45 and 1.35, and the relative throughput capacities were 1.0, 0.26 and 0.14, respectively. After 7 months storage, the sprout of potatoes was not observed at 10 krad irradiation while 57% of potatoes was sprouted at 5 krad. The contents of oxalic and malic acids were slightly increased by irradiation up to 100 krad while that of citric and succinic acids were not changed. The change in contents of these organic acids during storage was almost the same in both unirradiated and irradiated samples. Sucrose content was reached maximum after 8 days in 15 krad irradiated sample while it was increased through 40 days storage in 300 krad irradiated sample. The increase in sucrose contents by irradiation was higher in cortical tissue than in medullary tissue. (author)

Study on a multi-component palladium alloy membrane for the fusion fuel cycle

International Nuclear Information System (INIS)

Yoshida, Hiroshi; Okuno, Kenji; Nagasaki, Takanori; Noda, Kenji; Ishii, Yoshinobu; Takeshita, Hidefumi.

1985-11-01

A feasibility study on the material integrity with respect to the hydride formation and helium damage of the palladium alloy membrane was performed for an application of the palladium diffuser to a fusion fuel cleanup process. This study was conducted under the Japan/US Fusion Cooperation Program. Experimental works on the crystallography, hydrogen solubility and 3 He release characteristics were carried out with a multi-component palladium alloy(Pd-25Ag.Au.Ru). The excellent hydrogen permeability and mechanical properties of the membrane made of this alloy had been confirmed by authors' previous study. Based on the present study, this alloy membrane has high resistivity to the hydrogen embrittlement, and swelling and fracture due to the helium bubble formation under the practical operating conditions of the diffuser. (author)

Analysis of Gamma-irradiated Soybean Components by Electron Paramagnetic Resonance

International Nuclear Information System (INIS)

Oliveira, M.R. R. de; Quadrado, M.G.O.; Mastro, N.L. del

2007-01-01

Soybean (Glycine max) seeds contain besides oil and protein, important phytochemicals that have been shown in recent years to offer important health benefits. Soybean contains at least six classes of antioxidant compounds: flavonol, isoflavones, anthocyanins, proanthocyanidins, tocopherols, and poly carboxylic acids. An increasing number of studies have documented the significant value of many classes of these compounds, mainly isoflavones, not only as potent antioxidants, but also as antitumor agents and cardio protective compounds. Food irradiation is gaining increasing attention around the world but it is not a worldwide approved treatment yet. Electron paramagnetic resonance, EPR, is considered the most important technique to detect free-radicals on food. Results from a previous work showed that irradiated soybean could be detected by EPR only when higher doses were employed. This study was undertaken to investigate the radiation response of the diverse parts of the soy seed: hull or seed coat, cotyledons, hilum and hypocotyl axis or germ, from different soybean cultivars. Soybean samples were obtained from the National Soybean Research Center (Embrapa-Soja), Londrina, Brazil, separated in their components and gamma-irradiated in a Gamma cell 220 (AECL) with doses of 0.1 and 2.0 kGy at a dose rate of 2.9 kGy/h. EPR measurements were performed on an X-band spectrometer (ER 041 XG Microwave Bridge, Bruker). Both irradiation and EPR measurements were performed at room temperature (20-25 C). The results showed that the EPR signal intensity correlated with the ionizing radiation dose, although different cultivars presented differences in their radiation response. The main EPR peak corresponding to free radical presented differences in shape and intensity. The hull and the hilum presented signals higher and easier to be analyzed than the whole bean, indicating strong differences in radiation sensitivity of soybean components. (Author)

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

Fusion research in Hungary

International Nuclear Information System (INIS)

Zoletnik, S.

2004-01-01

Hungarian fusion research started in the 1970s, when the idea of installing a small tokamak experiment emerged. In return to computer equipment a soviet tokamak was indeed sent to Hungary and started to operate as MT-1 at the Central Research Institute for Physics (KFKI) in 1979. Major research topics included diagnostic development, edge plasma studies and investigation of disruptions. Following a major upgrade in 1992 (new vacuum vessel, active position control and PC network based data acquisition system) the MT-1M tokamak was used for the study of transport processes with trace impurity injection, micropellet ablation studies, X-ray tomography and laser blow-off diagnostic development. Although funding ceased in the middle of the 90's the group was held alive by collaborations with EU fusion labs: FZ -Juelich, IPP-Garching and CRPP-EPFL Lausanne. In 1998 the machine was dismantled due to reorganization of the Hungarian Academy of Sciences. New horizons opened to fusion research from 1999, when Hungary joined EURATOM and a fusion Association was formed. Since then fusion physics studies are done in collaboration with major EU fusion laboratories, Hungarian researchers also play an active role in JET diagnostics upgrade and ITER design. Major topics are pellet ablation studies, plasma turbulence diagnosis using Beam Emission Spectroscopy and other techniques, tomography and plasma diagnostics using various neutral beams. In fusion relevant technology R and D Hungary has less records. Before joining EURATOM some materials irradiation studies were done at the Budapest Research Reactor at KFKI-AEKI. The present day fusion technology programme focuses still on irradiation studies, nuclear material database and electromagnetic testing techniques. Increasing the fusion technology research activities is a difficult task, as the competition in Hungarian industry is very strong and the interest of organizations in long-term investments into R and D is rather weak and

Results of preliminary experiments on tritium decontamination by UV irradiation

International Nuclear Information System (INIS)

Oya, Yasuhisa; Shu, Wataru; O'hira, Shigeru; Hayashi, Takumi; Nishi, Masataka

2000-03-01

In the point of view of protection of workers from the radiation exposure and the limitation of the contamination with radioactive materials, it is important to decontaminate mobile tritium from plasma facing components of a nuclear fusion reactor at the beginning of their maintenance work. It is considered that the heating is the most effective method for decontamination. However, it is important to develop new decontamination method of adsorbed hydro-carbon based substances from the materials that cannot be heated or the inner pipe of double pipes. This report presents results of preliminary experiments performed for the development of the effective tritium decontamination technique pursuing under US/Japan collaborative program on technology for fusion-fuel processing (Annex IV). In the experiments, the effects of Ultra Violet (UV) irradiation on tritium removal from some kinds of materials, such as poly vinyl chloride -(CH 2 CHCl) n - film, polyethylene film and graphite samples coated by C 2 H 2 plasma were examined. As the result of UV irradiation, it was confined that hydrogen and carbon based compounds could be released from the specimen during UV irradiation. It is concluded that UV irradiation is one of the hopeful candidates for effective tritium decontamination. (author)

Evaluation of the Delta-T SPN1 radiometer for the measurement of solar irradiance components

Science.gov (United States)

Estelles, Victor; Serrano, David; Segura, Sara; Wood, John; Webb, Nick; Utrillas, Maria Pilar

2016-04-01

In this study we analyse the performance of an SPN1 radiometer built by Delta-T Devices Ltd. to retrieve global solar irradiance at ground and its components (diffuse, direct) in comparison with measurements from two Kipp&Zonen CMP21 radiometers and a Kipp&Zonen CHP1 pirheliometer, mounted on an active Solys-2 suntracker at the Burjassot site (Valencia, Spain) using data acquired every minute during years 2013 - 2015. The measurement site is close to sea level (60 m a.s.l.), near the Mediterranean coast (10 km) and within the metropolitan area of Valencia City (over 1.500.000 inhabitants). The SPN1 is an inexpensive and versatile instrument for the measurement of the three components of the solar radiation without any mobile part and without any need to azimuthally align the instrument to track the sun (http://www.delta-t.co.uk). The three components of the solar radiation are estimated from a combination of measurements performed by 7 different miniature thermopiles. The SPN1 pyranometer measures the irradiance between 400 and 2700 nm, and the nominal uncertainty for the individual readings is about 8% ± 10 W/m2 (5% for the daily averages). The pyranometer Kipp&Zonen CMP21 model is a secondary standard for the measurement of broadband solar global irradiance in horizontal planes. Two ventilated CMP21 are used for the measurement of the global and diffuse irradiances. The expected total daily uncertainty of the radiometer is estimated to be 2%. The pirheliometer Kipp&Zonen CHP1 is designed for the measurement of the direct irradiance. The principles are similar to the CMP21 pyranometer. The results of the comparison show that the global irradiance from the SPN1 compares very well with the CMP21, with absolute RMSD and MBD differences below the combined uncertainties (15 W/m2 and -5.4 W/m2, respectively; relative RMSD of 3.1%). Both datasets are very well correlated, with a correlation coefficient higher than 0.997 and a slope and intercept very close to 1 and 0

Neutron irradiation damage in transition metal carbides

International Nuclear Information System (INIS)

Matsui, Hisayuki; Nesaki, Kouji; Kiritani, Michio

1991-01-01

Effects of neutron irradiation on the physical properties of light transition metal carbides, TiC x , VC x and NbC x , were examined, emphasizing the characterization of irradiation induced defects in the nonstoichiometric composition. TiC x irradiated with 14 MeV (fusion) neutrons showed higher damage rates with increasing C/Ti (x) ratio. A brief discussion is made on 'cascade damage' in TiC x irradiated with fusion neutrons. Two other carbides (VC x and NbC x ) were irradiated with fission reactor neutrons. The irradiation effects on VC x were not so simple, because of the complex irradiation behavior of 'ordered' phases. For instance, complete disordering was revealed in an ordered phase, 'V 8 C 7 ', after an irradiation dose of 10 25 n/m 2 . (orig.)

An advanced fusion neutron source facility

International Nuclear Information System (INIS)

Smith, D.L.

1992-01-01

Accelerator-based 14-MeV-neutron sources based on modifications of the original Fusion Materials Irradiation Facility are currently under consideration for investigating the effects of high-fluence high-energy neutron irradiation on fusion-reactor materials. One such concept for a D-Li neutron source is based on recent advances in accelerator technology associated with the Continuous Wave Deuterium Demonstrator accelerator under construction at Argonne National Laboratory, associated superconducting technology, and advances in liquid-metal technology. In this paper a summary of conceptual design aspects based on improvements in technologies is presented

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

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

International Nuclear Information System (INIS)

2013-11-01

of materials under development for Generation IV concepts. International collaboration among MTRs and specialized facilities has been identified as integral to progress in fusion development as well as enhancing reactor utilization. This publication specifies which areas of research remain in the qualification of structural materials and components, and has detailed the characteristics of many research reactors and devices that can accomplish an important portion of these necessary studies. This publication is the outcome of two recent IAEA sponsored meetings under its programme to enhance the utilization and collaboration of research reactor and material test facilities: - Consultancy meeting on Role of Research Reactors in Materials Research for Nuclear Fusion Technology, 13-15 December 2010, IAEA, Vienna; - Technical meeting on Materials under High Energy and High Intensity Neutron Fluxes for Nuclear Fusion Technology, 27-29 June 2011, IAEA, Vienna. These meetings brought together representatives from MTRs, spallation neutron sources, multiple beam irradiation facilities, material scientists as well as fusion community representatives to discuss the current state of fusion research and to plot necessary studies and modes of research collaboration

Nuclear data requirements for fusion reactor nucleonics

International Nuclear Information System (INIS)

Bhat, M.R.; Abdou, M.A.

1980-01-01

Nuclear data requirements for fusion reactor nucleonics are reviewed and the present status of data are assessed. The discussion is divided into broad categories dealing with data for Fusion Materials Irradiation Test Facility (FMIT), D-T Fusion Reactors, Alternate Fuel Cycles and the Evaluated Data Files that are available or would be available in the near future

Effects of irradiation and drying on volatile components of fresh shiitake (Lentinus edodes Sing)

International Nuclear Information System (INIS)

Yang, Ming-Sheng; Chyau, Charng-Cherng; Horng, Deng-Tsen; Yang, Jui-Sen

1998-01-01

Fresh shiitake (Lentinus edodes Sing) was irradiated with doses of 0.5, 1 and 2 kGy using 60 Co. Effects of γ-irradiation and drying on the volatile composition of shiitake were studied by gas chromatography (GC) and GC-mass spectrometry. Irradiation above 1.0 kGy could inhibit the growth and mould decay of fresh shiitakes after harvesting. Irradiation with 2 kGy increased the eight-carbon volatile components of fresh shiitake. Treatment at 1 kGy irradiation of fresh shiitake produced some new volatile compounds in the dry product, such as methylethyl disulphide, sulphinylbis methane, methyl(methylthio)ethyl disulphide and N-(3-methylbutyl) acetamide. The eight-carbon compounds mostly disappeared after drying. The amount of sulphur-containing volatile compounds in dried shiitake became lower during irradiation. Irradiation with doses of 1 or 2 kGy of fresh shiitake did not increase the volatile content of shiitake after drying

Analysis of the influence of external irradiation component on the patients with thyroid cancer affected by the Chernobyl nuclear power plant accident

International Nuclear Information System (INIS)

Tepla, O.V.; Kovalenko, O.M.

2006-01-01

The definition possible relationship between the latent period and doses of external irradiation component on the thyroid gland in patients was estimated. Dose reconstruction from external irradiation component on the thyroid gland was applied in 99 patients with thyroid cancer affected by the Chernobyl accident. External irradiation component does not always correspond to the range of the doses that increase risk of thyroid cancer. No linear relation between the latent period duration and the dose of the external irradiation component on thyroid was revealed

Pacific Northwest Laboratory report on fusion energy research, April 1977 - June 1977

Energy Technology Data Exchange (ETDEWEB)

None

1977-07-01

The development of economic data for fusion power plants continued in a study estimating the potential impact of a shortage of materials important in fusion plant construction. In studies developing heat transfer and fluid flow design tools for fusion reactor blankets, preconceptual design studies were initiated to identify the potential design limits of water cooling in the first wall of Tokamak Next Step (TNS) concepts. In surface effects research clean gold samples were irradiated in the University of California (D,Be) neutron source for a neutron sputtering experiment. Light ion and neutron irradiation experiments have continued in studies of the effects of radiation on mechanical properties. The hardening response of 14 MeV neutron-irradiated nickel changed at high particle fluences (10/sup 16/ to 10/sup 17/ particles/cm/sup 2/) while the hardening response of 16 MeV proton-irradiated nickel did not, which may have been due to a difference in irradiation hardening mechanisms. The flux dependence of the damage microstructure and irradiation hardening of materials needs further study to clarify uncertainty about light ion and fusion neutron damage processes. Neutron irradiations of Ni, 316SS, and Nb wires and foils were completed. Work has continued in studies developing acoustic emission (AE) techniques for determining the prebreakdown behavior and failure mechanisms in electric insulators with potential applications in fusion reactors. Scoping experiments with the high-vacuum dielectric breakdown apparatus were conducted.

EU Development of High Heat Flux Components

International Nuclear Information System (INIS)

Linke, J.; Lorenzetto, P.; Majerus, P.; Merola, M.; Pitzer, D.; Roedig, M.

2005-01-01

The development of plasma facing components for next step fusion devices in Europe is strongly focused to ITER. Here a wide spectrum of different design options for the divertor target and the first wall have been investigated with tungsten, CFC, and beryllium armor. Electron beam simulation experiments have been used to determine the performance of high heat flux components under ITER specific thermal loads. Beside thermal fatigue loads with power density levels up to 20 MWm -2 , off-normal events are a serious concern for the lifetime of plasma facing components. These phenomena are expected to occur on a time scale of a few milliseconds (plasma disruptions) or several hundred milliseconds (vertical displacement events) and have been identified as a major source for the production of neutron activated metallic or tritium enriched carbon dust which is of serious importance from a safety point of view.The irradiation induced material degradation is another critical concern for future D-T-burning fusion devices. In ITER the integrated neutron fluence to the first wall and the divertor armour will remain in the order of 1 dpa and 0.7 dpa, respectively. This value is low compared to future commercial fusion reactors; nevertheless, a nonnegligible degradation of the materials has been detected, both for mechanical and thermal properties, in particular for the thermal conductivity of carbon based materials. Beside the degradation of individual material properties, the high heat flux performance of actively cooled plasma facing components has been investigated under ITER specific thermal and neutron loads

High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

Science.gov (United States)

Neu, R.

2006-04-01

In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials (PFMs). However, in a future fusion reactor, for economic reasons, a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which have to be fulfilled simultaneously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFMs but also point to operational restrictions, when using them as PFCs. These are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor-relevant scenarios which incorporate the experiences and measures.

High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

International Nuclear Information System (INIS)

Neu, R.

