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

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

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

Fractographic examination of HT-9 and 9Cr-1Mo Charpy specimens irradiated in the AD-2 test

International Nuclear Information System (INIS)

Gelles, D.S.; Hu, W.L.

1983-01-01

Fracture surface topologies have been examined using scanning electron microscopy for 20 selected half sized Charpy impact specimens of HT-9 and Modified 9Cr-1Mo in order to provide improved understanding of fracture toughness degradation as a result of irradiation for Path E alloys. The specimen matrix included unirradiated specimens and specimens irradiated in EBR-II in the AD-2 experiment. Also, hardness measurements have been made on selected irradiated Charpy specimens. The results of examinations indicate that irradiation hardening due to G-phase formation at 390 0 C is responsible for the large shift in ductile-to-brittle transition temperature (DBTT) found in HT-9. Toughness degradation in HT-9 observed following higher temperature irradiations is attributed to precipitation at delta ferrite stringers. Reductions in toughness as a consequence of irradiation in Modified 9Cr-1Mo are attributed to in-reactor precipitation of (V,Nb)C and M 23 C 6 . It is shown that crack propagation rates for ductile and brittle failure modes can be measured, that they differ by over an order of magnitude and that unexpected multiple shifts in fracture mode from ductile to brittle failure can be attributed to the effect of delta ferrite stringers on crack propagation rates

System for the continuous irradiation of specimens, especially for activation analysis

International Nuclear Information System (INIS)

Dieck, L.E.

1975-01-01

The system is to ensure a continuous irradiation of several specimens, especially for activation analysis. The specimens rotate in a hollow body which is axially movable along a rail. The rotation is effected by a turbine driven by hydraulic or pneumatic power and placed in the hollow body which can be used for example in the rabbit system of a nuclear reactor. The driving medium serves both as conveying medium for the system and drive for the rotating drum and as coolant for the specimens and results in radiation protection. The geometric arrangement and design of both the turbine and the whole system is decribed in detail. (UWI) [de

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-11-01

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

Beryllium irradiation embrittlement test programme. Material and specimen specification, manufacture and qualification

International Nuclear Information System (INIS)

Harries, D.R.; Dalle Donne, M.

1996-06-01

The report presents the specification, manufacture and qualification of the beryllium specimens to be irradiated in the BR2 reactor in Mol to investigate the effect of the neutron irradiation on the embrittlement as a function of temperature and beryllium oxide content. This work was been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhe and is supported by the European Union within the European Fusion Technology Program. (orig.)

Summary of the U.S. specimen matrix for the HFIR 13J varying temperature irradiation capsule

International Nuclear Information System (INIS)

Zinkle, S.J.

1998-01-01

The US specimen matrix for the collaborative DOE/Monbusho HFIR 13J varying temperature irradiation capsule contains two ceramics and 29 different metals, including vanadium alloys, ferritic/martensitic steels, pure iron, austenitic stainless steels, nickel alloys, and copper alloys. This experiment is designed to provide fundamental information on the effects of brief low-temperature excursions on the tensile properties and microstructural evolution of a wide range of materials irradiated at nominal temperatures of 350 and 500 C to a dose of ∼5 dpa. A total of 340 miniature sheet tensile specimens and 274 TEM disks are included in the US-supplied matrix for the irradiation capsule

Predicting crack instability behavior of burst tests from small specimens for irradiated Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

Davies, P.H.

1997-01-01

A scaling approach, based on the deformation J-integral at maximum load obtained from small specimens, is proposed for predicting the crack instability behavior of burst tests on irradiated Zr-2.5Nb pressure tubes. An assessment of this approach is carried out by comparison with other toughness criteria such as the modified J-integral and the plastic work dissipation rate approach. The largest discrepancy between the different parameters occurs for materials of intermediate toughness which exhibit the most stable crack growth and tunnelling up to maximum load. A study of one material of intermediate toughness suggests crack-front tunnelling has a significant influence on the results obtained from the 17-mm-wide specimens. It is shown that for a tube of intermediate toughness the different approaches can significantly underpredict the extent of stable crack growth before instability in a burst test even after correcting for tunnelling. The usefulness of a scaling approach in reducing the discrepancy between the small- and large-scale specimen results for this material is demonstrated

Irradiation of UO2 specimens with molten cores in a pressurized water loop. Test X-2-x

International Nuclear Information System (INIS)

Bain, A.S.

1961-08-01

Two Zircaloy-2 clad specimens containing stoichiometric UO 2 pellets were irradiated in a pressurized water loop for 379 hours at heat ratings sufficient to cause central melting of the UO 2 . There was no appearance of localized overheating or accelerated corrosion of the sheath, but the diametral increases were considerably larger than those observed in loop specimens irradiated at lower heat ratings. The length increases, however, were approximately the same as those measured for specimens at lower ratings. There was a clearly visible demarcation between UO 2 that had been molten and that which had not. The value of ∫ 500 o C Tm kdθ = 74 ± W/cm was essentially the same as that obtained from the short-duration tests in the Hydraulic Rabbit, indicating there is no marked decrease in thermal conductivity of the UO 2 fuel in irradiations up to 379 hours. (author)

Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high

Irradiation temperature measurement of the reactor pressure vessel surveillance specimen in the programmes of radiation degradation monitoring

International Nuclear Information System (INIS)

Kupca, L.; Stanc, S.; Simor, S.

2001-01-01

The information's about the special system of irradiation temperature measurement, used for reactor pressure vessel surveillance specimen, which are placed in reactor thermal shielding canals are presented in the paper. The system was designed and realized in the frame of Extended Surveillance Specimen Programme for NPP V-2 Jaslovske Bohunice and Modern Surveillance Specimen Programme for NPP Mochovce. Base design aspects, technical parameters of realization and results of measurement on the two units in Bohunice and Mochovce NPPs are presented too. (Authors)

Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

International Nuclear Information System (INIS)

Gerstenberg, H.; Kraehling, E.; Katheder, H.

1997-01-01

The shear strengths of various fibre reinforced resins being promising candidate insulators for superconducting coils to be used tinder a strong radiation load, e.g. in future fusion reactors were investigated prior and subsequent to reactor in-core irradiation at liquid helium temperature. A large number of sandwich-like (steel-bonded insulation-steel) specimens representing a widespread variety of materials and preparation techniques was exposed to irradiation doses of up to 5 x 10 7 Gy in form of fast neutrons and γ-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination of the ultimate shear strength. The results obtained from the different tested materials are compared. In addition an upgrade of the in-situ test rig installed at the Munich research reactor is presented, which allows combined shear/compression loading of low temperature irradiated specimens and provides a doubling of the testing rate

Analysis of ductile-brittle transition shifts for standard and miniature bending specimens of irradiated steel

International Nuclear Information System (INIS)

Korshunov, M.E.; Korolev, Yu.N.; Krasikov, E.A.; Gabuev, N.N.; Tykhmeev, D.Yu.

1996-01-01

A study is made to reveal if there is a correlation between shifts in temperature curves obtained when testing thin plates and standard specimens on impact bending and fracture toughness. The tests were carried out using steel 25Kh3NM specimens irradiated by 6 x 10 19 cm -2 neutron fluence. A conclusion is made about the possibility to evaluate the degree of radiation-induced embrittlement of reactor steels on the basis of thin plate testing under quasistatic loads [ru

Design, Fabrication and Test Report on a Verification Capsule (05M-06K) for the Control of a Neutron Irradiation Fluence of Specimens in HANARO

Energy Technology Data Exchange (ETDEWEB)

Choo, K. N.; Kim, B. G.; Kang, Y. H.; Cho, M. S.; Son, J. M.; Shin, Y. T.; Park, S. J.; Choi, M. H.; Lee, D. S.