2006-01-01

In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials (PFMs). However, in a future fusion reactor, for economic reasons, a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which have to be fulfilled simultaneously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFMs but also point to operational restrictions, when using them as PFCs. These are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor-relevant scenarios which incorporate the experiences and measures

Irradiation test of component for radiation-resistant small sized motor

International Nuclear Information System (INIS)

Nakamichi, M.; Ishitsuka, E.; Shimakawa, S.; Kan, S.

2009-01-01

A small-sized motor with a resistance to radiation was developed. This motor has been able to operate at a gamma-ray dose of a value 700 times as high as the specification of a commercial motor. The present work describes results of post-irradiation examinations (PIEs) to evaluate effects of neutron irradiation on the lifetime of some major components of the motor such as a bearing, a magnet and a fixation agent for a field coil wire. It became clear from the results of PIEs that the radiation-resistance dose of the motor using a Sm-Co magnet will be expected to be one order of magnitude higher than that of the motor using a Nb-Fe-B magnet.

The fusion blanket program at Chalk River

International Nuclear Information System (INIS)

Hastings, I.J.

1986-03-01

Work on the Fusion Blanket Program commenced at Chalk River in 1984 June. Co-funded by Canadian Fusion Fuels Technology Project and Atomic Energy of Canada Limited, the Program utilizes Chalk River expertise in instrumented irradiation testing, ceramics, tritium technology, materials testing and compound chemistry. This paper gives highlights of studies to date on lithium-based ceramics, leading contenders for the fusion blanket

High quality actively cooled plasma facing components for fusion

International Nuclear Information System (INIS)

Nygren, R.

1993-01-01

This paper interweaves some suggestions for developing actively-cooled PFCs (plasma facing components) for future fusion devices with supporting examples taken from the design, fabrication and operation of Tore Supra's Phase III Outboard Pump Limiter (OPL). This actively-cooled midplane limiter, designed for heat and particle removal during long pulse operation, has been operated in essentially thermally steady state conditions. From experience with testing to identify braze flaws in the OPL, recommendations are made to analyze the impact of joining flaws on thermal-hydraulic performance of PFCs and to validate a method of inspection for such flaws early in the design development. Capability for extensive in-service monitoring of future PFCs is also recommended and the extensive calorimetry and IR thermography used to confirm and update safe operating limits for power handling of the OPL are reviewed

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

Fusion energy studies

International Nuclear Information System (INIS)

Anon.

1978-01-01

The following topics are considered: (1) cryosorption vacuum pumping for fusion reactors, (2) TNS support studies, (3) tritium recovery from irradiated Li-Al and SAP, (4) actinide oxides, nitrides, and carbides, and (5) transition metal-actinide-C phase equilibria

Progress of laser nuclear fusion research

International Nuclear Information System (INIS)

Shiraga, Hiroyuki

2017-01-01

This paper describes the principle and features of nuclear fusion using laser, as well as its basic concepts such as high-temperature / high-density implosion system and fast ignition of fuel. At present, researches aiming at nuclear fusion ignition have been developing. As the current state of researches, this paper reviews the situations of FIREX (Fast Ignition Realization Experiment) project of Japan focusing on direct irradiation implosion and fast ignition system, as well as NIF (National Ignition Facility) project of the U.S. aiming at ignition combustion based on indirect irradiation implosion and central ignition system. In collaboration with the National Institute for Fusion Science, Osaka University started FIREX-1 project in 2003. It built a heating laser LFEX of 10 kJ/1 to 10ps, and started an implosion/heating integration experiment in 2009. Currently, it is developing experiment to achieve heating to 5 keV. At NIF, the self-heating of central sparks via energy of α particles generated in the nuclear fusion reaction has been realized. This paper also overviews R and D issues surrounding the lasers for reactors for use in laser nuclear fusion power generators. (A.O.)

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

Deuterium trapping at vacancy clusters in electron/neutron-irradiated tungsten studied by positron annihilation spectroscopy

Science.gov (United States)

Toyama, T.; Ami, K.; Inoue, K.; Nagai, Y.; Sato, K.; Xu, Q.; Hatano, Y.

2018-02-01

Deuterium trapping at irradiation-induced defects in tungsten, a candidate material for plasma facing components in fusion reactors, was revealed by positron annihilation spectroscopy. Pure tungsten was electron-irradiated (8.5 MeV at ∼373 K and to a dose of ∼1 × 10-3 dpa) or neutron-irradiated (at 573 K to a dose of ∼0.3 dpa), followed by post-irradiation annealing at 573 K for 100 h in deuterium gas of ∼0.1 MPa. In both cases of electron- or neutron-irradiation, vacancy clusters were found by positron lifetime measurements. In addition, positron annihilation with deuterium electrons was demonstrated by coincidence Doppler broadening measurements, directly indicating deuterium trapping at vacancy-type defects. This is expected to cause significant increase in deuterium retention in irradiated-tungsten.

Implication of irradiation effects on materials data for the design of near core components

International Nuclear Information System (INIS)

Dietz, W.; Breitling, H.

1995-01-01

For LWR's strict regulations exist for the consideration of irradiation in the design and surveillance of the reactor pressure vessel in the various codes (ASME, RCC-M, KTA) but less for near core components. For FBR's no firm rules exist either for the vessel nor the reactor internals. In this paper the German design practices for the loop type SNR-300 will be presented, and also some information from the surveillance programme of the KNK-reactor. Austenitic stainless steels have been mainly selected for the near core components. For some special applications Ni-alloys and a stabilized 2 1/4 Cr 1 Mo-alloy were specified. Considerations of the irradiation effects on material properties will be made for the various temperature and fluence levels around the core. The surveillance programmes will be described. Both, the consideration of irradiation effects in the elastic and inelastic analysis and the surveillance programmes had been a part of the licensing process for SNR-300. (author). 8 figs, 4 tabs

Effects of cryogenic reactor irradiation on organic insulators

International Nuclear Information System (INIS)

Kato, Teruo

1995-01-01

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

GfW-handbook for irradiation test guidelines for radiation hardness of electronic components

International Nuclear Information System (INIS)

Braeunig, D.; Wulf, F.; Gaebler, W.; Boden, A.

1982-12-01

The purpose of the report is to propose irradiation test methods so that a standardized application of the methods can lead to a better comparison of test results. The interaction of different radiation species with matter - ionization and displacement - is described. Application of appropriate radiation sources, dosimetry problems, and shielding for simulating space radiation effects by laboratory testing is discussed. The description and characteristics of the irradiation sources are presented. Flowcharts of the planning and running of irradiation tests are given. Guidelines for running the tests are established, test methods and test circuits are proposed. The test system offers the capability of measuring devices also of high complexity up to microprocessors. The test results are collected regularly and are published in GfW-Handbook TN53/08, 'Data Compilation of Irradiation Tested Electronic Components'. (orig./HP) [de

Comparative study of the tungsten irradiation conditions in IFMIF and DEMO

Energy Technology Data Exchange (ETDEWEB)

Simakov, S.P.; Pereslavtsev, P.; Fischer, U. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology; Moeslang, A. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany). Inst. for Material Research I

2010-05-15

The International Fusion Material Irradiation Facility (IFMIF) [1] will provide an accelerator based intense neutron source with a white spectrum extending up to 55 MeV for high fluence irradiations of fusion power reactor (FPR) candidate materials. Material samples located in test modules will be subjected to a radiation load anticipated for a fusion power reactor. The highest neutron flux is expected in the High Flux Test Module, which is considered in the IFMIF design to host around 1000 compactly packed stainless steel samples - the main structure materials of power fusion reactors. Another material subjected to the highest loads in a FPR is a tungsten. It is planned to be used as armour tiles for the divertor or the first wall. It turned out that no specific effort has been undertaken so far to search for a suitable irradiation location in the IFMIF Test Cell which provides a reasonable representation of the irradiation conditions in the divertor of a fusion power reactors. (orig.)

Development of bonding techniques of W and Cu-alloys for plasma facing components of fusion reactor with HIP method

International Nuclear Information System (INIS)

Saito, S.; Fukaya, K.; Ishiyama, S.; Eto, M.; Sato, K.; Akiba, M.

1998-01-01

W (tungsten) and Cu (copper)-alloys, like oxygen free high thermal conductivity (OFHC)-copper or dispersion strengthened (DS)-copper, are candidate materials for plasma facing components(PFC) of TOKAMAK type fusion reactor as armor tile and heat sink, respectively. However, PFC are exposed to cyclic high heat load and heavy irradiation by 14 MeV neutrons. Under these conditions, thermal stresses at bonding interface and irradiation damage will decrease the bonding strength between W and Cu alloys. Therefore, it is necessary to develop a reliable bonding techniques in order to make PFC with enough integrity. We have applied the hot isostatic press (HIP) method to bond W with Cu-alloys. In this experiments, to optimize HIP bonding conditions, four point bending tests were performed for different bonding conditions at temperatures from R.T. to 873 K and we obtained an optimum HIP bonding condition for W and OFHC-Cu as 1273 SK x 2 hours x 98 ∼ 147 MPa. Tensile tests were also performed at the same temperature range. The tensile strength of the bonded W / Cu was almost equal to that of OFHC Cu which was HIPed at the same conditions. Tensile specimens were broken at the bonding interface or OFHC-Cu side. Bonding tests of W and DS-Cu showed that HIP was not successful because tungsten oxide was produced at the bonding interface and residual stresses were not relaxed. Therefore, it was concluded that some insert materials will be needed to bond W and DS-Cu. (author)

A Review of Irradiation Effects on Organic-Matrix Insulation

International Nuclear Information System (INIS)

Simon, N.J.

1993-01-01

This review assesses the data base on epoxy and polyimide matrix insulation to determine whether organic electric insulation systems can be used in the toroidal field (TF) magnets of next generation fusion devices such as ITER* and TPX*. Owing to the difficulties of testing insulation under fusion reactor conditions, there is a considerable mismatch between the ITER requirements and the data that are currently available. For example, nearly all of the high-dose (5 x 10 7 to 10 8 Gy) data obtained on epoxy and polyimide matrix insulation employed gamma irradiation, electron irradiation, or reactor irradiation with a fast neutron fluence far below 10 23 /m 2 , the fluence expected for the insulation at the TF magnets, as set forth in ITER conceptual design documents. Also, the neutron spectrum did not contain a very high energy (E (ge) 5 MeV) component. Such data underestimate the actual damage that would be obtained with the neutron fluence and spectrum expected at a TF magnet. Experiments on a polyimide (Kapton) indicate that gamma or electron doses or mixed gamma and neutron reactor doses would have to be downgraded by a factor of up to ten to simulate fusion neutron doses. Even when neutrons did constitute a significant portion of the total dose, B-containing E-glass reinforcement was often used; therefore, excess damage from the 10 B + n → 7 Li + α reaction occurred near the glass-epoxy interface. This problem can easily be avoided by substituting B-free glass (R, S, or T types)

Progress on using deuteron-deuteron fusion generated neutrons for 40Ar/39Ar sample irradiation

Science.gov (United States)

Rutte, Daniel; Renne, Paul R.; Becker, Tim; Waltz, Cory; Ayllon Unzueta, Mauricio; Zimmerman, Susan; Hidy, Alan; Finkel, Robert; Bauer, Joseph D.; Bernstein, Lee; van Bibber, Karl

2017-04-01

We present progress on the development and proof of concept of a deuteron-deuteron fusion based neutron generator for 40Ar/39Ar sample irradiation. Irradiation with deuteron-deuteron fusion neutrons is anticipated to reduce Ar recoil and Ar production from interfering reactions. This will allow dating of smaller grains and increase accuracy and precision of the method. The instrument currently achieves neutron fluxes of ˜9×107 cm-2s-1 as determined by irradiation of indium foils and use of the activation reaction 115In(n,n')115mIn. Multiple foils and simulations were used to determine flux gradients in the sample chamber. A first experiment quantifying the loss of 39Ar is underway and will likely be available at the time of the presentation of this abstract. In ancillary experiments via irradiation of K salts and subsequent mass spectrometric analysis we determined the cross-sections of the 39K(n,p)39Ar reaction at ˜2.8 MeV to be 160 ± 35 mb (1σ). This result is in good agreement with bracketing cross-section data of ˜96 mb at ˜2.45 MeV and ˜270 mb at ˜4 MeV [Johnson et al., 1967; Dixon and Aitken, 1961 and Bass et al. 1964]. Our data disfavor a much lower value of ˜45 mb at 2.59 MeV [Lindström & Neuer, 1958]. In another ancillary experiment the cross section for 39K(n,α)36Cl at ˜2.8 MeV was determined as 11.7 ± 0.5 mb (1σ), which is significant for 40Ar/39Ar geochronology due to subsequent decay to 36Ar as well as for the determination of production rates of cosmogenic 36Cl. Additional experiments resolving the cross section functions on 39K between 1.5 and 3.6 MeV are on their way using the LICORNE neutron source of the IPN Orsay tandem accelerator. Results will likely be available at the time of the presentation of this abstract. While the neutron generator is designed for fluxes of ˜109 cm-2s-1, arcing in the sample chamber currently limits the power—straightforwardly correlated to the neutron flux—the generator can safely be run at. Further

The computerised accountancy system (MYDAS) for irradiated components in RNL's Mayfair Laboratory at Culcheth

International Nuclear Information System (INIS)

Stansfield, R.G.; Baker, A.R.

1985-09-01

The computerised Mayfair Accountancy System (MYDAS) has been developed to account for irradiated components in the Mayfair Laboratory at Culcheth and supersedes a card-index system. The computerised system greatly improves the availability of the data held and it ensures, by means of extensive data validation programs, that the data accurately represent the current inventory of irradiated components in the Laboratory. The system has been implemented on the Risley ICL 2966 main-frame computer and uses an IDMS database to store the data. The computer is accessed through the facilities of the Transaction Processing Management System (TPMS) providing rapid and secure access to the database from several visual display units and printers simultaneously. (author)

Recent developments concerning the fusion; Developpements recents sur la fusion

Energy Technology Data Exchange (ETDEWEB)

Jacquinot, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint Paul lez Durance (France); Andre, M. [CEA/DAM Ile de France, 91 - Bruyeres Le Chatel (France); Aymar, R. [ITER Joint Central Team Garching, Muenchen (Germany)] [and others

2000-09-04

Organized the 9 march 2000 by the SFEN, this meeting on the european program concerning the fusion, showed the utility of the exploitation and the enhancement of the actual technology (JET, Tore Supra, ASDEX) and the importance of the Europe engagement in the ITER program. The physical stakes for the magnetic fusion have been developed with a presentation of the progresses in the knowledge of the stability limits. A paper on the inertial fusion was based on the LMJ (Laser MegaJoule) project. The two blanket concepts chosen in the scope of the european program on the tritium blankets, have been discussed. These concepts will be validated by irradiation tests in the ITER-FEAT and adapted for a future reactor. (A.L.B.)

Effect of ITER components manufacturing cycle on the irradiation behaviour of 316L(N)-IG steel

International Nuclear Information System (INIS)

Rodchenkov, B.S.; Prokhorov, V.I.; Makarov, O.Yu.; Shamardin, V.K.; Kalinin, G.M.; Strebkov, Yu.S.; Golosov, O.A.

2000-01-01

The main options for the manufacturing of high heat flux (HHF) components is hot isostatic pressing (HIP) using either solid pieces or powder. There was no database on the radiation behaviour of these materials, and in particular stainless steel (SS) 316L(N)-IG with ITER components manufacturing thermal cycle. Irradiation of wrought steel, powder-HIP, solid-HIP and HIPed joints has been performed within the framework of an ITER task. Specimens cut from 316L(N)-IG plate, HIP products, and solid-HIP joints were irradiated in the SM-3 reactor in Dimitrovgrad up to 4 and 10 dpa at 175 deg. C and 265 deg. C. The paper describes the results of post-irradiation tensile and fracture toughness tests

Fusion reactor materials

Energy Technology Data Exchange (ETDEWEB)

none,

1989-01-01

This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

Fusion reactor materials

International Nuclear Information System (INIS)

1989-01-01

This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics

Integrated Computational study of Material Lifetime in a Fusion Reactor Environment

Energy Technology Data Exchange (ETDEWEB)

Gilbert, M.; Dudarev, S.; Packer, L.; Zheng, S.; Sublet, J.-C., E-mail: mark.gilbert@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Centre for Fusion Energy, Abingdon (United Kingdom)

2012-09-15

Full text: The high-energy, high-intensity neutron fluxes produced by the fusion plasma will have a significant life-limiting impact on reactor components in both experimental and commercial fusion devices. Not only do the neutrons bombarding the materials induce atomic displacement cascades, leading to the accumulation of structural defects, but they also initiate nuclear reactions, which cause transmutation of the elemental atoms. Understanding the implications associated with the resulting compositional changes is one of the key outstanding issues related to fusion energy research. Several complimentary computational techniques have been used to investigate the problem. Firstly, neutron-transport simulations, performed on a reference design for the demonstration fusion power plant (DEMO), quantify the variation in neutron irradiation conditions as a function of geometry. The resulting neutron fluxes and spectra are then used as input into inventory calculations, which allow for the compositional changes of a material to be tracked in time. These calculations reveal that the production of helium (He) gas atoms, whose presence in a material is of particular concern because it can accumulate and cause swelling and embrittlement, will vary significantly, even within the same component of a reactor. Lastly, a density-functional-based model for He-induced grain-boundary embrittlement has been developed to predict the life-limiting consequences associated with relatively low concentrations of He in materials situated at various locations in the DEMO structure. The results suggest that some important fusion materials may be significantly more susceptible to this type of failure than others. (author)

Geometric component of charge pumping current in nMOSFETs due to low-temperature irradiation

Science.gov (United States)

Witczak, S. C.; King, E. E.; Saks, N. S.; Lacoe, R. C.; Shaneyfelt, M. R.; Hash, G. L.; Hjalmarson, H. P.; Mayer, D. C.

2002-12-01

The geometric component of charge pumping current was examined in n-channel metal-oxide-silicon field effect transistors (MOSFETs) following low-temperature irradiation. In addition to the usual dependencies on channel length and gate bias transition time, the geometric component was found to increase with radiation-induced oxide-trapped charge density and decreasing temperature. A postirradiation injection of electrons into the gate oxide reduces the geometric component along with the density of oxide-trapped charge, which clearly demonstrates that the two are correlated. A fit of the injection data to a first-order model for trapping kinetics indicates that the electron trapping occurs predominantly at a single type of Coulomb-attractive trap site. The geometric component results primarily from the bulk recombination of channel electrons that fail to transport to the source or drain during the transition from inversion to accumulation. The radiation response of these transistors suggests that Coulomb scattering by oxide-trapped charge increases the bulk recombination at low temperatures by impeding electron transport. These results imply that the geometric component must be properly accounted for when charge pumping irradiated n-channel MOSFETs at low temperatures.

Fusion Canada issue 11

Energy Technology Data Exchange (ETDEWEB)

NONE

1990-06-01

A short bulletin from the National Fusion Program. Included in this issue is a report on operation at Tokamak de Varennes, CRITIC irradiations at AECL, Tritium systems at TFTR, physics contribution at ITER. 4 figs.

Modelling irradiation effects in fusion materials

DEFF Research Database (Denmark)

Victoria, M.; Dudarev, S.; Boutard, J.L.

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...

Safety analysis of water cooled components inside the JET thermonuclear fusion tokamak

International Nuclear Information System (INIS)

Ageladarakis, P.; O'Dowd, N.; Papastergiou, S.

1998-04-01

The transient thermal behaviour of a number of components, installed in the vessel of the world's largest Fusion Tokamak (JET) has been examined with a theoretical model, which simulated normal operational conditions and abnormal scenarios namely: Loss of Coolant Flow; Loss of Torus Vacuum; and combinations. A number of theoretical results related to water and cryogenically cooled devices have been validated by a comprehensive experimental campaign conducted both inside the JET plasma chamber and in a test rig. The performance of water cooled components which may be subjected to boiling or freeze-up risks in case of a Loss of Water Flow event has also been analysed. Time constants of transient temperature changes were determined by the model while protective actions were prescribed in order to safeguard the equipment against associated risks. A completely automatic safety protection system has been designed on the basis of these analyses and implemented in the routine JET operation. During operation of JET the safety code reacted several times within the specified time limits and protected the relevant components during real off-normal events. (author)

The use of industrial type control and monitoring components for a large fusion experiment

International Nuclear Information System (INIS)

Hemming, O.N.; Manduchi, G.; Luchetta, A.; Schmidt, V.; Vitturi, S.

1994-01-01

RFX is one of the large nuclear fusion experiments within the framework of the co-ordinated nuclear fusion research programme of the European Community. During the control system design phase in 1986, the increase in power and flexibility of industrial type programmable controllers lead to the decision for a complete physical split of control, monitoring and data acquisition functions according to speed requirements, allowing the exploitation of the relative advantages of both CAMAC and programmable controllers. The 'slow' control and monitoring functions (for about 4000 digital and 200 analog I/O signals with scanning times of similar 1 second) have been implemented using a series of networked industrial PLCs and personal computers. This has allowed us to choose from a wide range of off-the-shelf hardware and software components for the plant interface and to utilize specialized expertise from the industrial field for the application software implementation. The paper gives the expectations and results gained from this design choice and how it has influenced the decisions for the evolution of the system over the next few years with the utilization of new industrial hardware components. Details are also given regarding the system integration (via the Ethernet network) with the VAX-based CAMAC frontend fast control and data-acquisition system. ((orig.))

Annual report for the steering committee of the association Euratom-Belgian State for fusion 1994

International Nuclear Information System (INIS)

Moons, F.

1994-09-01

The work performed in 1994 at SCK/CEN in the field of technology for thermonuclear fusion is reported. It mainly embraces: (1) neutron irradiation effects on beryllium, (2) corrosion experiments on copper and molybdenum, (3) radiation hardening of opto-electronic components, (4) environmental tolerance and testing optical fibres assessment, (5) electrolytic recovery of elemental tritium from highly tritiated water, (6) in pile AISI 316L low cycle fatigue

FMIT: an accelerator-based neutron factory for fusion materials qualification

International Nuclear Information System (INIS)

Burke, R.J.; Hagan, J.W.; Trego, A.L.

1983-01-01

The Fusion Materials Irradiation Test Facility will provide a unique testing environment for irradiation of structural and special-purpose materials in support of fusion-power systems. The neutron source will be produced by a deuteron-lithium stripping reaction to generate high-energy neutrons to ensure materials damage characteristic of the deuterium-tritium power system. The facility, its testing role, the status, and major aspects of its design and supporting system development are described. Emphasis is given to programmatic elements and features incorporated in the accelerator and other systems to assure that the FMIT runs as a highly reliable fusion materials testing installation

Ceramics for fusion applications

International Nuclear Information System (INIS)

Clinard, F.W. Jr.

1986-01-01

Ceramics are required for a variety of uses in both near-term fusion devices and in commercial powerplants. These materials must retain adequate structural and electrical properties under conditions of neutron, particle, and ionizing irradiation; thermal and applied stresses; and physical and chemical sputtering. Ceramics such as Al 2 O 3 , MgAl 2 O 4 , BeO, Si 3 N 4 and SiC are currently under study for fusion applications, and results to date show widely-varying response to the fusion environment. Materials can be identified today which will meet initial operating requirements, but improvements in physical properties are needed to achieve satisfactory lifetimes for critical applications

Ceramics for fusion applications

International Nuclear Information System (INIS)

Clinard, F.W. Jr.

1987-01-01

Ceramics are required for a variety of uses in both near-term fusion devices and in commercial powerplants. These materials must retain adequate structural and electrical properties under conditions of neutron, particle and ionizing irradiation; thermal and applied stresses; and physical and chemical sputtering. Ceramics such as Al 2 O 3 , MgAl 2 O 4 , BeO, Si 3 N 4 and SiC are currently under study for fusion applications, and results to date show widely-varying responses to the fusion environment. Materials can be identified today that will meet initial operating requirements, but improvements in physical properties are needed to achieve satisfactory lifetimes for critical applications. (author)

Brazilian situation of blood component irradiation practice for the prevention of transfusion associated Graft-versus-Host disease

International Nuclear Information System (INIS)

Goes, E.G.; Borges, J.C.; Covas, D.T.; Motta, I.

1998-01-01

Transfusion-associated graft-versus-host disease (TA-GVHD) is a usually complication of transfusion of blood component containing T lymphocytes what recently has also involved immunocompetent patient. Gamma irradiation of cellular blood components has been the mainstay against TA-GVHD, nevertheless there is little information in the literature about current transfusion medicine practices regarding gamma irradiation of blood products. This work presents an overview of the Brazilian reality and suggests policies to optimize TA-GVHD prevention. (Author)

Brazilian situation of blood component irradiation practice for the prevention of transfusion associated Graft-versus-Host disease

Energy Technology Data Exchange (ETDEWEB)

Goes, E.G.; Borges, J.C. [EE/COPPE-UFRJ (Brazil); Covas, D.T. [Faculdade deMedicina-USP-RP (Brazil); Motta, I. [Instituto Nacional do Cancer- Rio deJaneiro (Brazil)

1998-12-31

Transfusion-associated graft-versus-host disease (TA-GVHD) is a usually complication of transfusion of blood component containing T lymphocytes what recently has also involved immunocompetent patient. Gamma irradiation of cellular blood components has been the mainstay against TA-GVHD, nevertheless there is little information in the literature about current transfusion medicine practices regarding gamma irradiation of blood products. This work presents an overview of the Brazilian reality and suggests policies to optimize TA-GVHD prevention. (Author)

The independence of irradiation creep in austenitic alloys of displacement rate and helium to dpa ratio

Energy Technology Data Exchange (ETDEWEB)

Garner, F.A.; Toloczko, M.B. [Pacific Northwest National Lab., Richland, WA (United States); Grossbeck, M.L. [Oak Ridge National Lab., TN (United States)

1997-04-01

The majority of high fluence data on the void swelling and irradiation creep of austenitic steels were generated at relatively high displacement rates and relatively low helium/dpa levels that are not characteristic of the conditions anticipated in ITER and other anticipated fusion environments. After reanalyzing the available data, this paper shows that irradiation creep is not directly sensitive to either the helium/dpa ratio or the displacement rate, other than through their possible influence on void swelling, since one component of the irradiation creep rate varies with no correlation to the instantaneous swelling rate. Until recently, however, the non-swelling-related creep component was also thought to exhibit its own strong dependence on displacement rate, increasing at lower fluxes. This perception originally arose from the work of Lewthwaite and Mosedale at temperatures in the 270-350{degrees}C range. More recently this perception was thought to extend to higher irradiation temperatures. It now appears, however, that this interpretation is incorrect, and in fact the steady-state value of the non-swelling component of irradiation creep is actually insensitive to displacement rate. The perceived flux dependence appears to arise from a failure to properly interpret the impact of the transient regime of irradiation creep.

Confinement inertial fusion. Power reactors of nuclear fusion by lasers

International Nuclear Information System (INIS)

Velarde, G.; Ahnert, C.; Aragones, J.M.; Leira, G; Martinez-Val, J.M.

1980-01-01

The energy crisis and the need of the nuclear fusion energy are analized. The nuclear processes in the laser interation with the ablator material are studied, as well as the thermohydrodinamic processes in the implossion, and the neutronics of the fusion. The fusion reactor components are described and the economic and social impact of its introduction in the future energetic strategies.(author)

Conceptual design of neutron diagnostic systems for fusion experimental reactor

International Nuclear Information System (INIS)

Iguchi, T.; Kaneko, J.; Nakazawa, M.

1994-01-01

Neutron measurement in fusion experimental reactors is very important for burning plasma diagnostics and control, monitoring of irradiation effects on device components, neutron source characterization for in-situ engineering tests, etc. A conceptual design of neutron diagnostic systems for an ITER-like fusion experimental reactor has been made, which consists of a neutron yield monitor, a neutron emission profile monitor and a 14-MeV spectrometer. Each of them is based on a unique idea to meet the required performances for full power conditions assumed at ITER operation. Micro-fission chambers of 235 U (and 238 U) placed at several poloidal angles near the first wall are adopted as a promising neutron yield monitor. A collimated long counter system using a 235 U fission chamber and graphite neutron moderators is also proposed to improve the calibration accuracy of absolute neutron yield determination

Two-pulse driving of D+D nuclear fusion within a single Coulomb exploding nanodroplet

International Nuclear Information System (INIS)

Last, Isidore; Jortner, Joshua; Peano, Fabio; Silva, Luis O.

2010-01-01

This paper presents a computational study of D+D fusion driven by Coulomb explosion (CE) within a single, homonuclear deuterium nanodroplet, subjected to double-pulse ultraintense laser irradiation. This irradiation scheme results in the attainment (by the first weaker pulse) of a transient inhomogeneous density profile, which serves as a target for the driving (by the second superintense pulse) of nonuniform CE that triggers overrun effects and induces intrananodroplet (INTRA) D+D fusion. Scaled electron and ion dynamics simulations were utilized to explore the INTRA D+D fusion yields for double-pulse, near-infrared laser irradiation of deuterium nanodroplets. The dependence of the INTRA yield on the nanodroplet size and on the parameters of the two laser pulses was determined, establishing the conditions for the prevalence of efficient INTRA fusion. The INTRA fusion yields are amenable to experimental observation within an assembly of nanodroplets. The INTRA D+D fusion can be distinguished from the concurrent internanodroplet D+D fusion reaction occurring in the macroscopic plasma filament and outside it in terms of the different energies of the neutrons produced in these two channels.

Irradiation-induced structure and property changes in tokamak plasma-facing, carbon-carbon composites

International Nuclear Information System (INIS)

Burchell, T.D.

1994-01-01

Carbon-carbon composites are an attractive choice for fusion reactor plasma-facing components 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 high thermal conductivity to manage the anticipated severe heat loads. Moreover, ignition machines such as ITER will produce large neutron fluxes. Consequently, the influence of neutron damage on the structure and properties of carbon-carbon composite materials must be evaluated. Data from two irradiation experiments are reported and discussed here. Carbon-carbon composite materials were irradiated in target capsules in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). A peak damage dose of 4.7 displacements per atom (dpa) at 600 degree C was attained. The carbon materials irradiated included uni-directional, two-directional, and three-directional carbon-carbon composites. Dimensional changes are reported for the composite materials and are related to single crystal dimensional changes through fiber and composite structural models. Moreover, the irradiation-induced dimensional changes are reported and discussed in terms of their architecture, fiber type, and graphitization temperature. The effect of neutron irradiation on thermal conductivity of two three-directional, carbon-carbon composites is reported and the recovery of thermal conductivity due to thermal annealing is discussed

The effect of fast neutron irradiation on the superconducting properties of REBCO coated conductors with and without artificial pinning centers

Science.gov (United States)

Fischer, D. X.; Prokopec, R.; Emhofer, J.; Eisterer, M.

2018-04-01

Superconductors are essential components of future fusion power plants. The magnet coils responsible for producing the field required for confining the fusion plasma are exposed to considerable neutron radiation. This makes irradiation studies necessary for understanding the radiation response of the superconductor. High temperature superconductors are promising candidates as magnet coil materials. YBCO and GdBCO tapes of several manufacturers were irradiated to fast neutron fluences of up to 3.9 × 1022 m-2 in the research reactor at the Atominstitut. Low energy neutrons contribute to the fission reactor spectrum but not to the expected spectrum at the fusion magnets. Low energy neutrons have to be shielded in irradiation experiments to avoid their substantial effect on the superconducting properties of tapes containing gadolinium. The critical current (I c) of the tapes in this study was examined at fields of up to 15 T and down to a temperature of 30 K. I c first increases upon irradiation and reaches a maximum at a certain fluence, which depends highly on temperature, being highest at low temperature. I c declines at high fluences and eventually degrades with respect to its initial value. Tapes with artificial pinning centers (APCs) degrade at lower fluences than tapes without them. The n-values decrease in all types of tapes after irradiation even when the critical currents are increased. The field dependence of the volume pinning force differs in pristine tapes with and without APCs but shows the same behavior after irradiation.

Development of 'low activation superconducting wire' for an advanced fusion reactor

International Nuclear Information System (INIS)

Hishinuma, Y.; Yamada, S.; Sagara, A.; Kikuchi, A.; Takeuchi, T.; Matsuda, K.; Taniguchi, H.