2007-02-15

As a part of a project for a capsule development and utilization for an irradiation test, a verification capsule (05M-06K) was designed, fabricated and tested for the development of new instrumented capsule technology for a more precise control of the irradiation fluence of a specimen, irrespective of the reactor operation condition. The basic structure of the 05M-06K capsule was based on the 04M-22K mock-up capsule which was successfully designed and out-pile tested to confirm the various key technologies necessary for the fluence control of a specimen. 21 square and round shaped specimens made of STS 304 were inserted into the capsule. The capsule was constructed in 5 stages with specimens and an independent electric heater at each stage. Each of the five specimens which were accommodated in the 1st stage (top) of the capsule can be taken out of the HANARO core during a normal reactor operation. The specimen is extracted by a specimen extraction mechanism using a steel wire. During the out-pile test, the temperatures of the specimens were measured by 12 thermocouples installed in the capsule. The capsule was successfully out-pile tested in a single channel test loop. The obtained results will be used for a safety evaluation of the new irradiation capsule for controlling the irradiation fluence of specimens in HANARO.

FFTF irradiation of fracture mechanics specimens for out-of-core structures

International Nuclear Information System (INIS)

King, D.C.

1978-09-01

The National Program Plan has established data requirements for out-of-core structures for FBRs. Significant FFTF irradiation space with moderate gamma heating levels is required to irradiate relatively large fracture mechanics specimens to total neutron fluences ranging between 5 x 10 21 and 5 x 10 22 n/cm 2 and temperatures which range between 400 0 C (750 0 F) and 650 0 C (1200 0 F). Priority 1 data on stainless steel welds requires a test volume of 7443 cm 3 (454 in 3 ). Priority 2 data on 304 and 316 SS and Inconel 718 materials and Inconel 718 welds requires 2760 cm 3 (168 in 3 ). Priority 3 data on stainless steels, other nickel-base alloys, and ferritics requires 33,118 cm 3 (2021 in 3 ). Priority 4 data at elevated temperatures on stainless steels, other nickel-base alloys and ferritics requires 69,182 cm 3

Analysis of stress-induced Burgers vector anisotropy in pressurized tube specimens of irradiated ferritic-martensitic steel: JLF-1

International Nuclear Information System (INIS)

Gelles, D.S.; Shibayama, T.

1998-01-01

A procedure for determining the Burgers vector anisotropy in irradiated ferritic steels allowing identification of all a and all a/2 dislocations in a region of interest is applied to a pressurized tube specimen of JLF-1 irradiated at 430 C to 14.3 x 10 22 n/cm 2 (E > 0.1 MeV) or 61 dpa. Analysis of micrographs indicates large anisotropy in Burgers vector populations develop during irradiation creep

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-07-01

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

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

International Nuclear Information System (INIS)

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

1993-07-01

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

In-situ tritium recovery from Li2O irradiated in fast neutron flux - Beatrix-II temperature change specimen

International Nuclear Information System (INIS)

Slagle, O.D.; Hollenberg, G.W.; Kurasawa, T.; Verrall, R.A.

1992-01-01

The Beatrix-II irradiation experiment is an in-situ tritium release experiment to evaluate the stability and tritium release characteristics of Li 2 O under fast neutron irradiation to extended burnups. A thin annular ring specimen capable of temperature changes was irradiated in Phase I of the experiment to a lithium burnup of 5%. The primary emphasis of the test plan was to determine the effect and interrelationship of gas composition and temperature on the tritium inventory with increasing temperature and a series of specific temperature changes were carried out at intervals throughout the experiment to characterize the effect of burnup. Decreasing the amount of hydrogen in the sweep gas resulted in an increase in the tritium inventory in the Li 2 O specimen. The tritium recovery during startup and shutdown was observed to be strongly influenced by the composition of the sweep gas

Effect of neutron irradiation on vanadium alloys

International Nuclear Information System (INIS)

Braski, D.N.

1986-01-01

Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600 0 C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520 0 C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys

Effect of neutron irradiation on vanadium alloys

Energy Technology Data Exchange (ETDEWEB)

Braski, D.N.

1986-01-01

Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

Accelerated irradiation growth of zirconium alloys

International Nuclear Information System (INIS)

Griffiths, M.; Gilbert, R.W.; Fidleris, V.

1989-01-01

This paper discusses how sponge zirconium and Zr-2.5 wt% Nb, Zircaloy, or Excel alloys all exhibit accelerated irradiation growth compared with high-purity crystal-bar zirconium for irradiation temperatures between 550 to 710 K and fluences between 0.1 to 10 x 10 25 n · m -2 (E > 1 MeV). There is generally an incubation period or fluence before the onset of accelerated or breakaway growth, which is dependent on the particular material being irradiated, its metallurgical condition before irradiation, and the irradiation temperature. Transmission electron microscopy has shown that there is a correlation between accelerated irradiation growth and the appearance of c-component vacancy loops on basal planes. Measurements in some specimens indicate that a significant fraction of the strain can be directly attributed to the loops themselves. There is considerable evidence to show that their formation is dependent both on the specimen purity and on the irradiation temperature. Materials that have a high interstitial-solute content contain c-component loops and exhibit high growth rates even at low fluences ( 2 :5 n · m -2 , E > 1 MeV). For sponge zirconium and the Zircaloys, c-component loop formation and the associated acceleration of growth (breakaway) during irradiation occurs because the intrinsic interstitial solute (mainly, oxygen, carbon and nitrogen) in the zirconium matrix is supplemented by interstitial iron, chromium, and nickel from the radiation-induced dissolution of precipitates. (author)

Gamma irradiation test report of simulated grout specimens for gas generation/liquid advection

International Nuclear Information System (INIS)

Hinman, C.A.

1994-01-01

This report presents the results from an irradiation test performed on four specimens of grout that were fabricated from synthetic Double Shell Slurry Feed (DSSF) liquid waste. The objective was to investigate the radiolytic generation of gases and the potential for advective rejection of waste liquids from the grout matrix and to provide experimental information for the validation of the C-Cubed calculated model. It has been demonstrated that a number of gases can be formed within the grout due to radiolytic decomposition of various chemical components that make up the grout. This observation leads to the conjecture that the potential exists for the rejection of a portion of the 60 vol% free liquid from the grout matrix driven by pressurization by these gases. It was found that, for the specimen geometries used in this test series, and for peak radiation dose accumulation rates on the order of 4 to 60 times of the initial rate expected in the grout vaults (300 Rads/hr), no liquid rejection was observed from 2% to 35% of the target exposure expected in the grout vaults (1E+08 Rads). When the irradiation rate exceeded the projected grout vault dose rate by a factor of 200 a small amount of liquid rejection was observed from one of two specimens that had received 20% more than the goal exposure. Because of the differences in the magnitudes of the relative radiation field strengths between this study and an actual grout vault, it is concluded that the potential for liquid rejection by internal gas pressurization from presently configured grout waste forms is very low for the expected conditions

Ductile fracture toughness of heavy section pressure vessel steel plate. A specimen-size study of ASTM A 533 steels

International Nuclear Information System (INIS)

Williams, J.A.

1979-09-01

The ductile fracture toughness, J/sub Ic/, of ASTM A 533, Grade B, Class 1 and ASTM A 533, heat treated to simulate irradiation, was determined for 10- to 100-mm thick compact specimens. The toughness at maximum specimen load was also measured to determine the conservatism of J/sub Ic/. The toughness of ASTM A 533, Grade B, Class 1 steel was 349 kJ/m 2 and at the equivalent upper shelf temperature, the heat treated material exhibited 87 kJ/m 2 . The maximum load fracture toughness was found to be linearly proportional to specimen size, and only specimens which failed to meet ASTM size criteria exhibited maximum load toughness less than J/sub Ic/

Development of reconstitution technique of irradiated specimens. 3. Report for FY 1995 and FY 1996 on JAERI-IHI cooperated research program (joint research)

Energy Technology Data Exchange (ETDEWEB)

Nishiyama, Yutaka; Fukaya, Kiyoshi; Onizawa, Kunio; Suzuki, Masahide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakamura, Terumi; Kaihara, Shoichiro; Yoshida, Kazuo; Sato, Akira

1998-10-01

The cooperated research between Japan Atomic Energy Research Institute and Ishikawajima-Harima Heavy Industries Co., Ltd. on the development of reconstitution technique of irradiated reactor pressure vessel surveillance specimens has been performed from FY 1993. In FY 1993-1994, the method of surface activated joining (SAJ) was applied to reconstitution of Charpy impact specimens. Some verification tests using unirradiated reactor pressure vessel plate materials have shown that SAJ is feasible for a reconstitution technique, in particular, owing to low joining temperature. The present paper reports the results of the cooperated research performed in FY 1995-1996. To improve the quality of the SAJ, the configuration of the end tab surface to be joined with the insert material was modified. The torque measured during joining was also introduced in joining parameters. A nondestructive inspection, temperature measurements in the specimens during joining were performed. The effect of joining on Charpy impact properties was discussed. For practical application of the technique to irradiated specimens, we confirmed that the impact specimens with joining interface gave rise to no failure at the joining position during impact test after neutron irradiation. (author)

Technology development on analysis program for measuring fracture toughness of irradiated specimens

International Nuclear Information System (INIS)

Shibata, Akira; Takada, Fumiki

2007-03-01

The fracture toughness which represents resistance for brittle or ductile fracture is one of the most important material property concerning linear and non-linear fracture mechanics analyses. In order to respond to needs of collecting data relating to fracture toughness of pressure vessel and austenitic stainless steels, fracture toughness test for irradiated materials has been performed in JMTR hot laboratory. On the other hand, there has been no computer program for analysis of fracture toughness using the test data obtained from the test apparatus installed in the hot cell. Therefore, only load-displacement data have been provided to users to calculate fracture toughness of irradiated materials. Recently, request of analysis of fracture toughness have been increased. Thus a computer program, which calculates the amount of the crack extension, the compliance and the fracture toughness from the data acquired from the test apparatus installed in the hot cell, has been developed. In the program unloading elastic compliance method is applied based on ASTM E1820-01. Through the above development, the request for the fracture toughness analysis can be satisfied and the fracture toughness of irradiated test specimens can be provided to users. (author)

Extra spots in the electron diffraction patterns of neutron irradiated zirconium and its alloys

International Nuclear Information System (INIS)

Madden, P.K.

1977-01-01

Specimens of neutron irradiated zirconium and its alloys were examined in the transmission electron microscope. Groups of extra spots, often exhibiting four-fold symmetry, were observed in thin foil electron diffraction patterns of these specimens. The 'extra-spot' structure, like the expected black-dot/small scale dislocation loop neutron irradiated damage, is approximately 100 A in size. Its nature is uncertain. It may be related to irradiation damage or to some artefact introduced during specimen preparation. If it is the latter, then published irradiation damage defect size distributions and determined irradiation growth strains of other investigators, may require modification. The present inconclusive results indicate that extra-spot structure is likely to consist of oxide particles, but may correspond to hydride precipitation or decoration effects, or even, to electron beam effects. (author)

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

International Nuclear Information System (INIS)

Oyamatsu, Yasuko; Someya, Hiroyuki

2013-02-01

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

Post irradiation fracture properties of precipitation-strengthened alloy D21

International Nuclear Information System (INIS)

Huang, F.H.

1986-03-01

The precipitation strengthened alloys have the potential for use in fuel cladding and duct applications for liquid metal reactors due to their high strength and low swelling rate. Unfortunately, these high strength alloys tend to exhibit poor fracture toughness, and the effects of neutron irradiation on the fracture properties of the material are of concern. Compact tension specimens of alloy D21 were irradiated in the Experimental Breeder Reactor II to a fluence of 2.7 x 10 22 n/cm 2 (E > 0.1 MeV) at 425, 500, 550 and 600 0 C. Fracture toughness tests on these specimens wre performed using electric potential techniques at temperatures ranging from 205 to 425 C. The material exhibited low postirradiation fracture toughness which increased with either increasing test or irradiation temperature. The tearing modulus, however, increased with increasing irradiation temperature but decreased with increasing test temperature. Results wre analyzed using the J-integral approach. The fracture toughness of irradiated D21 was evaluated essentially following the procedure recommended in ASTM Test Method E813. It was found that the data elimination limits illustrated in E813 were too large for the specimens tested, although the thickness criterion was satisfied. The precautions needed to determine J/sub 1c/ based on a reduced data qualification range were disussed

Feasibility Study of Laser Cutting for Fabrication of Tensile Specimen

International Nuclear Information System (INIS)

Jin, Y. G.; Baik, S. J.; Kim, G. S.; Heo, G. S.; Yoo, B. O.; Ahn, S. B.; Chun, Y. B.

2015-01-01

The specimen fabrication technique was established to machine the specimen from the irradiated materials. The wire cut EDM(electric discharge machine) was modified to fabricate the mechanical testing specimens from irradiated components and fuel claddings. The oxide layer removal system was also developed because the oxide layer on the surface of the irradiated components and claddings interrupted the applying the electric current during the processing. However, zirconium oxide is protective against further corrosion as well as beneficial to mechanical strength for the tensile deformation of the cladding. Thus, it is important to fabricate the irradiated specimens without removal of oxide layer on the surface of the irradiated structural components and claddings. In the present study, laser cutting system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the feasibility of the laser cutting system was studied for the fabrication of various types of irradiated specimens in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser beam machining system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the dimensions were compared for the feasibility of the laser cutting system. The effect of surface oxide layer was also investigated for machining process of the zircaloy-4 fuel cladding and it was found that laser beam machining could be a useful tool to fabricate the specimens with surface oxide layer

Feasibility Study of Laser Cutting for Fabrication of Tensile Specimen

Energy Technology Data Exchange (ETDEWEB)

Jin, Y. G.; Baik, S. J.; Kim, G. S.; Heo, G. S.; Yoo, B. O.; Ahn, S. B.; Chun, Y. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The specimen fabrication technique was established to machine the specimen from the irradiated materials. The wire cut EDM(electric discharge machine) was modified to fabricate the mechanical testing specimens from irradiated components and fuel claddings. The oxide layer removal system was also developed because the oxide layer on the surface of the irradiated components and claddings interrupted the applying the electric current during the processing. However, zirconium oxide is protective against further corrosion as well as beneficial to mechanical strength for the tensile deformation of the cladding. Thus, it is important to fabricate the irradiated specimens without removal of oxide layer on the surface of the irradiated structural components and claddings. In the present study, laser cutting system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the feasibility of the laser cutting system was studied for the fabrication of various types of irradiated specimens in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser beam machining system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the dimensions were compared for the feasibility of the laser cutting system. The effect of surface oxide layer was also investigated for machining process of the zircaloy-4 fuel cladding and it was found that laser beam machining could be a useful tool to fabricate the specimens with surface oxide layer.