2011-01-01

In the D-T burning plasma reactor beyond ITER such as DEMO and fusion power plants assuming the steady-state and long time operation, it will be necessary to consider carefully induced radioactivity and neutron irradiation properties on the all components for fusion reactors. The decay time of the induced radioactivity can control the schedule and scenarios of the maintenance and shutdown on the fusion reactor. V 3 Ga and MgB 2 compound have shorter decay time within 1 years and they will be desirable as a candidate material to realize 'low activation and high magnetic field superconducting magnet' for advanced fusion reactor. However, it is well known that J c -B properties of V 3 Ga and MgB 2 wires are lower than that of the Nb-based A15 compound wires, so the J c -B enhancements on the V 3 Ga and MgB 2 wires are required in order to apply for an advanced fusion reactor. We approached and succeeded to developing the new process in order to improve J c properties of V 3 Ga and MgB 2 wires. In this paper, the recent activities for the J c improvements and detailed new process in V 3 Ga and MgB 2 wires are investigated. (author)

Observation of stars produced during cold fusion

International Nuclear Information System (INIS)

Matsumoto, T.

1992-01-01

It has been indicated tht multiple-neutron nuclei such as quad-neutrons can be emitted during cold fusion. These multiple-neutrons might bombard the nuclei of materials outside a cold fusion cell to cause nuclear reactions. In this paper, observations of nuclear emulsions that were irradiated during a cold fusion experiment with heavy water and palladium foil are described. Various traces, like stars, showing nuclear reactions caused by the multiple-neutrons have been clearly observed

Irradiation tests of a small-sized motor with radiation resistance

International Nuclear Information System (INIS)

Nakamichi, M.; Ishitsuka, E.; Shimakawa, S.; Kan, S.

2007-01-01

In the Test Blanket Module (TBM) of the International Thermonuclear Experimental Reactor (ITER), tritium production and release behavior will be studied using neutrons from fusion reactions, as the blanket development for a demonstration (DEMO) reactor. For development of the TBM, in-pile functional tests are planned, including an integrated irradiation experiment of a fusion blanket mock-up for pulsed operation simulating the ITER operation mode, using the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Agency (JAEA).Due to be installed in an irradiation rig, a small-sized motor has to be developed for rotating a neutron absorber with a window to realize the simulated pulse operation. Since degradation of materials of the motor may be caused by radiation damage due to neutron and gamma-ray irradiation, it is important to examine the soundness of the motor materials under the neutron and gamma irradiation.In the present study, a small-sized motor with increased radiation resistance was developed as follows. A design of a commercial alternate current (AC) servomotor was adopted in the base structure, and some components of the motor were replaced by those made of radiation-proof materials, through elimination of organic materials. Polyester-coated wire for field coil and epoxy for fixed resin were replaced by polyimide-coated wire and polysiloxane filled with MgO and Al 2 O 3 , respectively. Furthermore, inorganic lubricant (Mo-based coating of 4 micro meter in thickness) was treated on the surface of a gear, instead of organic (polyphenylether) oil.Radiation-induced degradation of the components of the developed small-sized motor was examined using JMTR and the Japan Research Reactor No.4 (JRR-4) of JAEA. The motor was operating normally up to a gamma-ray dose of 7 x 10 8 Gy, a fast neutron (E>1 MeV) fluence of 2 x 10 21 m -2 and a thermal neutron (E 22 m -2 . The irradiated gamma-ray dose for this motor is about 700 times as high as the operation

Irradiation effect on chemical components of oil palm empty fruit bunch and palm press fibre

International Nuclear Information System (INIS)

Zainon Othman; Mat Rasol Awang; Hassan Hamdani Mutaat; Tamikazu Kume; Hitoshi Ito; Shinpei Matsuhashi; Ishigaki, I.

1998-01-01

Physico-chemical properties of empty fruit bunch (EFB) and palm press fibre (PPF), which are major by-products of the oil palm industries, were studied for upgrading their utilisation as animal feed by radiation-fermentation process. Comparative analyses of raw EFB and PPF from 3 different mills showed significant variations in some of their chemical components. Significant differences were also observed between the chemical components of EFB and PPF samples. The water holding capacities (WHC) of both EFB and PPF suggested their suitability for use as fermentation media. Gamma irradiation of up to 50 kGy have little effect on the components of both EFB and PPF. Irradiation dose of 25 kGy appeared to produce enhancement effect on cellulase hydrolysis of holocellulose and alpha-cellulose of EFB but a retarding effect on hydrolysis of PPF

Effects of Gamma Irradiation on Active Components in Essential Oils of Cinnamomum verum J.S.Presl

International Nuclear Information System (INIS)

Thongphasuk, Piyanuch; Thongphasuk, Jarunee; Eamsiri, Jarurut; Pongpat, Suchada

2009-07-01

Full text: Gamma irradiation is one of the methods utilized to reduce microbial contamination of medicinal herbs. Since irradiation may also affect active compounds of the irradiated herbs, the objective of this research is to study the effect of gamma irradiation (10 and 25 kGy) from cobalt-60 on active compounds in essential oils of Cinnamomum verum J.S.Presl by using GC-MS. The results showed that gamma irradiation at the dose of 10 and 25 kGy does not significantly affect active components in essential oils such as alpha-pinene, camphene, 1,8-cineole, alpha-copaene, benzaldehyde, linalool, bornyl acetate, terpinen-4-0l, alpha-terpineol, benzylacetaldehyde, Z-cinnamaldehyde, E-cinnamaldehyde, and cinnamic acid

Some applications of fission-based testing capabilities in the development of fusion technology

International Nuclear Information System (INIS)

Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

1981-10-01

The testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for fusion testing when they do become available. Fission testing is capable of filling many gaps in fusion reactor design information, and should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, work was performed on the irradiation of magnet insulators. This work is continuing with an improved test environment. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, a conceptual design is presented for a fission-based Integrated Test Facility (ITF), which can accommodate entire wall/blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

Radiation damage in CTR magnet components

International Nuclear Information System (INIS)

Ullmaier, H.

1976-01-01

Data are reviewed (already existing or to be acquired) which should allow prediction of the behavior of large superconducting coils in the radiation field of a future fusion reactor. The electrical and mechanical stability of such magnets is determined by the irradiation induced deterioration of the magnet components, i.e., (a) changes in critical current, field and temperature of the superconductor (NbTi, A-15 phases), (b) resistivity increase in the stabilizer (Cu, Al), and (c) changes in mechanical and dielectric properties of insulators and spacers. Recent low temperature simulation experiments (with fission neutrons and heavy ions) show that the superconductor will not be the critical component of a fusion magnet--at least as far as radiation damage is concerned. Much more severe is the loss of stability due to the resistivity increase of the stabilizing material. It seems, however, that the magnitude of this effect can be predicted rather reliably and therefore taken into account in the coil design. Almost no data exist about the low temperature behavior of insulator and spacer materials in a radiation field. Furthermore, very little is known about the nature of the radiation damage in non-metals, which makes extrapolations of the few existing data to other materials or to other doses highly speculative. Only future experiments can decide if the insulators will be the limiting component of a CTR magnet or not

Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

International Nuclear Information System (INIS)

Abdou, M.A.; Boyd, R.D.; Easor, J.R.

1984-06-01

A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted

Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

Energy Technology Data Exchange (ETDEWEB)

Abdou, M.A.; Boyd, R.D.; Easor, J.R.; Gauster, W.B.; Gordon, J.D.; Mattas, R.F.; Morgan, G.D.; Ulrickson, M.A,; Watson, R.D.; Wolfer, W.G,

1984-06-01

A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted.

Structural integrity of stainless steel components exposed to neutron irradiation. Change in failure strength of cracked components due to cold working

International Nuclear Information System (INIS)

Kamaya, Masayuki; Hojo, Tomohiro; Mochizuki, Masahito

2015-01-01

Load carrying capacity of austenitic stainless steel component is increased due to hardening caused by neutron irradiation if no crack is included in the component. On the other hand, if a crack is initiated in the reactor components, the hardening may decrease the load carrying capacity due to reduction in fracture toughness. In this paper, in order to develop a failure assessment procedure of irradiated cracked components, characteristics of change in failure strength of stainless steels due to cold working were investigated. It was experimentally shown that the proof and tensile strengths were increased by the cold working, whereas the fracture toughness was decreased. The fracture strengths of a cylinder with a circumferential surface crack were analyzed using the obtained material properties. Although the cold working altered the failure mode from plastic collapse to the unsteady ductile crack growth, it did not reduce failure strengths even if 50% cold working was applied. The increase in failure strength was caused not only by increase in flow stress but also by reduction in J-integral value, which was brought by the change in stress-strain curve. It was shown that the failure strength of the hardened stainless steel components could be derived by the two-parameter method, in which the change in material properties could be reasonably considered. (author)

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.)

The FUBR-1B experiment, irradiation of lithium ceramics to high burnups under large temperature gradients

International Nuclear Information System (INIS)

Hollenberg, G.W.; Knight, R.C.; Densley, P.J.; Pember, L.A.; Johnson, C.E.; Poeppel, R.B.; Yang, L.

1985-01-01

Solid breeder materials used for supplying the tritium for fueling fusion power reactors will be required to withstand a variety of severe environmental conditions such as irradiation damage, thermal stresses and chemical reactions while continuing to produce tritium and not interfering with other essential components in the complex blanket region. In the FUBR-1B experiment several solid breeder candidates are being subjected to the most hostile conditions foreseen in a fusion reactor's blanket. Some material, such as Li 2 O, Li 8 ZrO 6 , and Li 4 SiO 4 , possess high lithium atom densities which are reflected in high tritium breeding ratios. Other material, such as LiAlO 2 and Li 2 ZrO 3 , appear to have exceptional irradiation stability. Verifying the magnitude of these differences will allow national selection between design options

Development of vanadium base alloys for fusion first-wall/blanket applications

International Nuclear Information System (INIS)

Smith, D.L.; Chung, H.M.; Loomis, B.A.; Matsui, H.; Votinov, S.; VanWitzenburg, W.

1994-01-01

Vanadium alloys have been identified as a leading candidate material for fusion first-wall/blanket applications. Certain vanadium alloys exhibit favorable safety and environmental characteristics, good fabricability, high temperature and heat load capability, good compatibility with liquid metals and resistance to irradiation damage effects. The current focus is on vanadium alloys with (3-5)% Cr and (3-5)% Ti with a V-4Cr-4Ti alloy as the leading candidate. Preliminary results indicate that the crack-growth rates of certain alloys are not highly sensitive to irradiation. Results from the Dynamic Helium Charging Experiment (DHCE) which simulates fusion relevant helium/dpa ratios are similar to results from neutron irradiated material. This paper presents an overview of the recent results on the development of vanadium alloys for fusion first wall/blanket applications

Fusion reactor materials

International Nuclear Information System (INIS)

Rowcliffe, A.F.; Burn, G.L.; Knee', S.S.; Dowker, C.L.

1994-02-01

This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

Fusion: A necessary component of US energy policy

International Nuclear Information System (INIS)

Correll, D.L. Jr.

1989-01-01

US energy policy must ensure that its security, its economy, or its world leadership in technology development are not compromised by failure to meet the nation's electrical energy needs. Increased concerns over the greenhouse effect from fossil-fuel combustion mean that US energy policy must consider how electrical energy dependence on oil and coal can be lessened by conservation, renewable energy sources, and advanced energy options (nuclear fission, solar energy, and thermonuclear fusion). In determining how US energy policy is to respond to these issues, it will be necessary to consider what role each of the three advanced energy options might play, and to determine how these options can complement one another. This paper reviews and comments on the principal US studies and legislation that have addressed fusion since 1980, and then suggests a research, development, and demonstration program that is consistent with the conclusions of those prior authorities and that will allow us to determine how fusion technology can fit into a US energy policy that takes a balanced, long term view of US needs. 17 refs

The materials production and processing facility at the Spanish National Centre for fusion technologies (TechnoFusion)

International Nuclear Information System (INIS)

Munoz, A.; Monge, M.A.; Pareja, R.; Hernandez, M.T.; Jimenez-Rey, D.; Roman, R.; Gonzalez, M.; Garcia-Cortes, I.; Perlado, M.; Ibarra, A.

2011-01-01

In response to the urgent request from the EU Fusion Program, a new facility (TechnoFusion) for research and development of fusion materials has been planned with support from the Regional Government of Madrid and the Ministry of Science and Innovation of Spain. TechnoFusion, the National Centre for Fusion Technologies, aims screening different technologies relevant for ITER and DEMO environments while promoting the contribution of international companies and research groups into the Fusion Programme. For this purpose, the centre will be provided with a large number of unique facilities for the manufacture, testing (a triple-beam multi-ion irradiation, a plasma-wall interaction device, a remote handling for under ionizing radiation testing) and analysis of critical fusion materials. Particularly, the objectives, semi-industrial scale capabilities and present status of the TechnoFusion Materials Production and Processing (MPP) facility are presented. Previous studies revealed that the MPP facility will be a very promising infrastructure for the development of new materials and prototypes demanded by the fusion technology and therefore some of them will be here briefly summarized.

The materials production and processing facility at the Spanish National Centre for fusion technologies (TechnoFusion)

Energy Technology Data Exchange (ETDEWEB)

Munoz, A., E-mail: rpp@fis.uc3m.es [Departamento de Fisica, UC3M, Avda de la Universidad 30, 28911 Leganes, Madrid (Spain); Monge, M.A.; Pareja, R. [Departamento de Fisica, UC3M, Avda de la Universidad 30, 28911 Leganes, Madrid (Spain); Hernandez, M.T. [LNF-CIEMAT, Avda, Complutense, 22, 28040 Madrid (Spain); Jimenez-Rey, D. [CMAM, UAM, C/Faraday 3, 28049, Madrid (Spain); Roman, R.; Gonzalez, M.; Garcia-Cortes, I. [LNF-CIEMAT, Avda, Complutense, 22, 28040 Madrid (Spain); Perlado, M. [IFN, ETSII, UPM, C/Jose Gutierrez Abascal, 2, 28006 Madrid (Spain); Ibarra, A. [LNF-CIEMAT, Avda, Complutense, 22, 28040 Madrid (Spain)

2011-10-15

In response to the urgent request from the EU Fusion Program, a new facility (TechnoFusion) for research and development of fusion materials has been planned with support from the Regional Government of Madrid and the Ministry of Science and Innovation of Spain. TechnoFusion, the National Centre for Fusion Technologies, aims screening different technologies relevant for ITER and DEMO environments while promoting the contribution of international companies and research groups into the Fusion Programme. For this purpose, the centre will be provided with a large number of unique facilities for the manufacture, testing (a triple-beam multi-ion irradiation, a plasma-wall interaction device, a remote handling for under ionizing radiation testing) and analysis of critical fusion materials. Particularly, the objectives, semi-industrial scale capabilities and present status of the TechnoFusion Materials Production and Processing (MPP) facility are presented. Previous studies revealed that the MPP facility will be a very promising infrastructure for the development of new materials and prototypes demanded by the fusion technology and therefore some of them will be here briefly summarized.

Damage analysis and fundamental studies for fusion reactor materials development

International Nuclear Information System (INIS)

Odette, G.R.; Lucas, G.E.

1991-09-01

The philosophy of the program at the University of California Santa Barbara has been to develop a fundamental understanding of both the basic damage processes and microstructural evolution that take place in a material during neutron irradiation and the consequent dimensional and mechanical property changes. This fundamental understanding can be used in conjunction with empirical data obtained from a variety of irradiation facilities to develop physically-based models of neutron irradiation effects in structural materials. The models in turn can be used to guide alloy development and to help extrapolate the irradiation data base (expected to be largely fission reactor based) to the fusion reactor regime. This philosophy is consistent with that of the national and international programs for developing structural materials for fusion reactors

Electro-chemically-based technologies for processing of tungsten components in fusion technology

International Nuclear Information System (INIS)

Holstein, N.; Konys, J.; Krauss, W.; Lorenz, J.

2010-01-01

In fusion technology layers and bulk components fabricated from tungsten and W-alloys are used as functional materials, e.g. as coatings of blanket modules or T-permeation barriers and also as structural components in a He-cooled divertor. Their application under high heat loads and temperatures is besides manufacturing, also challenging regarding joining, caused e.g. by expansion mismatches in combination with steel or other diffusion issues. Driven by these needs, electro-chemically-based technologies were analyzed concerning their advantages in processing in the fields of soft structuring of tungsten alloys and in deposition of functional scales. The Electro-Chemistry (EC) of tungsten is characterized by its affection to build up passivation layers in aqueous media during the initial oxidation, which is the result of an unavoidable basic electrochemical reaction with water (W + 3H 2 O → WO 3 + 3H 2 ), although the element standard potential is situated between common EC material like iron and copper. (orig.)

Present status of SiCf/SiC composites as low-activation structural materials of fusion reactor in Japan

International Nuclear Information System (INIS)

Kohyama, A.; Katoh, Y.; Hasegawa, A.; Noda, T.

2001-01-01

The outline of research subjects on SiCf/SiC composites to apply to the structural components of fusion reactors are described and present status on material development of SiCf/SiC composites in Japan is reviewed. Irradiation experiments of the composites using fission reactors conducted by international collaborations to clarify their radiation response and to optimize the fabrication processes are introduced. (author)

Near-surface thermal characterization of plasma facing components using the 3-omega method

International Nuclear Information System (INIS)

Dechaumphai, Edward; Barton, Joseph L.; Tesmer, Joseph R.; Moon, Jaeyun; Wang, Yongqiang; Tynan, George R.; Doerner, Russell P.; Chen, Renkun

2014-01-01

Near-surface regime plays an important role in thermal management of plasma facing components in fusion reactors. Here, we applied a technique referred to as the ‘3ω’ method to measure the thermal conductivity of near-surface regimes damaged by ion irradiation. By modulating the frequency of the heating current in a micro-fabricated heater strip, the technique enables the probing of near-surface thermal properties. The technique was applied to measure the thermal conductivity of a thin ion-irradiated layer on a tungsten substrate, which was found to decrease by nearly 60% relative to pristine tungsten for a Cu ion dosage of 0.2 dpa

Joint ICFRM-14 (14. international conference on fusion reactor materials) and IAEA satellite meeting on cross-cutting issues of structural materials for fusion and fission applications. PowerPoint presentations

International Nuclear Information System (INIS)

2009-01-01

The Conference was devoted to the challenges in the development of new materials for advanced fission, fusion and hybrid reactors. The topics discussed include fuels and materials research under the high neutron fluence; post-irradiation examination; development of radiation resistant structural materials utilizing fission research reactors; core materials development for the advanced fuel cycle initiative; qualification of structural materials for fission and fusion reactor systems; application of charged particle accelerators for radiation resistance investigations of fission and fusion structural materials; microstructure evolution in structural materials under irradiation; ion beams and ion accelerators

Repetitive Solid Spherical Pellet Injection and Irradiation toward the Repetitive-mode Fast-Ignition Fusion miniReactor CANDY

International Nuclear Information System (INIS)

HANAYAMA, Ryohei; KOMEDA, Osamu; NISHIMURA, Yasuhiko; MORI, Yoshitaka; ISHII, Katsuhiro; NAKAYAMA, Suisei; OKIHARA, Shinichiro; FUJITA, Kazuhisa; SEKINE, Takashi; SATO, Nakahiro; KAWASHIMA, Toshiyuki; KAN, Hirofumi; KURITA, Takashi; NAKAMURA, Naoki; KONDO, Takuya; FUJINE, Manabu; AZUMA, Hirozumi; HIOKI, Tatsumi; KAKENO, Mitsutaka; MOTOHIRO, Tomoyoshi

2016-01-01

Pellet injection and repetitive laser illumination are key technologies for realizing inertial fusion energy [1-4]. Neutron generator using lasers also requires a repeating pellet target supplier. Here we present the first demonstration of target injection and neutron generation[5]. We injected more than 1300 spherical deuterated polystyrene(C 8 D 8 ) bead pellet targets during 23 minutes at 1 Hz(Fig. 1). After the pellet targets fell for a distance of 18 cm, we applied the synchronized laser-diode-pumped ultra-intense laser HAMA. The laser intensity at the focal point is 5 x 10 18 W/cm 2 , which is high enough to generate neutrons. As a result of the irradiation, we produced 2.45-MeV DD neutrons. Figure 2 shows the neutron time-of-flight signals detected by plastic scintillators coupled to photomultipliers. The neutron energy was calculated by the time-of-flight method. The maximum neutron yield was 9.5 x 10 4 /4π sr. The result is a step toward fusion power and also suggests possible industrial neutron sources. (paper)

Toxicological safety and stability of the components of an irradiated Korean medicinal herb, Paeoniae Radix

International Nuclear Information System (INIS)

Yu, Y.-B.; Jeong, I.-Y.; Park, H.-R.; Oh, Heon; Jung, Uhee; Jo, S.-K.

2004-01-01

As utilization of medicinal herbs in food and bio-industry increases, mass production and the supply of herbs with a high quality are required. As the use of fumigants and preservatives for herbs is being restricted, safe hygienic technologies are demanded. To consider the possibility of the application of irradiation technology for this purpose, the genotoxicological safety and stability of the active components of the γ-irradiated Paeoniae Radix were studied. The herb was irradiated with γ-rays at a practical dosage of 10 kGy, and then it was extracted with hot water. The genotoxicity of the extract of the irradiated herb was examined in two short-term in vitro tests: (1) Ames test in Salmonella typhimurium; (2) Micronucleus test in cultured Chinese hamster ovary (CHO) cells. The extract of the irradiated herb did not show mutagenicity in the Ames test of the Salmonella reverse mutation assay, and did not show cytogenetic toxicity in the culture of the CHO cells. HPLC chromatogram of paeoniflorin in the irradiated Paeoniae Radix was similar with that of the non-irradiated sample. The quantity of paeoniflorin did not change significantly with irradiation. These results suggest that γ-irradiated Paeoniae Radix is toxicologically safe and chemically stable

Toxicological safety and stability of the components of an irradiated Korean medicinal herb, Paeoniae Radix

Science.gov (United States)

Yu, Young-Beob; Jeong, Ill-Yun; Park, Hae-Ran; Oh, Heon; Jung, Uhee; Jo, Sung-Kee

2004-09-01

As utilization of medicinal herbs in food and bio-industry increases, mass production and the supply of herbs with a high quality are required. As the use of fumigants and preservatives for herbs is being restricted, safe hygienic technologies are demanded. To consider the possibility of the application of irradiation technology for this purpose, the genotoxicological safety and stability of the active components of the γ-irradiated Paeoniae Radix were studied. The herb was irradiated with γ-rays at a practical dosage of 10 kGy, and then it was extracted with hot water. The genotoxicity of the extract of the irradiated herb was examined in two short-term in vitro tests: (1) Ames test in Salmonella typhimurium; (2) Micronucleus test in cultured Chinese hamster ovary (CHO) cells. The extract of the irradiated herb did not show mutagenicity in the Ames test of the Salmonella reverse mutation assay, and did not show cytogenetic toxicity in the culture of the CHO cells. HPLC chromatogram of paeoniflorin in the irradiated Paeoniae Radix was similar with that of the non-irradiated sample. The quantity of paeoniflorin did not change significantly with irradiation. These results suggest that γ-irradiated Paeoniae Radix is toxicologically safe and chemically stable.

Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

Energy Technology Data Exchange (ETDEWEB)

Ulrickson, M.A.; Stevens, P.L.; Hino, T.; Hirohata, Y. [eds.

1996-12-01

This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage.

Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

International Nuclear Information System (INIS)

Ulrickson, M.A.; Stevens, P.L.; Hino, T.; Hirohata, Y.

1996-12-01

This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage

Design of intense neutron source for fusion material study and the role of universities

International Nuclear Information System (INIS)

Ishino, Shiori

1993-01-01

Need and requirement for the intense neutron source for fusion materials study have been discussed for many years. Recently, international climate has been becoming gradually maturing to consider this problem more seriously because of the recognition of crucial importance of solving materials problems for fusion energy development. The present symposium was designed to discuss the problems associated with the intense neutron source for material irradiation studies which will have a potential for the National Institute for Fusion Science to become one of the important future research areas. The symposium comprises five sessions; first, the role of materials research in fusion development strategies was discussed followed by a brief summary of current IFMIF (International Fusion Materials Irradiation Facility) activity. Despite the pressing need for intense fusion neutron source, currently available neutron sources are reactor or accelerator based sources of which FFTF and LASREF were discussed. Then, various concepts of intense neutron source candidates were presented including ESNIT, which are currently under design by JAERI. In the fourth session, discussions were made on the study of materials with the intense neutron source from the viewpoint of materials scientists and engineers as the user of the facility. This is followed by discussions on the role of universities from the two stand points, namely, fusion irradiation studies and fusion materials development. Finally summary discussions were made by the participants, indicating important role fundamental studies in universities for the full utilization of irradiation data and the need of pure 14 MeV neutron source for fundamental studies together with the intense surrogate neutron sources. (author)

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

International Nuclear Information System (INIS)

Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi

1999-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-02-01

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

Chemical analysis developments for fusion materials studies

International Nuclear Information System (INIS)

McCown, J.J.; Baldwin, D.L.; Keough, R.F.; Van der Cook, B.P.

1985-04-01

Several projects at Hanford under the management of the Westinghouse Hanford Company have involved research and development (R and D) on fusion materials. They include work on the Fusion Materials Irradiation Test Facility and its associated Experimental Lithium System; testing of irradiated lithium compounds as breeding materials; and testing of Li and Li-Pb alloy reactions with various atmospheres, concrete, and other reactor materials for fusion safety studies. In the course of these projects, a number of interesting and challenging analytical chemistry problems were encountered. They include sampling and analysis of lithium while adding and removing elements of interest; sampling, assaying and compound identification efforts on filters, aerosol particles and fire residues; development of dissolution and analysis techniques for measuring tritium and helium in lithium ceramics including oxides, aluminates, silicates and zirconates. An overview of the analytical chemistry development problems plus equipment and procedures used will be presented

Development of fusion first-wall radiation damage facilities

International Nuclear Information System (INIS)

McElroy, R.J.; Atkins, T.

1986-11-01

The report describes work performed on the development of fusion-reactor first-wall simulation facilities on the Variable Energy Cyclotron, at Harwell, United Kingdom. Two irradiation facilities have been constructed: i) a device for helium and hydrogen filling up to 1000 ppm for post-irradiation mechanical properties studies, and ii) a helium implantation and damage facility for simultaneous injection of helium and radiation damage into a specimen under stress. These facilities are now fully commissioned and are available for investigations of first-wall radiation damage and for intercorrelation of fission- and fusion -reactor materials behaviour. (U.K.)

The management of fusion waste

International Nuclear Information System (INIS)

Hancox, R.; Butterworth, G.J.

1990-01-01

Fusion reactors based on the deuterium-tritium fuel cycle will generate radioactive waste as a result of neutron irradiation of the structural materials and absorption of the tritium fuel. An important issue is whether the volume of this waste and the risks associated with it can be reduced to a sufficiently low level that the environmental advantage of fusion can be maintained without incurring unacceptable additional costs. Information is presented on the radioactive waste expected from the decommissioning of three generations of fusion devices - the JET experiment, NET, and power reactors. The characteristics and probable volumes of this waste are considered, together with the risks associated with its disposal. (author)

The management of fusion waste

International Nuclear Information System (INIS)

Hancox, R.; Butterworth, G.J.

1991-01-01

Fusion reactors based on the deuterium-tritium fuel cycle will generate radioactive waste as a result of neutron irradiation of the structural materials and absorption of the tritium fuel. An important issue is whether the volume of this waste and the risks associated with it can be reduced to a sufficiently low level that the environmental advantage of fusion can be maintained without incurring unacceptable additional costs. Information is presented on the radioactive waste expected from the decommissioning of three generations of fusion devices - the JET experiment, NET, and power reactors. The characteristics and probable volumes of this waste are considered, together with the risks associated with its disposal. (orig.)

[Data fusion and multi-components quantitative analysis for identification and quality evaluation of Gentiana rigescens from different geographical origins].

Science.gov (United States)

Wang, Qin-Qin; Shen, Tao; Zuo, Zhi-Tian; Huang, Heng-Yu; Wang, Yuan-Zhong

2018-03-01

The accumulation of secondary metabolites of traditional Chinese medicine (TCM) is closely related to its origins. The identification of origins and multi-components quantitative evaluation are of great significance to ensure the quality of medicinal materials. In this study, the identification of Gentiana rigescens from different geographical origins was conducted by data fusion of Fourier transform infrared (FTIR) spectroscopy and high performance liquid chromatography (HPLC) in combination of partial least squares discriminant analysis; meanwhile quantitative analysis of index components was conducted to provide an accurate and comprehensive identification and quality evaluation strategy for selecting the best production areas of G. rigescens. In this study, the FTIR and HPLC information of 169 G. rigescens samples from Yunnan, Sichuan, Guangxi and Guizhou Provinces were collected. The raw infrared spectra were pre-treated by multiplicative scatter correction, standard normal variate (SNV) and Savitzky-Golay (SG) derivative. Then the performances of FTIR, HPLC, and low-level data fusion and mid-level data fusion for identification were compared, and the contents of gentiopicroside, swertiamarin, loganic acid and sweroside were determined by HPLC. The results showed that the FTIR spectra of G. rigescens from different geographical origins were different, and the best pre-treatment method was SNV+SG-derivative (second derivative, 15 as the window parameter, and 2 as the polynomial order). The results showed that the accuracy rate of low- and mid-level data fusion (96.43%) in prediction set was higher than that of FTIR and HPLC (94.64%) in prediction set. In addition, the accuracy of low-level data fusion (100%) in the training set was higher than that of mid-level data fusion (99.12%) in training set. The contents of the iridoid glycosides in Yunnan were the highest among different provinces. The average content of gentiopicroside, as a bioactive marker in Chinese

Hybrid indirect-drive/direct-drive target for inertial confinement fusion

Energy Technology Data Exchange (ETDEWEB)

Perkins, Lindsay John

2018-02-27

A hybrid indirect-drive/direct drive for inertial confinement fusion utilizing laser beams from a first direction and laser beams from a second direction including a central fusion fuel component; a first portion of a shell surrounding said central fusion fuel component, said first portion of a shell having a first thickness; a second portion of a shell surrounding said fusion fuel component, said second portion of a shell having a second thickness that is greater than said thickness of said first portion of a shell; and a hohlraum containing at least a portion of said fusion fuel component and at least a portion of said first portion of a shell; wherein said hohlraum is in a position relative to said first laser beam and to receive said first laser beam and produce X-rays that are directed to said first portion of a shell and said fusion fuel component; and wherein said fusion fuel component and said second portion of a shell are in a position relative to said second laser beam such that said second portion of a shell and said fusion fuel component receive said second laser beam.

Fusion-neutron effects on magnetoresistivity of copper stabilizer materials

International Nuclear Information System (INIS)

Guinan, M.W.; Van Konynenburg, R.A.

1983-01-01

The objective of this work is to quantify the changes which occur in the magnetoresistivity of coppers (having various purities and pretreatments, and at magnetic fields up to 12 T during the course of sequential fusion neutron irradiations at about 4 0 K and anneals to room temperature. In conjunction with work in progress by Coltman and Klabunde of ORNL, the results should lead to engineering design data for the stabilizers of superconducting magnets in fusion reactors. These magnets are expected to be irradiated during reactor operation and warmed to room temperature periodically during maintenance

Power balancing of multibeam laser fusion lasers

International Nuclear Information System (INIS)

Seka, W.; Morse, S.; Letzring, S.; Kremens, R.; Kessler, T.J.; Jaanimagi, P.; Keck, R.; Verdon, C.; Brown, D.

1989-01-01

The success of laser fusion depends to a good degree on the ability to compress the target to very high densities of ≥1000 times liquid DT. To achieve such compressions require that the irradiation nonuniformity must not exceed ∼1% rms over the whole time of the compression, particularly during the early phases of irradiation. The stringent requirements for the irradiation uniformity for laser fusion have been known for quite some time but until recently the energy balance was mistakenly equated to power balance. The authors describe their effort on energy balance and irradiation patterns on the target. They significantly improved the laser performance with respect to overall intensity distributions on target including the implementation of distributed (random) phase plates in each high power beam. However, the slightly varying performance of the third harmonic conversion crystals in the twenty-four beams of their laser system was generally compensated for by appropriately adjusted 1.054μm input laser energy. Computational analysis of the results of the recent high density campaign are shown

IFMIF, International Fusion Materials Irradiation Facility conceptual design activity cost report

International Nuclear Information System (INIS)

Rennich, M.J.