Comparison of immersion density and improved microstructural characterization methods for measuring swelling in small irradiated disk specimens

International Nuclear Information System (INIS)

Sawai, T.; Suzuki, M.; Hishinuma, A.; Maziasz, P.J.

1992-01-01

The procedure of obtaining microstructural data from reactor-irradiated specimens has been carefully checked for accuracy by comparison of swelling data obtained from transmission electron microscopy (TEM) observations of cavities with density-change data measured using the Precision Densitometer at Oak Ridge National Laboratory (ORNL). Comparison of data measured by both methods on duplicate or, in some cases, on the same specimen has shown some appreciable discrepancies for US/Japan collaborative experiments irradiated in the High Flux Isotope Reactor (HFIR). The contamination spot separation (CSS) method was used in the past to determine the thickness of a TEM foil. Recent work has revealed an appreciable error in this method that can result in an overestimation of the foil thickness. This error causes lower swelling values to be measured by TEM microstructural observation relative to the Precision Densitometer. An improved method is proposed for determining the foil thickness by the CSS method, which includes a correction for the systematic overestimation of foil thickness. (orig.)

Fractographic examination of reduced activation ferritic/martensitic steel charpy specimens irradiated to 30 dpa at 370{degrees}C

Energy Technology Data Exchange (ETDEWEB)

Gelles, D.S.; Hamilton, M.L. [Pacific Northwest National Lab., Richland, WA (United States); Schubert, L.E. [Univ. of Missouri, Rolla, MO (United States)

1996-10-01

Fractographic examinations are reported for a series of reduced activation ferritic/Martensitic steel Charpy impact specimens tested following irradiation to 30 dpa at 370{degrees}C in FFTF. One-third size specimens of six low activation steels developed for potential application as structural materials in fusion reactors were examined. A shift in brittle fracture appearance from cleavage to grain boundary failure was noted with increasing manganese content. The results are interpreted in light of transmutation induced composition changes in a fusion environment.

Recent results on the neutron irradiation of ITER candidate copper alloys irradiated in DR-3 at 250{degrees}C to 0.3 dpa

Energy Technology Data Exchange (ETDEWEB)

Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Singh, B.N.; Toft, P.; Eldrup, M.

1997-04-01

Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment with additional specimens re-aged and given a reactor bakeout treatment at 350{degrees}C for 100 h. CuAl-25 was also heat treated to simulate the effects of a bonding thermal cycle on the material. A number of heat treated specimens were neutron irradiated at 250{degrees}C to a dose level of {approximately}0.3 dpa in the DR-3 reactor as Riso. The main effect of the bonding thermal cycle heat treatment was a slight decrease in strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post irradiation tests at 250{degrees}C showed a severe loss of ductility in the case of the CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens exhibited a reasonable amount of uniform elongation, with CuCrZr possessing a lower strength.

Ring ductility of irradiated Inconel 706 and Nimonic PE16

International Nuclear Information System (INIS)

Huang, F.H.; Fish, R.L.

1984-01-01

The tensile ductility of fast neutron-irradiated, precipitation-hardened alloys Inconel 706 and Nimonic PE16 has been observed to be very low for certain test conditions. Explanations for the low ductility behavior have been sought by examination of broken tensile specimens with microscopy and other similar techniques. A ring compression test provides a method of evaluating the ductility of irradiated cladding specimens. Unlike the conventional uniaxial tensile testing in which the tensile specimen is deformed uniformly, the ring specimen is subjected to localized bending where the crack is initiated. The ductility can be estimated through an analysis of the bending of a ring in terms of strain hardening. Ring sections from irradiated, solution-treated Inconel 706 and Nimonic PE16 were compressed in the diametral direction to provide load-deflection records over a wide range of irradiation and test temperatures. Results showed that ductility in both alloys decreased with increasing test temperatures. The poorest ductility was exhibited at different irradiation temperatures in the two alloys - near 550 0 C for PE16 and 460 to 520 0 C for Inconel 706. The ring ductility data indicate that the grain boundary strength is a major factor in controlling the ductility of the PE16 alloy

Elastic-plastic analysis of the SS-3 tensile specimen

International Nuclear Information System (INIS)

Majumdar, S.

1998-01-01

Tensile tests of most irradiated specimens of vanadium alloys are conducted using the miniature SS-3 specimen which is not ASTM approved. Detailed elastic-plastic finite element analysis of the specimen was conducted to show that, as long as the ultimate to yield strength ratio is less than or equal to 1.25 (which is satisfied by many irradiated materials), the stress-plastic strain curve obtained by using such a specimen is representative of the true material behavior

Fracture toughness measurements with subsize disk compact specimens

International Nuclear Information System (INIS)

Alexander, D.J.

1994-01-01

Special fixtures and test methods have been developed for testing small disk compact specimens (1.25 mm diam by 4.6 mm thick). Specimens of European type 316L austenitic stainless steel were irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 or 250 C and tested over a temperature range from 20 to 250 C. Results show that irradiation to this dose level at these temperatures reduces the fracture toughness but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250 C is more damaging than at 90 C, causing larger decreases in the fracture toughness. The testing shows that it is possible to generate useful fracture toughness data with a small disk compact specimens

Stress-stain relations of irradiated stainless steels below 673 K

International Nuclear Information System (INIS)

Jitsukawa, S.; Hishinuma, A.; Grossbeck, M.L.

1992-01-01

Most specimens, irrespective of thermo-mechnaical treatment, exhibited proof stress levels of above 800 MPa and uniform elongations below 1% after irradiation in the the High Flux Isotope Reactor (HFIR). Only the solution annealed specimens irradiated at a low temperature of 328 k showed uniform elongations larger than 5% and proof stresses smaller than 800 MPa. Irradiation in the High Flux Reactor (HFR) caused more hardening than did irradiation in the HFIR. Ductility loss and change in work hardening characteristics by HFR irradiation were evaluated from reduction of area values. Residual ductility was revealed to be larger than 0.5 in natural strain, and the irradiation was estimated to have a small effect on work hardening characteristics and on fracture stress. The ductility of the irradiated alloys was found to be about 58% of that for the unirradiated alloys, as has been previously reported for irradiation in the HFIR. It was also demonstrated that true stress-strain relations, except for the fracture conditions, could be represented by Swift's type constitutive equation. (orig.)

Assessment of plastic flow and fracture properties with small specimens test techniques for IFMIF-designed specimens

International Nuclear Information System (INIS)

Spaetig, P.; Campitelli, E.N.; Bonade, R.; Baluc, N.