1996-12-01

This report documents the cost estimate for the International Fusion Materials Irradiation Facility (IFMIF) at the completion of the Conceptual Design Activity (CDA). The estimate corresponds to the design documented in the Final IFMIF CDA Report. In order to effectively involve all the collaborating parties in the development of the estimate, a preparatory meeting was held at Oak Ridge National Laboratory in March 1996 to jointly establish guidelines to insure that the estimate was uniformly prepared while still permitting each country to use customary costing techniques. These guidelines are described in Section 4. A preliminary cost estimate was issued in July 1996 based on the results of the Second Design Integration Meeting, May 20--27, 1996 at JAERI, Tokai, Japan. This document served as the basis for the final costing and review efforts culminating in a final review during the Third IFMIF Design Integration Meeting, October 14--25, 1996, ENEA, Frascati, Italy. The present estimate is a baseline cost estimate which does not apply to a specific site. A revised cost estimate will be prepared following the assignment of both the site and all the facility responsibilities

Preliminary results of the International Fusion Materials Irradiation Facility deuteron injector

Energy Technology Data Exchange (ETDEWEB)

Gobin, R.; Adroit, G.; Bogard, D.; Bourdelle, G.; Chauvin, N.; Delferriere, O.; Gauthier, Y.; Girardot, P.; Guiho, P.; Harrault, F.; Jannin, J. L.; Loiseau, D.; Mattei, P.; Roger, A.; Sauce, Y.; Senee, F.; Vacher, T. [Commissariat a l' Energie Atomique et aux Energie Alternatives, CEA/Saclay, DSM/IRFU, 91191-Gif/Yvette (France)

2012-02-15

In the framework of the IFMIF-EVEDA project (International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities), CEA/IRFU is in charge of the design, construction, and characterization of the 140 mA continuous deuteron injector, including the source and the low energy beam line. The electron cyclotron resonance ion source which operates at 2.45 GHz is associated with a 4-electrode extraction system in order to minimize beam divergence at the source exit. Krypton gas injection is foreseen in the 2-solenoid low energy beam line. Such Kr injection will allow reaching a high level of space charge compensation in order to improve the beam matching at the radio frequency quadrupole (RFQ) entrance. The injector construction is now completed on the Saclay site and the first plasma and beam production has been produced in May 2011. This installation will be tested with proton and deuteron beams either in pulsed or continuous mode at Saclay before shipping to Japan. In this paper, after a brief description of the installation, the preliminary results obtained with hydrogen gas injection into the plasma chamber will be reported.

Code of a Tokamak Fusion Energy Facility ITER

International Nuclear Information System (INIS)

Yasuhide Asada; Kenzo Miya; Kazuhiko Hada; Eisuke Tada

2002-01-01

The technical structural code for ITER (International Thermonuclear Experimental Fusion Reactor) and, as more generic applications, for D-T burning fusion power facilities (hereafter, Fusion Code) should be innovative because of their quite different features of safety and mechanical components from nuclear fission reactors, and the necessity of introducing several new fabrication and examination technologies. Introduction of such newly developed technologies as inspection-free automatic welding into the Fusion Code is rationalized by a pilot application of a new code concept of s ystem-based code for integrity . The code concept means an integration of element technical items necessary for construction, operation and maintenance of mechanical components of fusion power facilities into a single system to attain an optimization of the total margin of these components. Unique and innovative items of the Fusion Code are typically as follows: - Use of non-metals; - Cryogenic application; - New design margins on allowable stresses, and other new design rules; - Use of inspection-free automatic welding, and other newly developed fabrication technologies; - Graded approach of quality assurance standard to cover radiological safety-system components as well as non-safety-system components; - Consideration on replacement components. (authors)

Effect of gamma irradiation on the fitness component of the pulse beetle, callosobruchus chinensis

International Nuclear Information System (INIS)

Abdelrahman, A.M.; Aboulnasr, A.E.; Roushdy, H.M.; Ahmed, M.Y.; Haiba, I.M.

1984-01-01

The present work deals with the fines components of the F 1 generation of pulse beetle, callosobruchus chinensis, developed from adults irradiated (as newly emerged adults) with higher doses and fractionation of sterilizing doses. The results obtained show a gradual decrease in egg production and hatchability. The dosages caused complete sterility in females and males respectively, shortening the life span of developed adults. The dose 200,000 rad caused immediate death to the irradiated adults. Fractionation of the sterilizing dose had no effect on either longevity or the percent of egg hatchability

Exploring liquid metal plasma facing component (PFC) concepts-Liquid metal film flow behavior under fusion relevant magnetic fields

International Nuclear Information System (INIS)

Narula, M.; Abdou, M.A.; Ying, A.; Morley, N.B.; Ni, M.; Miraghaie, R.; Burris, J.

2006-01-01

The use of fast moving liquid metal streams or 'liquid walls' as a plasma contact surface is a very attractive option and has been looked upon with considerable interest over the past several years, both by the plasma physics and fusion engineering programs. Flowing liquid walls provide an ever replenishing contact surface to the plasma, leading to very effective particle pumping and surface heat flux removal. A key feasibility issue for flowing liquid metal plasma facing component (PFC) systems, pertains to their magnetohydrodynamic (MHD) behavior under the spatially varying magnetic field environment, typical of a fusion device. MHD forces hinder the development of a smooth and controllable liquid metal flow needed for PFC applications. The present study builds up on the ongoing research effort at UCLA, directed towards providing qualitative and quantitative data on liquid metal free surface flow behavior under fusion relevant magnetic fields

Effects of fusion relevant transient energetic radiation, plasma and thermal load on PLANSEE double forged tungsten samples in a low-energy plasma focus device

Science.gov (United States)

Javadi, S.; Ouyang, B.; Zhang, Z.; Ghoranneviss, M.; Salar Elahi, A.; Rawat, R. S.

2018-06-01

Tungsten is the leading candidate for plasma facing component (PFC) material for thermonuclear fusion reactors and various efforts are ongoing to evaluate its performance or response to intense fusion relevant radiation, plasma and thermal loads. This paper investigates the effects of hot dense decaying pinch plasma, highly energetic deuterium ions and fusion neutrons generated in a low-energy (3.0 kJ) plasma focus device on the structure, morphology and hardness of the PLANSEE double forged tungsten (W) samples surfaces. The tungsten samples were provided by Forschungszentrum Juelich (FZJ), Germany via International Atomic Energy Agency, Vienna, Austria. Tungsten samples were irradiated using different number of plasma focus (PF) shots (1, 5 and 10) at a fixed axial distance of 5 cm from the anode top and also at various distances from the top of the anode (5, 7, 9 and 11 cm) using fixed number (5) of plasma focus shots. The virgin tungsten sample had bcc structure (α-W phase). After PF irradiation, the XRD analysis showed (i) the presence of low intensity new diffraction peak corresponding to β-W phase at (211) crystalline plane indicating the partial structural phase transition in some of the samples, (ii) partial amorphization, and (iii) vacancy defects formation and compressive stress in irradiated tungsten samples. Field emission scanning electron microscopy showed the distinctive changes to non-uniform surface with nanometer sized particles and particle agglomerates along with large surface cracks at higher number of irradiation shots. X-ray photoelectron spectroscopy analysis demonstrated the reduction in relative tungsten oxide content and the increase in metallic tungsten after irradiation. Hardness of irradiated samples initially increased for one shot exposure due to reduction in tungsten oxide phase, but then decreased with increasing number of shots due to increasing concentration of defects. It is demonstrated that the plasma focus device provides

An assessment of magnetic effects in ferromagnetic martensitic steels for use in fusion machines

International Nuclear Information System (INIS)

Lechtenberg, T.; Dahms, C.; Attaya, H.

1984-01-01

Interest in the 9-12%Cr class of martensitic stainless steels has accelerated since these materials were included in the U.S. Alloy Development for Irradiation Performance (ADIP) task funded by the Office of Fusion Energy in 1979. This program is focused on developing structural materials for fusion reactor first wall/breeding blanket components where the neutron damage is most severe. This area of a fusion reactor will be required to tolerate damage levels on the order of 110 dpa( 1 ). As a part of ADIP, study of the martensitic steels is focused on establishing the feasibility of using these materials. The interest in martensitic steels stems from their potential to tolerate high levels of radiation damage without significant degradation of material properties. Martensitic steels have a body-centered-cubic crystal structure that, unlike face-centered-cubic structure of austenitic steels, exhibits very little swelling under neutron irradiation( 2 ). One of the outstanding issues with martensitic steels is the possible parasitic stresses associated with ferromagnetic interaction with the magnetic fields. This paper is divided into two parts, the first reviews previous work on magnetic effects to the structure and plasma; the second presents new calculations of stresses on a coolant pipe in a Starfire model assumed to be made of 12Cr-1Mo steel(HT-9)

Fusion technologies for Laser Inertial Fusion Energy (LIFE∗

Directory of Open Access Journals (Sweden)

Kramer K.J.

2013-11-01

Full Text Available The Laser Inertial Fusion-based Energy (LIFE engine design builds upon on going progress at the National Ignition Facility (NIF and offers a near-term pathway to commercial fusion. Fusion technologies that are critical to success are reflected in the design of the first wall, blanket and tritium separation subsystems. The present work describes the LIFE engine-related components and technologies. LIFE utilizes a thermally robust indirect-drive target and a chamber fill gas. Coolant selection and a large chamber solid-angle coverage provide ample tritium breeding margin and high blanket gain. Target material selection eliminates the need for aggressive chamber clearing, while enabling recycling. Demonstrated tritium separation and storage technologies limit the site tritium inventory to attractive levels. These key technologies, along with the maintenance and advanced materials qualification program have been integrated into the LIFE delivery plan. This describes the development of components and subsystems, through prototyping and integration into a First Of A Kind power plant.

High stability space frame for a large fusion laser

International Nuclear Information System (INIS)

Hurley, C.A.; Myall, J.O.

1975-01-01

The Shiva laser system is a large neodymium glass laser target irradiation facility being constructed at LLL to perform laser fusion experiments. A frame is being constructed to support the large number of laser components that make up the Shiva system. Twenty laser chains composed of amplifiers, spatial filters, polarizers, rotators, and mirrors will be arranged in an optimum geometry so that each beam arrives at the target simultaneously and within alignment tolerances. This frame is capable of supporting approximately 600 individual component assemblies and maintaining a tolerance of +-4-μrad rotation between any two points over a period of 100 s. Consideration has been given to the positional stability and support of the components, the geometrical array of stacked beams with respect to the oscillator and target, the flow of utilities (e.g., power cables and cooling gas pipes), good accessibility for operation and maintenance, and adaptability for change and growth

IFMIF (International Fusion Materials Irradiation Facility) key element technology phase task description

Energy Technology Data Exchange (ETDEWEB)

Ida, M.; Nakamura, H.; Sugimoto, M.; Yutani, T.; Takeuchi, H. [eds.] [Japan Atomic Energy Research Inst., Tokai Research Establishment, Fusion Neutron Laboratory, Tokai, Ibaraki (Japan)

2000-08-01

In 2000, a 3 year Key Element technology Phase (KEP) of the International Fusion Materials Irradiation Facility (IFMIF) has been initiated to reduce the key technology risk factors needed to achieve continuous wave (CW) beam with the desired current and energy and to reach the corresponding power handling capabilities in the liquid lithium target system. In the KEP, the IFMIF team (EU, Japan, Russian Federation, US) will perform required tasks. The contents of the tasks are described in the task description sheet. As the KEP tasks, the IFMIF team have proposed 27 tasks for Test Facilities, 12 tasks for Target, 26 tasks for Accelerator and 18 tasks for Design Integration. The task description by RF is not yet available. The task items and task descriptions may be added or revised with the progress of KEP activities. These task description sheets have been compiled in this report. After 3 years KEP, the results of the KEP tasks will be reviewed. Following the KEP, 3 years Engineering Validation Phase (EVP) will continue for IFMIF construction. (author)

IFMIF (International Fusion Materials Irradiation Facility) conceptual design activity reduced cost report

International Nuclear Information System (INIS)

2000-02-01

This report describes the results of a preliminary reevaluation of the design and cost of the International Fusion Materials Irradiation Facility (IFMIF) Project in response to the request from the 28th FPCC meeting in January 1999. Two major ideas have been considered: 1) reduction of the total construction cost through elimination of the previously planned facility upgrade and 2) a facility deployment in 3 stages with capabilities for limited experiments in the first stage. As a result, the size and complexity of the facility could be significantly reduced, leading to substantial cost savings. In addition to these two ideas, this study also included a critical review of the original CDA specification with the objective of elimination of nonessential items. For example, the number of lithium targets was reduced from two to one. As a result of these changes in addition to the elimination of the upgrade, the total cost estimate was very substantially reduced from 797.2 MICF to 487.8 MICF, where 1 MICF = 1 Million of the IFMIF Conversion Units (approximately $1M US January, 1996). (author)

Overview of materials research for fusion reactors

International Nuclear Information System (INIS)

Muroga, T.; Gasparotto, M.; Zinkle, S.J.

2002-01-01

Materials research for fusion reactors is overviewed from Japanese, EU and US perspectives. Emphasis is placed on programs and strategies for developing blanket structural materials, and recent highlights in research and development for reduced activation ferritic martensitic steels, vanadium alloys and SiC/SiC composites, and in mechanistic experimental and modeling studies. The common critical issue for the candidate materials is the effect of irradiation with helium production. For the qualification of materials up to the full lifetime of a DEMO and Power Plant reactors, an intense neutron source with relevant fusion neutron spectra is crucial. Elaborate use of the presently available irradiation devices will facilitate efficient and sound materials development within the required time scale

5. IEA International workshop on beryllium technology for fusion. Book of abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

The collection includes the abstracts of reports presented to the 5-th IEA international workshop on beryllium technology for fusion. The themes of reports are as follows: status of beryllium technology for fusion in Russia; manufacturing and testing of Be armoured first wall mock-up for ITER; development of the process of diffusion welding of metals stainless steel-copper-beryllium into a single composite; some features of beryllium-laser beam interaction; the effect of irradiation dose on tritium and helium release from neutron irradiated beryllium; thermal properties of neutron irradiated Be{sub 12}Ti. The results of investigating the mechanical properties variation and swelling of beryllium under high temperature neutron irradiation are presented.

5. IEA International workshop on beryllium technology for fusion. Book of abstracts

International Nuclear Information System (INIS)

2001-01-01

The collection includes the abstracts of reports presented to the 5-th IEA international workshop on beryllium technology for fusion. The themes of reports are as follows: status of beryllium technology for fusion in Russia; manufacturing and testing of Be armoured first wall mock-up for ITER; development of the process of diffusion welding of metals stainless steel-copper-beryllium into a single composite; some features of beryllium-laser beam interaction; the effect of irradiation dose on tritium and helium release from neutron irradiated beryllium; thermal properties of neutron irradiated Be 12 Ti. The results of investigating the mechanical properties variation and swelling of beryllium under high temperature neutron irradiation are presented [ru

Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

Energy Technology Data Exchange (ETDEWEB)

Bourham, Mohamed A.; Gilligan, John G.

1999-08-14

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

Fusion technology programme

International Nuclear Information System (INIS)

Finken, D.

1985-10-01

KfK is involved in the European Fusion Programme predominantly in the NET and Fusion Technology part. The following fields of activity are covered: Studies for NET, alternative confinement concepts, and needs and issues of integral testing. Research on structural materials. Development of superconducting magnets. Gyrotron development (part of the Physics Programme). Nuclear technology (breeding materials, blanket design, tritium technology, safety and environmental aspects of fusion, remote maintenance). Reported here are status and results of work under contracts with the CEC within the NET and Technology Programme. The aim of the major part of this R and D work is the support of NET, some areas (e.g. materials, safety and environmental impact, blanket design) have a wider scope and address problems of a demonstration reactor. In the current working period, several new proposals have been elaborated to be implemented into the 85/89 Euratom Fusion Programme. New KfK contributions relate to materials research (dual beam and fast reactor irradiations, ferritic steels), to blanket engineering (MHD-effects) and to safety studies (e.g. magnet safety). (orig./GG)

FUSION Yearbook. Association Euratom-Tekes. Annual Report 2004

International Nuclear Information System (INIS)

Karttunen, S.; Rantamaeki, K.

2005-05-01

This report summarises the results of the Tekes FUSION technology programme and the fusion research activities by the Association Euratom-Tekes in 2004. The research areas are fusion physics, plasma engineering, fusion technology and a smaller effort to socioeconomic studies. Fusion technology research is carried out in close collaboration with Finnish industry. The emphasis in fusion physics and plasma engineering is in theoretical and computational studies on turbulent transport and modelling of radio-frequency heating experiments and the real time control of transport barriers in JET plasmas, predictive integrated modelling of tokamak plasmas, and studies on material transport in the edge plasmas supported by surface analysis of the JET divertor and limiter tiles. The work in fusion technology for the EFDA Technology Programme and ITER is strongly focused into vessel/in-vessel materials covering research and characterisation of first wall materials, mechanical testing of reactor materials under neutron irradiation, characterisation of irradiated Ti-alloys, simulations of carbon and tungsten sputtering, joining and welding methods and surface physics studies on plasma facing materials. A second domain of fusion technology consists of remote handling systems including water hydraulic manipulators for the ITER divertor maintenance as well as prototyping of intersector welding and cutting robot. Virtual modelling is an essential element in the remote handling engineering. Preparations to host the ITER divertor test platform (DTP2) were completed in 2004 and the DTP2 facility will be hosted by VTT. Some effort was also devoted to neutronics, socio-economic and power plant studies. Several EFDA technology tasks were successfully completed in 2004. (orig.)

Critical plasma-wall interaction issues for plasma-facing materials and components in near-term fusion devices

International Nuclear Information System (INIS)

Federici, G.; Coad, J.P.; Haasz, A.A.; Janeschitz, G.; Noda, N.; Philipps, V.; Roth, J.; Skinner, C.H.; Tivey, R.; Wu, C.H.

2000-01-01

The increase in pulse duration and cumulative run-time, together with the increase of the plasma energy content, will represent the largest changes in operation conditions in future fusion devices such as the International Thermonuclear Experimental Reactor (ITER) compared to today's experimental facilities. These will give rise to important plasma-physics effects and plasma-material interactions (PMIs) which are only partially observed and accessible in present-day experiments and will open new design, operation and safety issues. For the first time in fusion research, erosion and its consequences over many pulses (e.g., co-deposition and dust) may determine the operational schedule of a fusion device. This paper identifies the most critical issues arising from PMIs which represent key elements in the selection of materials, the design, and the optimisation of plasma-facing components (PFCs) for the first-wall and divertor. Significant advances in the knowledge base have been made recently, as part of the R and D supporting the engineering design activities (EDA) of ITER, and some of the most relevant data are reviewed here together with areas where further R and D work is urgently needed

Fusion materials semiannual progress report for period ending June 30, 1997

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-08-01

This is the twenty-second in a series of semiannual technical progress reports on fusion materials. This report combines the full spectrum of research and development activities on both metallic and non-metallic materials with primary emphasis on the effects of the neutronic and chemical environment on the properties and performance of materials for in-vessel components. This effort forms one element of the materials program being conducted in support of the Fusion Energy Sciences Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. Topics covered here are: vanadium alloys; silicon carbide composites; ferritic/martensitic steels; austenitic stainless steels; insulating ceramics and optical materials; solid breeding materials; radiation effects mechanistic studies and experimental methods; dosimetry damage parameters; activation calculations; materials engineering and design requirements; irradiation facilities; test matrices; and experimental methods.

Fusion materials semiannual progress report for period ending June 30, 1997

International Nuclear Information System (INIS)

1997-08-01

This is the twenty-second in a series of semiannual technical progress reports on fusion materials. This report combines the full spectrum of research and development activities on both metallic and non-metallic materials with primary emphasis on the effects of the neutronic and chemical environment on the properties and performance of materials for in-vessel components. This effort forms one element of the materials program being conducted in support of the Fusion Energy Sciences Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. Topics covered here are: vanadium alloys; silicon carbide composites; ferritic/martensitic steels; austenitic stainless steels; insulating ceramics and optical materials; solid breeding materials; radiation effects mechanistic studies and experimental methods; dosimetry damage parameters; activation calculations; materials engineering and design requirements; irradiation facilities; test matrices; and experimental methods

Fusion Programme SCK-CEN - Annual report 2009

International Nuclear Information System (INIS)

Massaut, V.

2009-01-01

This report summarizes the Research and Development work carried out at SCK-CEN on fusion technology in the year 2009. This covers mostly the work done under the EFDA agreement as well as the new developments carried out within the so-called Broader Approach of fusion such as - studies on structural and first wall materials for ITER and DEMO - studies and testing on the radiation resistance of instruments and componenets for the diagnostic and remote handling - development of irradiation devices and systems for the testing of fusion materials under representative environment.

Fusion Programme SCK-CEN - Annual report 2009

Energy Technology Data Exchange (ETDEWEB)

Massaut, V

2009-10-15

This report summarizes the Research and Development work carried out at SCK-CEN on fusion technology in the year 2009. This covers mostly the work done under the EFDA agreement as well as the new developments carried out within the so-called Broader Approach of fusion such as - studies on structural and first wall materials for ITER and DEMO - studies and testing on the radiation resistance of instruments and componenets for the diagnostic and remote handling - development of irradiation devices and systems for the testing of fusion materials under representative environment.

Failure rate data for fusion safety and risk assessment

International Nuclear Information System (INIS)

Cadwallader, L.C.

1993-01-01

The Fusion Safety Program (FSP) at the Idaho National Engineering Laboratory (INEL) conducts safety research in materials, chemical reactions, safety analysis, risk assessment, and in component research and development to support existing magnetic fusion experiments and also to promote safety in the design of future experiments. One of the areas of safety research is applying probabilistic risk assessment (PRA) methods to fusion experiments. To apply PRA, we need a fusion-relevant radiological dose code and a component failure rate data base. This paper describes the FSP effort to develop a failure rate data base for fusion-specific components

Dependence between fusion temperatures and chemical components of a certain type of coal using classical, non-parametric and bootstrap techniques

Energy Technology Data Exchange (ETDEWEB)

Gonzalez-Manteiga, W.; Prada-Sanchez, J.M.; Fiestras-Janeiro, M.G.; Garcia-Jurado, I. (Universidad de Santiago de Compostela, Santiago de Compostela (Spain). Dept. de Estadistica e Investigacion Operativa)

1990-11-01

A statistical study of the dependence between various critical fusion temperatures of a certain kind of coal and its chemical components is carried out. As well as using classical dependence techniques (multiple, stepwise and PLS regression, principal components, canonical correlation, etc.) together with the corresponding inference on the parameters of interest, non-parametric regression and bootstrap inference are also performed. 11 refs., 3 figs., 8 tabs.

Fusion-neutron effects on magnetoresistivity of copper stabilizer materials

Energy Technology Data Exchange (ETDEWEB)

Guinan, M.W.; Van Konynenburg, R.A.

1983-02-24

The objective of this work is to quantify the changes which occur in the magnetoresistivity of coppers (having various purities and pretreatments, and at magnetic fields up to 12 T during the course of sequential fusion neutron irradiations at about 4/sup 0/K and anneals to room temperature. In conjunction with work in progress by Coltman and Klabunde of ORNL, the results should lead to engineering design data for the stabilizers of superconducting magnets in fusion reactors. These magnets are expected to be irradiated during reactor operation and warmed to room temperature periodically during maintenance.

Study of damages by neutron irradiation in lithium aluminates

International Nuclear Information System (INIS)

Palacios G, O.

1999-01-01

Lithium aluminates proposed to the production of tritium in fusion nuclear reactors, due to the thermal stability that they present as well as the behavior of the aluminium to the irradiation. As a neutron flux with profile (≅ 14 Mev) of a fusion reactor is not available. A irradiation experiment was designed in order to know the micro and nano structure damages produced by fast and thermal neutrons in two irradiation positions of the fusion nuclear reactor Triga Mark III: CT (Thermal Column) and SIFCA (System of Irradiation Fixed of Capsules). In this work samples of lithium aluminate were characterized by XRD (X-Ray Diffraction), TEM (Transmission Electron Microscopy) and SEM (Scanning Electron Microscopy). Two samples were prepared by two methods: a) coalition method and b) peroxide method. This characterization comprised original and irradiated samples. The irradiated sample amounted to 4 in total: one for each preparation method and one for each irradiation position. The object of this analysis was to correlate with the received neutron dose the damages suffered by the samples with the neutron irradiation during long periods (440 H), in their micro and nano structure aspects; in order to understand the changes as a function of the irradiation zone (with thermal and fast neutron flux) and the preparation methods of the samples and having as an antecedent the irradiation in SIFCA position by short times (2h). The obtained results are referred to the stability of γ -aluminate phase, under given conditions of irradiation and defined nano structure arrangement. They also refer to the proposals of growth mechanism and nucleation of new phases. The error associated with the measurement of neutron dose is also discussed. (Author)

Progress in the US program to develop low-activation structural materials for fusion

International Nuclear Information System (INIS)

Kurtz, R.J.; Jones, R.H.; Bloom, E.E.; Rowcliffe, A.F.; Smith, D.L.; Odette, G.R.; Wiffen, F.W.

1999-01-01

It has long been recognized that attainment of the safety and environmental potential of fusion energy requires the successful development of low-activation materials for the first wall, blanket and other high heat flux structural components. Only a limited number of materials potentially possess the physical, mechanical and low-activation characteristics required for this application. The current US structural materials research effort is focused on three candidate materials: advanced ferritic steels, vanadium alloys, and silicon carbide composites. Recent progress has been made in understanding the response of these materials to neutron irradiation. (author)

Current status of reduced-activation ferritic/martensitic steels R and D for fusion energy

International Nuclear Information System (INIS)

Kimura, Akihiko

2005-01-01

Reduced-activation ferritic/martensitic (RAF/M) steels have been considered to be the prime candidate for the fusion blanket structural material. The irradiation data obtained up to now indicates rather high feasibility of the steels for application to fusion reactors because of their high resistance to degradation of material performance caused by both the irradiation-induced displacement damage and transmutation helium atoms. The martensitic structure of RAF/M steels consists of a large number of lattice defects before the irradiation, which strongly retards the formation of displacement damage through absorption and annihilation of the point defects generated by irradiation. Transmutation helium can be also trapped at those defects in the martensitic structure so that the growth of helium bubbles at grain boundaries is suppressed. The major properties of the steels are well within our knowledge, and processing technologies are mostly developed for fusion application. RAF/M steels are now certainly ready to proceed to the next stage, that is, the construction of International Thermo-nuclear Experimental Reactor Test Blanket Modules (ITER-TBM). Oxide dispersion strengthening (ODS) steels have been developed for higher thermal efficiency of fusion power plants. Recent irradiation experiments indicated that the steels were quite highly resistant to neutron irradiation embrittlement, showing hardening accompanied by no loss of ductility. High-Cr ODS steels whose chromium concentration was in the range from 14 to 19 mass% showed high resistance to corrosion in supercritical pressurized water. It is shown that the 14Cr-ODS steel is susceptible to neither hydrogen nor helium embrittlement. A combined utilization of ODS steels with RAF/M steels will be effective to realize fusion power early at a reasonable thermal efficiency. (author)

Evaluation of alternative methods of simulating asymmetric bulk heating in fusion reactor blanket/shield components

International Nuclear Information System (INIS)

Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Wadkins, R.P.; Wessol, D.E.

1981-10-01

As a part of Phase O, Test Program Element-II of the Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program, a study was conducted by EG and G Idaho, Inc., to identify, characterize, and recommend alternative approaches for simulating fusion bulk heating in blanket/shield components. This is the report on that effort. Since the usefulness of any simulation approach depends upon the particular experiment considered, classes of problem types (thermal-hydraulic, thermomechanical, etc.) and material types (structure, solid breeder, etc.) are developed. The evaluation of the various simulation approaches is performed for the various significant combinations of problem class and material class. The simulation approaches considered are discrete-source heating, direct resistance, electromagnetic induction, microwave heating, and nuclear heating. From the evaluations performed for each experiment type, discrete - source heating emerges as a good approach for bulk heating simulation in thermal - hydraulics experiments, and nuclear heating appears to be a good approach in experiments addressing thermomechanics and combined thermal-hydraulic/thermomechanics

GfW-handbook for data compilation of irradiation tested electronic components

International Nuclear Information System (INIS)

Wulf, F.; Braeunig, D.; Gaebler, W.

1981-06-01

The present 2. edition of the Data Compilation of Irradiation Tested Electronic Components represents a continuation of the 1. edition and is published as a loose-leaf handbook. In addition to the 190 reports provided in the 1. issue the present handbook contains further 44 test reports of currently used semiconductor devices in a comprehensive but easily to handle graphical and tabular presentation. Statistical values are given in order to facilitate the parts life time evaluation in a radiative environment. (orig.) [de

Workshop on beryllium for fusion applications. Proceedings. IEA Implementing Agreement for a Programme of Research and Development on Fusion Materials

International Nuclear Information System (INIS)

Dalle Donne, M.

1993-12-01

As shown by recent developments beryllium has become one of the most important materials in the development of fusion reactors. It is practically the only neutron multiplier available for blankets with ceramic breeder materials and can be used with liquid metal breeders as well. It is one of the most likely materials to be used on the surface of the first walls and of the divertor. The neutron irradiation behavior of beryllium in a fusion reactor is not well know. Beryllium was extensively irradiated about 25-40 years ago and has been used since then in material testing reactors as reflector. In the meantime, however, beryllium has been improved quite considerably. Today it is possible to obtain commercially beryllium which is much more isotropic and contains smaller ammounts of oxide. There are already indications that these new kinds of beryllium behave better under irradiation. (orig.)

Radiation damage related to fusion-reactor materials

International Nuclear Information System (INIS)

Brimhall, J.L.

1982-03-01

Three reasons why the fusion irradiation environment could produce different damage microstructure will be discussed: (1) the primary damage state induced by a higher energy primary knock-on-spectra; (2) increased helium generation due to high eta, α cross sections; and (3) pulsed operation of the irradiation environment. Examples of the microstructural damage caused by each of the above will be given

Re-weldability tests of irradiated 316L(N) stainless steel using laser welding technique

International Nuclear Information System (INIS)

Yamada, Hirokazu; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, George; Kohno, Wataru; Morishima, Yasuo

2002-01-01

SS316L(N)-IG is the candidate material for the in-vessel and ex-vessel components of fusion reactors such as ITER (International Thermonuclear Experimental Reactor). This paper describes a study on re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. The laser welding process is used for re-welding of the water cooling branch pipeline repairs. It is clarified that re-welding of SS316L(N)-IG irradiated up to about 0.2 dpa (3.3 appm He) can be carried out without a serious deterioration of tensile properties due to helium accumulation. Therefore, repair of the ITER blanket cooling pipes can be performed by the laser welding process

Re-weldability tests of irradiated 316L(N) stainless steel using laser welding technique

Science.gov (United States)

Yamada, Hirokazu; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, George; Kohno, Wataru; Morishima, Yasuo

2002-12-01

SS316L(N)-IG is the candidate material for the in-vessel and ex-vessel components of fusion reactors such as ITER (International Thermonuclear Experimental Reactor). This paper describes a study on re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. The laser welding process is used for re-welding of the water cooling branch pipeline repairs. It is clarified that re-welding of SS316L(N)-IG irradiated up to about 0.2 dpa (3.3 appm He) can be carried out without a serious deterioration of tensile properties due to helium accumulation. Therefore, repair of the ITER blanket cooling pipes can be performed by the laser welding process.

23. Symposium On Fusion Technology (SOFT), Venice - A personal view

International Nuclear Information System (INIS)

Spears, W.R.

2004-01-01

This conference, examining the advances in our leading-edge technology, took place on 22-24 September 2004 against the wonderful and historic backdrop of Venice, at a monastery of the Cini Foundation, on the Island of St. Giorgio, directly opposite St. Marks. The strong connection between the ancient and modern was brought home to us in the very first talk, from the Mayor of Venice and MEP Prof. P. Costa, who reminded us of Venice's particular problem with global warming, and urged us to do our part to develop an energy source that should help to avoid it drowning. Prof. Sir C. Llewellyn-Smith, head of the UK Fusion Programme and Chairman of Euratom CCE-FU, took up this theme and elaborated how we should reach our goal, showing in particular the urgency of pursuing a fast track, proceeding with ITER and the International Fusion Materials Irradiation Facility (IFMIF) without further delay, and envisaging that the subsequent machine would be prototypical of future commercial fusion power plants. The conference proceeded through plenary and oral sessions, and through poster sessions, covering plasma heating, fuelling, control and diagnostics, magnets and power supplies, plasma-facing components, blanket and vessel, remote handling, materials technology, the experiences gained on existing experiments, and projections for future experiments and fusion power plants. There were 570 participants, from 25 countries, of whom a third came from outside Europe

The European fusion program and the role of the research reactors

International Nuclear Information System (INIS)

Laesser, R.; Andreani, R.; Diegele, E.