2005-01-01

The primary mission of the International Fusion Material Irradiation Facility (IFMIF) is to generate a material database to be used for the design of various components, for the licensing and for the assessment of the safe operation of a demonstration fusion reactor. IFMIF is an accelerator-based high-energy neutron source whose irradiation volume is quite limited (0.5 l for the high fluence volume). This requires the use of small specimens to measure the irradiation-induced changes on the physical and mechanical properties of materials. In this paper, we developed finite element models to better analyze the results obtained with two different small specimen test techniques applied to the tempered martensitic steel F82H-mod. First, one model was used to reconstruct the load-deflection curves of small ball punch tests, which are usually used to extract standard tensile parameters. It was shown that a reasonable assessment of the overall plastic flow can be done with small ball punch tests. Second, we investigated the stress field sensitivity at a crack tip to the constitutive behavior, for a crack modeled in plane strain, small-scale yielding and fracture mode I conditions. Based upon a local criterion for cleavage, that appears to be the basis to account for the size and geometry effects on fracture toughness, we showed that the details of the constitutive properties play a key role in modeling the irradiation-induced fracture toughness changes. Consequently, we suggest that much more attention and efforts have to be paid in investigating the post-yield behavior of the irradiated specimens and, in order to reach this goal, we recommend the use of not only tensile specimens but also that of compression ones in the IFMIF irradiation matrices. (author)

Surface, structural and tensile properties of proton beam irradiated zirconium

Energy Technology Data Exchange (ETDEWEB)

Rafique, Mohsin; Chae, San; Kim, Yong-Soo, E-mail: yongskim@hanyang.ac.kr

2016-02-01

This paper reports the surface, structural and tensile properties of proton beam irradiated pure zirconium (99.8%). The Zr samples were irradiated by 3.5 MeV protons using MC-50 cyclotron accelerator at different doses ranging from 1 × 10{sup 13} to 1 × 10{sup 16} protons/cm{sup 2}. Both un-irradiated and irradiated samples were characterized using Field Emission Scanning Electron Microscope (FESEM), X-ray Diffraction (XRD) and Universal Testing Machine (UTM). The average surface roughness of the specimens was determined by using Nanotech WSxM 5.0 develop 7.0 software. The FESEM results revealed the formation of bubbles, cracks and black spots on the samples’ surface at different doses whereas the XRD results indicated the presence of residual stresses in the irradiated specimens. Williamson–Hall analysis of the diffraction peaks was carried out to investigate changes in crystallite size and lattice strain in the irradiated specimens. The tensile properties such as the yield stress, ultimate tensile stress and percentage elongation exhibited a decreasing trend after irradiation in general, however, an inconsistent behavior was observed in their dependence on proton dose. The changes in tensile properties of Zr were associated with the production of radiation-induced defects including bubbles, cracks, precipitates and simultaneous recovery by the thermal energy generated with the increase of irradiation dose.

Surface, structural and tensile properties of proton beam irradiated zirconium

Science.gov (United States)

Rafique, Mohsin; Chae, San; Kim, Yong-Soo

2016-02-01

This paper reports the surface, structural and tensile properties of proton beam irradiated pure zirconium (99.8%). The Zr samples were irradiated by 3.5 MeV protons using MC-50 cyclotron accelerator at different doses ranging from 1 × 1013 to 1 × 1016 protons/cm2. Both un-irradiated and irradiated samples were characterized using Field Emission Scanning Electron Microscope (FESEM), X-ray Diffraction (XRD) and Universal Testing Machine (UTM). The average surface roughness of the specimens was determined by using Nanotech WSxM 5.0 develop 7.0 software. The FESEM results revealed the formation of bubbles, cracks and black spots on the samples' surface at different doses whereas the XRD results indicated the presence of residual stresses in the irradiated specimens. Williamson-Hall analysis of the diffraction peaks was carried out to investigate changes in crystallite size and lattice strain in the irradiated specimens. The tensile properties such as the yield stress, ultimate tensile stress and percentage elongation exhibited a decreasing trend after irradiation in general, however, an inconsistent behavior was observed in their dependence on proton dose. The changes in tensile properties of Zr were associated with the production of radiation-induced defects including bubbles, cracks, precipitates and simultaneous recovery by the thermal energy generated with the increase of irradiation dose.

Test methodology and technology of fracture toughness for small size specimens

Energy Technology Data Exchange (ETDEWEB)

Wakai, E.; Takada, F.; Ishii, T.; Ando, M. [Japan Atomic Energy Agency, Naga-gun, Ibaraki-ken (Japan); Matsukawa, S. [JNE Techno-Research Co., Kanagawa-ken (Japan)

2007-07-01

Full text of publication follows: Small specimen test technology (SSTT) is required to investigate mechanical properties in the limited availability of effective irradiation volumes in test reactors and accelerator-based neutron and charged particle sources. The test methodology guideline and the manufacture processes for very small size specimens have not been established, and we would have to formulate it. The technology to control exactly the load and displacement is also required in the test technology under the environment of high dose radiation produced from the specimens. The objective of this study is to examine the test technology and methodology of fracture toughness for very small size specimens. A new bend test machine installed in hot cell has been manufactured to obtain fracture toughness and DBTT (ductile - brittle transition temperature) of reduced-activation ferritic/martensitic steels for small bend specimens of t/2-1/3PCCVN (pre-cracked 1/3 size Charpy V-notch) with 20 mm length and DFMB (deformation and fracture mini bend specimen) with 9 mm length. The new machine can be performed at temperatures from -196 deg. C to 400 deg. C under unloading compliance method. Neutron irradiation was also performed at about 250 deg. C to about 2 dpa in JMTR. After the irradiation, fracture toughness and DBTT were examined by using the machine. Checking of displacement measurement between linear gauge of cross head's displacement and DVRT of the specimen displacement was performed exactly. Conditions of pre-crack due to fatigue in the specimen preparation were also examined and it depended on the shape and size of the specimens. Fracture toughness and DBTT of F82H steel for t/2-1/3PCCVN, DFMB and 0.18DCT specimens before irradiation were examined as a function of temperature. DBTT of smaller size specimens of DFMB was lower than that of larger size specimen of t/2-1/3PCCVN and 0.18DCT. The changes of fracture toughness and DBTT due to irradiation were also

Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

Energy Technology Data Exchange (ETDEWEB)

Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Shiba, K. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

2007-07-01

Full text of publication follows: Neutron irradiation of 9-12% Cr ferritic/martensitic steels below 425-450 deg. C produces microstructural defects that cause an increase in yield stress and ultimate tensile strength. This irradiation hardening causes embrittlement, which is observed in Charpy impact and toughness tests as an increase in ductile-brittle transition temperature (DBTT). Based on observations that show little change in strength in these steels irradiated above 425-450 deg. C, the general conclusion has been that no embrittlement occurs above this irradiation-hardening temperature regime. In a recent study of F82H steel irradiated at 300, 380, and 500 deg. C, irradiation hardening-an increase in yield stress-was observed in tensile specimens irradiated at the two lower temperatures, but no change was observed for the specimens irradiated at 500 deg. C. As expected, an increase in DBTT occurred for the Charpy specimens irradiated at 300 and 380 deg. C. However, there was an unexpected increase in the DBTT of the specimens irradiated at 500 deg. C. The observed embrittlement was attributed to the irradiation-accelerated precipitation of Laves phase. This conclusion was based on results from a detailed thermal aging study of F82H, in which tensile and Charpy specimens were aged at 500, 550, 600, and 650 deg. C to 30,000 h. These studies indicated that there was a decrease in yield stress at the two highest temperatures and essentially no change at the two lowest temperatures. Despite the strength decrease or no change, the DBTT increased for Charpy specimens irradiated at all four temperatures. Precipitates were extracted from thermally aged specimens, and the amount of precipitate was correlated with the increase in transition temperature. Laves phase was identified in the extracted precipitates by X-ray diffraction. Earlier studies on conventional elevated-temperature steels also showed embrittlement effects above the irradiation-hardening temperature

irradiation growth in annealed Zr2.5wt%Nb at 3530K

International Nuclear Information System (INIS)

Rogerson, A.; Murgatroyd, R.A.

1978-10-01

Zr 2.5wt%Nb growth specimens have been irradiated at 353 0 K to a fast neutron dose of approximately 4.0 x 10 25 n/m 2 . Specimens were taken from the longitudinal and transverse directions of a nominally annealed, seam-welded tube and irradiated in both the stress relieved and fully annealed conditions. Growth in these specimens is characterised by large positive and negative strains in the longitudinal and transverse directions respectively, with dimensional changes in weld material exhibiting intermediate growth behaviour. The results are compared with growth data on both annealed and cold worked Zircaloy-2 at 353 0 K and discussed in terms of the effect of texture, grain size, and cold work on irradiation growth. It is concluded that the continuation of growth to high doses in annealed Zr-2.5wt%Nb at 353 0 K results from interstitial induced dislocation climb with vacancies diffusing to grain boundaries. (author)

Fracture mechanics behaviour of neutron irradiated Alloy A-286

International Nuclear Information System (INIS)

Mills, W.J.; James, L.A.

The effect of fast-neutron irradiation on the fatigue-crack propagation and fracture toughness behaviour of Alloy A-286 was characterized using fracture mechanics techniques. The fracture toughness was found to decrease continuously with increasing irradiation damage at both 24 deg. C and 427 deg. C. In the unirradiated and low fluence conditions, specimens displayed appreciable plasticity prior to fracture, and equivalent Ksub(Ic) values were determined from Jsub(Ic) fracture toughness results. At high irradiation exposure levels, specimens exhibited a brittle Ksub(Ic) fracture mode. The 427 deg. C fracture toughness fell from 129 MPa√m in the unirradiated condition to 35 MPa√m at an exposure of 16.2 dpa (total fluence of 5.2x10 22 n/cm 2 ). Room temperature fracture toughness values were consistently 40 to 60 percent higher than the 427 deg. C values. Electron fractography revealed that the reduction in fracture resistance was attributed to a fracture mechanism transition from ductile microvoid coalescence to channel fracture. Fatigue-crack propagation tests were conducted at 427 deg. C on specimens irradiated at 2.4 dpa and 16.2 dpa. Crack growth rates at the lower exposure level were comparable to those in unirradiated material, while those at the higher exposure were slightly higher than in unirradiated material. (author)

Characterisation of neutron irradiation damage in zirconium alloys - a 'Round Robin' experiment

International Nuclear Information System (INIS)

Kelly, P.M.; Blake, R.G.; Jostsons, A.

1977-01-01

The nature of the damage structure in the neutron-irradiated zirconium specimens supplied as part of an international 'Round Robin' experiment has been studied using transmission electron microscopy. The damage structure consists entirely of a/3 dislocation loops and no evidence has been found for c component loops. Both vacancy and interstitial loops were found in specimens where inside/outside contrast analysis was possible. Quantitative measurements of loop size distributions and loop concentrations are reported. All specimens exhibited corduroy contrast to varying degress. (author)

Bright field electron microscopy of biological specimens

International Nuclear Information System (INIS)

Johansen, B.V.

1976-01-01

A preirradiation procedure is described which preserves negatively stained morphological features in bright field electron micrographs to a resolution of about 1.2 nm. Prior to microscopy the pre-irradiation dose (1.6 x 10 -3 C cm -2 ) is given at low electron optical magnification at five different areas on the grid (the centre plus four 'corners'). This pre-irradiation can be measured either with a Faraday cage or through controlled exposure-developing conditions. Uranyl formate stained T2 bacteriophages and stacked disk aggregates of Tobacco Mosaic Virus (TMV) protein served as test objects. A comparative study was performed on specimens using either the pre-irradiation procedure or direct irradiation by the 'minimum beam exposure' technique. Changes in the electron diffraction pattern of the stain-protein complex and the disappearance of certain morphological features in the specimens were both used in order to compare the pre-irradiation method with the direct exposure technique. After identical electron exposures the pre-irradiation approach gave a far better preservation of specimen morphology. Consequently this procedure gives the microscopist more time to select and focus appropriate areas for imaging before deteriorations take place. The investigation also suggested that microscopy should be carried out between 60,000 and 100,000 times magnification. Within this magnification range, it is possible to take advantage of the phase contrast transfer characteristics of the objective lens while the electron load on the object is kept at a moderate level. Using the pre-irradiation procedure special features of the T2 bacteriophage morphology could be established. (author)

Development of Reconstitution Technology for Surveillance Specimens

International Nuclear Information System (INIS)

Yasushi Atago; Shunichi Hatano; Eiichiro Otsuka

2002-01-01

The Japan Power Engineering and Inspection Corporation (JAPEIC) has been carrying out the project titled 'Nuclear Power Plant Integrated Management Technology (PLIM)' consigned by Japanese Ministry of Economy, Trade and Industry (METI) since 1996FY as a 10-years project. As one of the project themes, development of reconstitution technology for reactor pressure vessel (RPV/RV) surveillance specimens, which are installed in RPVs to monitor the neutron irradiation embrittlement on RPV/RV materials, is now on being carried out to deal with the long-term operation of nuclear power plants. The target of this theme is to establish the technical standard for applicability of reconstituted surveillance specimens including the reconstitution of the Charpy specimens and Compact Tension (CT) specimens. With the Charpy specimen reconstitution, application of 10 mm length inserts is used, which enables the conversion of tests from the LT-direction to the TL-direction. This paper presents the basic data from Charpy and CT specimens of RPV materials using the surveillance specimens obtained for un-irradiated materials including the following. 1) Reconstitution Technology of Charpy Specimens. a) The interaction between plastic zone and Heat Affected Zone (HAZ). b) The effects of the possible deviations from the standard specimens for the reconstituted specimens. 2) Reconstitution Technology of CT specimens. a) The correlation between fracture toughness and plastic zone width. Because the project is now in progress, this paper describes the outline of the results obtained as of the end of 2000 FY. (authors)

Fracture toughness measurements with subsize disk compact specimens

International Nuclear Information System (INIS)

Alexander, D.J.

1992-01-01

Special fixtures and test methods are necessary to facilitate the fracture toughness testing of small disk compact specimens of irradiated candidate materials for first-wall fusion applications. New methods have been developed for both the unloading compliance and potential drop techniques of monitoring crack growth. Provisions have been made to allow the necessary probes and instrumentation to be installed remotely using manipulators for testing of irradiated specimens in a hot cell. Laboratory trials showed that both unloading compliance and potential drop gave useful results. Both techniques gave similar data, and predicted the final crack extension within allowable limits. The results from the small disk compact specimens were similar to results from conventional compact specimen 12.7 mm thick. However, the slopes of the J-R curves from the larger specimens were lower, suggesting that the smaller disk compact specimens may have lost some constraint due to their size. The testing shows that it should be possible to generate useful J-R curve fracture toughness data from the small disk compact specimens

The ARBOR irradiation project

International Nuclear Information System (INIS)

Petersen, C.; Shamardin, V.; Fedoseev, A.; Shimansky, G.; Efimov, V.; Rensman, J.