2005-01-01

The main objectives of the European long-term Fusion Technology Program are i) investigation of DEMO breeding blankets options, ii) development of low activation materials resistant to high neutron fluence, iii) construction of IFMIF for validation of DEMO materials, and iv) promotion of modelling efforts for the understanding of radiation damage. A large effort is required for the development and performance verification of the materials subjected to the intense neutron irradiation encountered in fusion reactors. In the absence of a strong 14.1 MeV neutron source fission materials research reactors are used. Elaborate in-pile and post-irradiation examinations are performed. In addition, the modelling effort is increased to predict the damage by a 'true' fusion spectrum in the future. Even assuming that a positive decision for IFMIF construction can be reached, the operation of a limited number of materials test reactors is needed to perform irradiation studies on large samples and for screening. (author)

Investigation of irradiation effects on highly integrated leading-edge electronic components of diagnostics and control systems for LHD deuterium operation

Science.gov (United States)

Ogawa, K.; Nishitani, T.; Isobe, M.; Murata, I.; Hatano, Y.; Matsuyama, S.; Nakanishi, H.; Mukai, K.; Sato, M.; Yokota, M.; Kobuchi, T.; Nishimura, T.; Osakabe, M.

2017-08-01

High-temperature and high-density plasmas are achieved by means of real-time control, fast diagnostic, and high-power heating systems. Those systems are precisely controlled via highly integrated electronic components, but can be seriously affected by radiation damage. Therefore, the effects of irradiation on currently used electronic components should be investigated for the control and measurement of Large Helical Device (LHD) deuterium plasmas. For the precise estimation of the radiation field in the LHD torus hall, the MCNP6 code is used with the cross-section library ENDF B-VI. The geometry is modeled on the computer-aided design. The dose on silicon, which is a major ingredient of electronic components, over nine years of LHD deuterium operation shows that the gamma-ray contribution is dominant. Neutron irradiation tests were performed in the OKTAVIAN at Osaka University and the Fast Neutron Laboratory at Tohoku University. Gamma-ray irradiation tests were performed at the Nagoya University Cobalt-60 irradiation facility. We found that there are ethernet connection failures of programmable logic controller (PLC) modules due to neutron irradiation with a neutron flux of 3  ×  106 cm-2 s-1. This neutron flux is equivalent to that expected at basement level in the LHD torus hall without a neutron shield. Most modules of the PLC are broken around a gamma-ray dose of 100 Gy. This is comparable with the dose in the LHD torus hall over nine years. If we consider the dose only, these components may survive more than nine years. For the safety of the LHD operation, the electronic components in the torus hall have been rearranged.

New facility for post irradiation examination of neutron irradiated beryllium

International Nuclear Information System (INIS)

Ishitsuka, Etsuo; Kawamura, Hiroshi

1995-01-01

Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800 degrees C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and 60 Co;7.4 MBq/day

RTNS-II irradiations and operations

International Nuclear Information System (INIS)

Logan, C.M.; Heikkinen, D.W.

1982-01-01

The objectives of this work are operation of RTNS-II (a 14-MeV neutron source facility), machine development, and support of the experimental program that utilizes this facility. Experimenter services include dosimetry handling, scheduling, coordination, and reporting. RTNS-II is dedicated to materials research for the fusion power program. Its primary use is to aid in the development of models of high-energy neutron effects. Such models are needed in interpreting and projecting to the fusion environment engineering data obtained in other neutron spectra. Irradiations were performed for a total of twenty-nine different experimenters during this quarter. A JOEL 200 CX TEM and other post-irradiation test equipment have been installed

RTNS-II fusion materials irradiation facility

International Nuclear Information System (INIS)

Heikkinen, D.W.; Tuckerman, D.B.; Davis, J.C.; Massoletti, D.J.; Short, D.W.

1986-01-01

The Rotating Target Neutron Source (RTNS-II) facility provides an intense source of 14-MeV neutrons for the fusion energy programs of Japan and the United States. Each of the two identical accelerator-based neutron sources is capable of providing source strengths in excess of 3 x 10 13 n/s using deuteron beam currents up to 150 mA. The present status of the facility, as well as the various upgrade options, will be described in detail

Polyploidization of rat hepatocytes due to cell fusion under the effect of radiation of different let

International Nuclear Information System (INIS)

Gil'yano, N.Ya.; Malinovskij, O.V.; Khair, M.B.; Baldychev, A.S.; Smolin, V.A.

1988-01-01

The method of flow cytometry was used to study polyploidization of hepatocytes following X-, γ-, and neutron-irradiation. Ionizing radiation was shown to induce cell polyploidization by two different ways: (1) cells and nuclei fusion, and (2) restriction of mitosis after DNA replication. RBE of 14 MeV neutrons with respect to fusion was about 5x10 3 . With neutron irradiation, the densitivity of cells by fusion was not lower than that by chromosome mutations

Irradiation effects in tungsten-copper laminate composite

Energy Technology Data Exchange (ETDEWEB)

Garrison, L.M., E-mail: garrisonlm@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Katoh, Y. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, L.L. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Byun, T.S. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Reiser, J.; Rieth, M. [Karlsruhe Institute of Technology, Karlsruhe (Germany)

2016-12-01

Tungsten-copper laminate composite has shown promise as a structural plasma-facing component as compared to tungsten rod or plate. The present study evaluated the tungsten-copper composite after irradiation in the High Flux Isotope Reactor (HFIR) at temperatures of 410–780 °C and fast neutron fluences of 0.02–9.0 × 10{sup 25} n/m{sup 2}, E > 0.1 MeV, 0.0039–1.76 displacements per atom (dpa) in tungsten. Tensile tests were performed on the composites, and the fracture surfaces were analyzed with scanning electron microscopy. Before irradiation, the tungsten layers had brittle cleavage failure, but the overall composite had 15.5% elongation at 22 °C. After only 0.0039 dpa this was reduced to 7.7% elongation, and no ductility was observed after 0.2 dpa at all irradiation temperatures when tensile tested at 22 °C. For elevated temperature tensile tests after irradiation, the composite only had ductile failure at temperatures where the tungsten was delaminating or ductile. - Highlights: • Fusion reactors need a tough, ductile tungsten plasma-facing material. • The unirradiated tungsten-copper laminate is more ductile than tungsten alone. • After neutron irradiation, the composite has significantly less ductility. • The tungsten behavior appears to dominate the overall composite behavior.

Effect of ultraviolet irradiation on chromatin and its components from Yoshida ascites tumour cells

International Nuclear Information System (INIS)

Ramakrishnan, N.; Patil, M.S.; Pradhan, D.S.

1981-01-01

A study has been made of the effect of U.V. irradiation on Yoshida ascites tumour chromatin and its non-DNA components. The extractability of total histones was increased from 6% to 17% with an increase in U.V. incident radiation dose from 500J/m 2 to 2000J/m 2 . The polyacrylamide gel electrophoresis pattern of chromosomal proteins was examined after irradiation of the chromatin, and the effect of U.V. irradiation of chromatin on histones was also investigated. The results indicated that cross-linking of DNA with chromosomal proteins is an important category of U.V. radiation-induced lesions discerned in U.V. irradiated chromatin. Histones and several non-histone proteins seemed to undergo U.V. radiation-induced cross-linking with DNA, which was taken as indicative of their close association with DNA in the chromatin structure. It is suggested that the cross-link formation between DNA and non-histone proteins may be due to sequence-specific association of non-histone proteins with DNA. (U.K.)

Conceptual design of inertial confinement fusion power plant

International Nuclear Information System (INIS)

Mima, Kunioki; Yamanaka, Tatsuhiko; Nakai, Sadao

1994-01-01

Presented is the status of the conceptual design studies of inertial confinement fusion reactors. The recent achievements of the laser fusion research enable us to refine the conceptual design of the power plant. In the paper, main features of several new conceptual designs of ICF reactor; KOYO, SIRIUS-P, HYLIFE-II and so on are summarized. In particular, the target design and the reactor chamber design are described. Finally, the overview of the laser fusion reactor and the irradiation system is also described. (author)

Swelling and swelling resistance possibilities of austenitic stainless steels in fusion reactors

International Nuclear Information System (INIS)

Maziasz, P.J.

1983-01-01

Fusion reactor helium generation rates in stainless steels are intermediate to those found in EBR-II and HFIR, and swelling in fusion reactors may differ from the fission swelling behavior. Advanced titanium-modified austenitic stainless steels exhibit much better void swelling resistance than AISI 316 under EBR-II (up to approx. 120 dpa) and HFIR (up to approx. 44 dpa) irradiations. The stability of fine titanium carbide (MC) precipitates plays an important role in void swelling resistance for the cold-worked titanium-modified steels irradiated in EBR-II. Futhermore, increased helium generation in these steels can (a) suppress void conversion, (b) suppress radiation-induced solute segregation (RIS), and (c) stabilize fine MC particles, if sufficient bubble nucleation occurs early in the irradation. The combined effects of helium-enhanced MC stability and helium-suppressed RIS suggest better void swelling resistance in these steels for fusion service than under EBR-II irradiation

Transient temperature response of in-vessel components due to pulsed operation in tokamak fusion experimental reactor (FER)

International Nuclear Information System (INIS)

Minato, Akio; Tone, Tatsuzo

1985-12-01

A transient temperature response of the in-vessel components (first wall, blanket, divertor/limiter and shielding) surrounding plasma in Tokamak Fusion Experimental Reactor (FER) has been analysed. Transient heat load during start up/shut down and pulsed operation cycles causes the transient temperature response in those components. The fatigue lifetime of those components significantly depends upon the resulting cyclic thermal stress. The burn time affects the temperature control in the solid breeder (Li 2 O) and also affects the thermo-mechanical design of the blanket and shielding which are constructed with thick structure. In this report, results of the transient temperature response obtained by the heat transfer and conduction analyses for various pulsed operation scenarios (start up, shut down, burn and dwell times) have been investigated in view of thermo-mechanical design of the in-vessel components. (author)

Repair welding of fusion reactor components

International Nuclear Information System (INIS)

Chin, B.A.

1993-01-01

Experiments have shown that irradiated Type 316 stainless steel is susceptible to heat-affected-zone (HAZ) cracking upon cooling when welded using the gas tungsten arc (GTA) process under lateral constraint. The cracking has been hypothesized to be caused by stress-assisted helium bubble growth and rupture at grain boundaries. This study utilized an experimental welding setup which enabled different compressive stresses to be applied to the plates during welding. Autogenous GTA welds were produced in Type 316 stainless steel doped with 256 appm helium. The application of a compressive stress, 55 MPa, during welding suppressed the previously observed catastrophic cracking. Detailed examinations conducted after welding showed a dramatic change in helium bubble morphology. Grain boundary bubble growth along directions parallel to the weld was suppressed. Results suggest that stress-modified welding techniques may be used to suppress or eliminate helium-induced cracking during joining of irradiated materials

Fusion Simulation Project Workshop Report

Science.gov (United States)

Kritz, Arnold; Keyes, David

2009-03-01

The mission of the Fusion Simulation Project is to develop a predictive capability for the integrated modeling of magnetically confined plasmas. This FSP report adds to the previous activities that defined an approach to integrated modeling in magnetic fusion. These previous activities included a Fusion Energy Sciences Advisory Committee panel that was charged to study integrated simulation in 2002. The report of that panel [Journal of Fusion Energy 20, 135 (2001)] recommended the prompt initiation of a Fusion Simulation Project. In 2003, the Office of Fusion Energy Sciences formed a steering committee that developed a project vision, roadmap, and governance concepts [Journal of Fusion Energy 23, 1 (2004)]. The current FSP planning effort involved 46 physicists, applied mathematicians and computer scientists, from 21 institutions, formed into four panels and a coordinating committee. These panels were constituted to consider: Status of Physics Components, Required Computational and Applied Mathematics Tools, Integration and Management of Code Components, and Project Structure and Management. The ideas, reported here, are the products of these panels, working together over several months and culminating in a 3-day workshop in May 2007.

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.)

Creep fatigue assessment for EUROFER components

Energy Technology Data Exchange (ETDEWEB)

Özkan, Furkan, E-mail: oezkan.furkan@partner.kit.edu; Aktaa, Jarir

2015-11-15

Highlights: • Design rules for creep fatigue assessment are developed to EUROFER components. • Creep fatigue assessment tool is developed in FORTRAN code with coupling MAPDL. • Durability of the HCPB-TBM design is discussed under typical fusion reactor loads. - Abstract: Creep-fatigue of test blanket module (TBM) components built from EUROFER is evaluated based on the elastic analysis approach in ASME Boiler Pressure Vessel Code (BPVC). The required allowable number of cycles design fatigue curve and stress-to-rupture curve to estimate the creep-fatigue damage are used from the literature. Local stress, strain and temperature inputs for the analysis of creep-fatigue damage are delivered by the finite element code ANSYS utilizing the Mechanical ANSYS Parametric Design Language (MAPDL). A developed external FORTRAN code used as a post processor is coupled with MAPDL. Influences of different pulse durations (hold-times) and irradiation on creep-fatigue damage for the preliminary design of the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) are discussed for the First Wall component of the TBM box.

Final Report: Safety of Plasma-Facing Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

International Nuclear Information System (INIS)

Bourham, Mohamed A.; Gilligan, John G.

1999-01-01

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m 2 over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER

Interaction of X-irradiated mouse cells in heterokaryons

International Nuclear Information System (INIS)

Hofmanova, J.; Spurna, V.

1985-01-01

The frequency of heterokaryon formation and the ability of DNA synthesis in the system of mouse X-irradiated L fibroblasts and non-irradiated or irradiated LS/BL lymphosarcoma cells were studied. The frequency of heterokaryons after fusion of one or both irradiated parental cells was 3 to 6 times higher than in the non-irradiated cell cultures. In these heterokaryons we found 1.5 to 3 times more nuclei of irradiated L cells capable of DNA synthesis than in the population of non-fused irradiated cells. (author)

Time-dependent tritium inventories and flow rates in fuel cycle components of a tokamak fusion reactor

International Nuclear Information System (INIS)

Kuan, W.

1995-01-01

Time-dependent inventories and flow rates for several components of the fuel cycle are modeled and studied through the use of a new modular-type model for the dynamic simulation of the fuel cycle in a fusion reactor. The complex dynamic behavior in the modeled subsystems is analyzed using this new model. Preliminary results using fuel cycle design configurations similar to ITER are presented and analyzed. The inventories and flow rates inside the primary vacuum pumping, fuel cleanup unit and isotope separation system are studied. Ways to minimize the tritium inventory are also assessed. This was performed by looking at various design options that could be used to minimize tritium inventory for specific components. (orig.)

Laser-induced fusion of human embryonic stem cells with optical tweezers

Energy Technology Data Exchange (ETDEWEB)

Chen Shuxun; Wang Xiaolin; Sun Dong [Department of Mechanical and Biomedical Engineering, City University of Hong Kong (Hong Kong); Cheng Jinping; Han Cheng, Shuk [Department of Biology and Chemistry, City University of Hong Kong (Hong Kong); Kong, Chi-Wing [Stem Cell and Regenerative Medicine Consortium, and Departments of Medicine and Physiology, LKS Faculty of Medicine, University of Hong Kong (Hong Kong); Li, Ronald A. [Stem Cell and Regenerative Medicine Consortium, and Departments of Medicine and Physiology, LKS Faculty of Medicine, University of Hong Kong (Hong Kong); Center of Cardiovascular Research, Mount Sinai School of Medicine, New York, New York 10029 (United States)

2013-07-15

We report a study on the laser-induced fusion of human embryonic stem cells (hESCs) at the single-cell level. Cells were manipulated by optical tweezers and fused under irradiation with pulsed UV laser at 355 nm. Successful fusion was indicated by green fluorescence protein transfer. The influence of laser pulse energy on the fusion efficiency was investigated. The fused products were viable as gauged by live cell staining. Successful fusion of hESCs with somatic cells was also demonstrated. The reported fusion outcome may facilitate studies of cell differentiation, maturation, and reprogramming.