2002-01-01

The irradiation project 'ARBOR', for 'Associated Reactor Irradiation in BOR 60', includes 150 mini-tensile/low cycle fatigue specimens and 150 mini-Charpy (KLST) specimens of nine different RAFM steels. Specimens began irradiation on 22 November 2000 in an specially designed irradiation rig in BOR 60, in a fast neutron flux (>0.1 MeV) of 1.8x10 15 n/cm 2 s and with direct sodium cooling at a temperature less than 340 deg. C. Tensile, low cycle fatigue and Charpy specimens of the following materials are included: EUROFER 97, F82H mod., OPTIFER IVc, EUROFER 97 with different boron contents, ODS-EUROFER 97, as well as EUROFER 97 electron-beam welded and reference bulk material, from NRG, Petten

Demonstration of Laser Cutting System for Tube Specimen

Energy Technology Data Exchange (ETDEWEB)

Jin, Y. G.; Kim, G. S.; Heo, G. S.; Baik, S. J.; Kim, H. M.; Ahn, S. B. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The oxide layer removal system was also developed because the oxide layer on the surface of the irradiated fuel cladding and components interrupted the applying the electric current during the processing. However, it was found that the mechanical testing data of the irradiated specimens with removal of oxide layer was less reliable than the specimens with oxide layer . The laser cutting system using Nd:YAG with fiber optic beam delivery has great potential in material processing applications of the irradiated fuel cladding and components due to non-contact process. Thus, the oxide layer doesn't interrupt the fabrication process during the laser cutting system. In the present study, the laser cutting system was designed to fabricate the mechanical testing specimens from the unirradiated fuel cladding with and without oxide. The feasibility of the laser cutting system was demonstrated for the fabrication of various types of unirradiated specimens. The effect of surface oxide layer was also investigated for machining process of the zirlo fuel cladding and it was found that laser beam machining could be a useful tool to fabricate the specimens with surface oxide layer. Based on the feasibility studies and demonstration, the design of the laser cutting machine for fully or partially automatic and remotely operable system will be proposed and made.

Evaluation of irradiation hardening of proton irradiated stainless steels by nanoindentation

International Nuclear Information System (INIS)

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

2014-01-01

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

Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

International Nuclear Information System (INIS)

Haggag, F.M.; Nanstad, R.K.; Byrne, S.T.

1991-01-01

The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low [121 to 210 degrees C (250--410 degrees F)] compared to those for commercial light-water reactors (LWRs) [∼288 degrees C (550 degrees F)]. The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 x10 18 neutron/cm 2 [0.68 x 10 18 neutron/cm 2 (>1 MeV)] at temperatures of 288, 204, 163, and 121 degrees C (550, 400, 325, and 250 degrees F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs

Results of charpy V-notch impact testing of structural steel specimens irradiated at ∼30 degrees C to 1 x 1016 neutrons/cm2 in a commercial reactor cavity

International Nuclear Information System (INIS)

Iskander, S.K.; Stoller, R.E.

1997-04-01

A capsule containing Charpy V-notch (CVN) and mini-tensile specimens was irradiated at ∼ 30 degrees C (∼ 85 degrees F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 10 16 neutrons/cm 2 (> 1MeV). The capsule included six CVN impact specimens of archival High Flux Isotope Reactor A212 grade B ferritic steel and five CVN impact specimens of a well-studied A36 structural steel. This irradiation was part of the ongoing study of neutron-induced damage effects at the low temperature and flux experienced by reactor supports. The plant operators shut down the plant before the planned exposure was reached. The exposure of these specimens produced no significant irradiation-induced embrittlement. Of interest were the data on unirradiated specimens in the L-T orientation machined from a single plate of A36 structural steel, which is the same specification for the structural steel used in some reactor supports. The average CVN energy of five unirradiated specimens obtained from one region of the plate and tested at room temperature was ∼ 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of ∼ 220%. The CVN impact energies for all 18 specimens ranged from a low of 32 J to a high of 111 J. Moreover, it appears that the University of Kansas CVN impact energy data of the unirradiated specimens at the 100-J level are shifted toward higher temperatures by about 20 K. The results were an example of the extent of scatter possible in CVN impact testing. Generic values for the CVN impact energy of A36 should be used with caution in critical applications

The ARBOR irradiation project

Energy Technology Data Exchange (ETDEWEB)

Petersen, C. E-mail: claus.petersen@imf.fzk.de; Shamardin, V.; Fedoseev, A.; Shimansky, G.; Efimov, V.; Rensman, J

2002-12-01

The irradiation project 'ARBOR', for 'Associated Reactor Irradiation in BOR 60', includes 150 mini-tensile/low cycle fatigue specimens and 150 mini-Charpy (KLST) specimens of nine different RAFM steels. Specimens began irradiation on 22 November 2000 in an specially designed irradiation rig in BOR 60, in a fast neutron flux (>0.1 MeV) of 1.8x10{sup 15} n/cm{sup 2} s and with direct sodium cooling at a temperature less than 340 deg. C. Tensile, low cycle fatigue and Charpy specimens of the following materials are included: EUROFER 97, F82H mod., OPTIFER IVc, EUROFER 97 with different boron contents, ODS-EUROFER 97, as well as EUROFER 97 electron-beam welded and reference bulk material, from NRG, Petten.

Pre- and post-irradiation properties of copper alloys at 250 deg. C following bonding and bakeout thermal cycles

International Nuclear Information System (INIS)

Singh, B.N.; Edwards, D.J.; Eldrup, M.; Toft, P.

1997-01-01

Screening experiments were carried out to investigate the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties and electrical resistivity of the oxide dispersion strengthened (GlidCop, CuAl-25) and the precipitation hardened (CuCrZr, CuNiBe) copper alloys. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing, and bonding thermal treatment followed by re-ageing and the reactor bakeout treatment at 350 deg. C for 100 h. Tensile specimens of CuAl-25 were given the heat treatment corresponding to the bonding thermal cycle. A number of heat treated specimens were neuron irradiated at 250 deg. C to a dose level of ∼ 0.3 dpa in the DR-3 reactor at Risoe. Both unirradiated and irradiated specimens with various heat treatments were tensile tested at 250 deg. C. The microstructure and electrical resistivity of these specimens were determined in the unirradiated as well as irradiated conditions. The post-deformation microstructure of the irradiated specimens was also investigated. The fracture surfaces of both unirradiated and irradiated specimens were examined. Results of these investigations are reported in the present report. The main effect of the bonding thermal cycle heat treatment was a slight decrease in strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post irradiation tests at 250 deg. C showed a severe loss of ductility in the case of CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens, on the other hand, exhibited a reasonable amount of uniform elongation. The results are briefly discussed in terms of thermal and irradiation stability of precipitates and particles and irradiation-induced segregation, precipitation and recovery of dislocation microstructure. (au) 7 tabs., 28 ills., 15 refs

Fatigue performance of HFIR-irradiated Nimonic PE-16 at 4300C

International Nuclear Information System (INIS)

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

1983-01-01

Nimonic PE-16 was irradiated in the HFIR to 6 to 9 dpa and 560 to 1000 at. ppM He at 430 0 C. Postirradiation fatigue tests revealed a reduction in fatigue life by about a factor of 10 at 430 0 C. In contrast to AISI type 316 stainless steel, no endurance limit was observed. All irradiated specimens exhibited some intergranular fracture with an increasing tendency toward cleavage-like intragranular fracture for low strain ranges

Fracture toughness of irradiated Zr-2.5Nb pressure tube from KAPS-2 evaluated using disk compact tension specimens

International Nuclear Information System (INIS)

Shah, Priti Kotak; Dubey, J.S.; Balakrishnan, K.S.; Shriwastaw, R.S.; Dhotre, M.P.; Bhandekar, A.; Pandit, K.M.; Anantharaman, S.

2013-12-01

The report gives the results of the fracture toughness tests carried out over the range of temperatures on specimens prepared from the irradiated S-07Zr-2.5Nb pressure tube removed from Kakrapar Atomic Power Station-2 (KAPS-2) as a part of materials surveillance programme. The pressure tube had experienced ∼ 8 effective full power years (EFPY) of reactor operation and had hydrogen equivalent (H eq ) content less than 20 ppm along the tube length. The fracture toughness tests have been carried out using 30 mm Disk Compact Tension (DCT) specimens, that were punched out of the irradiated pressure tube. The disk punching was carried out using specially made shielded enclosure and hydraulic press. Fatigue pre-cracking and fracture toughness tests were performed using servo-hydraulic universal testing machine with Direct Current Potential Drop (DCPD) equipment to monitor the crack length. The tests were carried out at different test temperature from ambient to 300℃. The fracture toughness values have been used to estimate the critical pressure for the tube. The fracture properties indicate that such tubes have sufficient toughness to satisfy the Leak-Before-Break (LBB) criterion for in-reactor operation. (author)

Tension test system for irradiated small specimens operated by remote control

International Nuclear Information System (INIS)

Okada, Akira

1993-01-01

A robot-based tension test system has been developed to aid in the mechanical testing of highly radioactive specimens. This system reduces radiation hazards from specimens and allows for the uniform precision of testing results independent of experimenters' skills. The robot system is designed to accommodate a miniaturized tension specimen with a gage section 5.5 by 1.2 mm, with a total length and width of 12.5 and 2.3 mm, respectively, and thickness of about 0.2 mm. The system is composed of a manipulating robot, a vibrational-type specimen feeder, a rotating-type specimen tray, a specimen observation system, a simulated tension text fixture, and a microcomputer for controlling the system. This system accomplishes specimen arrangement in the specimen tray, specimen transportation and loading to the test fixture and testing, and removal of the broken specimen from the fixture. These procedures are performed quickly, safely, and with uniform testing precision by computer control remotely by an unskilled experimenter

Final report on graphite irradiation test OG-2

International Nuclear Information System (INIS)

Price, R.J.; Beavan, L.A.

1975-01-01

Results are presented of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on specimens of nuclear graphites irradiated in capsule OG-2. About half the irradiation space was allocated to H-451 near-isotropic petroleum-coke-based graphite or its subsized prototype grade H-429. Most of these specimens had been previously irradiated. Virgin specimens of another near-isotropic graphite, grade TS-1240, were irradiated. Some previously irradiated specimens of needle-coke-based H-327 graphite and pitch-coke-based P 3 JHAN were also included

Evaluation of Specimen Geometric Effect for Laser Flash Thermal Diffusivity Test

International Nuclear Information System (INIS)

Park, Dae Gyu; Kim, Hee Moon; Song, Woong Sub; Baik, Seung Je; Ryu, Woo Seok; Ahn, Sang Bok; Joo, Young Sun

2012-01-01

KAERI(Korea Atomic Energy Research Institute) is developing a new type of nuclear reactor, the so called 'SMART' (System Integrated Modular Advanced Reactor) reactor. Alloy 690 was selected as the candidate material for the heat exchanger tube of of SMART's steam generator. The SMART R and D is now facing the stage of engineering verification and standard design approval for application of DEMO reactors. Therefore, the material performance under the relevant environment needs to be evaluated. The one of the important material performance issues is thermal conductivity, which the engineering database is necessary for the steam generator design. However, the neutron post irradiation characteristics of alloy 690 are little known. As a result, a PIE (Post Irradiation Examination) of the thermal properties have been plan for a 4 times, so called base line test, 1 st irradiation test, 2 nd and 3 rd irradiation test. But there is some constraint to perform thermal diffusivity test owing to test specimen. Originally thermal diffusivity test are planed using disk shape with 9 mm diameter and 1 mm thick specimen. Due to mismatch of neutron irradiation schedule, thermal diffusivity will be tested by different shape and size specimens at 1 st irradiation test. Therefore, verification of geometric and size effect are necessary for test specimen in order to achieve accurate test results

Design and use of nonstandard tensile specimens for irradiated materials testing

International Nuclear Information System (INIS)

Panayotou, N.F.

1984-10-01

Miniature, nonstandard, tensile-type specimens have been developed for use in radiation effects experiments at high energy neutron sources where the useful radiation volume is as small as a few cubic centimeters. The end result of our development is a sheet-type specimen, 12.7 mm long with a 5.1 mm long, 1.0 mm wide gage section, which is typically fabricated from 0.25 mm thick sheet stock by a punching technique. Despite this miniature geometry, it has been determined that the data obtained using these miniature specimens are in good agreement with data obtained using much larger specimens. This finding indicates that miniature tensile specimen data may by used for engineering design purposes. Furthermore, it is clear that miniature tensile specimen technology is applicable to fields other than the study of radiation effects. This paper describes the miniature specimen technology which was developed and compares the data obtained from these miniature specimens to data obtained from much larger specimens. 9 figures

Oxide glass structure evolution under swift heavy ion irradiation

International Nuclear Information System (INIS)

Mendoza, C.; Peuget, S.; Charpentier, T.; Moskura, M.; Caraballo, R.; Bouty, O.; Mir, A.H.; Monnet, I.; Grygiel, C.; Jegou, C.

2014-01-01

Highlights: • Structure of SHI irradiated glass is similar to the one of a hyper quenched glass. • D2 Raman band associated to 3 members ring is only observed in irradiated glass. • Irradiated state seems slightly different to an equilibrated liquid quenched rapidly. - Abstract: The effects of ion tracks on the structure of oxide glasses were examined by irradiating a silica glass and two borosilicate glass specimens containing 3 and 6 oxides with krypton ions (74 MeV) and xenon ions (92 MeV). Structural changes in the glass were observed by Raman and nuclear magnetic resonance spectroscopy using a multinuclear approach ( 11 B, 23 Na, 27 Al and 29 Si). The structure of irradiated silica glass resembles a structure quenched at very high temperature. Both borosilicate glass specimens exhibited depolymerization of the borosilicate network, a lower boron coordination number, and a change in the role of a fraction of the sodium atoms after irradiation, suggesting that the final borosilicate glass structures were quenched from a high temperature state. In addition, a sharp increase in the concentration of three membered silica rings and the presence of large amounts of penta- and hexacoordinate aluminum in the irradiated 6-oxide glass suggest that the irradiated glass is different from a liquid quenched at equilibrium, but it is rather obtained from a nonequilibrium liquid that is partially relaxed by very rapid quenching within the ion tracks

Vanderbilt University Gamma Irradiation of Nano-modified Concrete (2017 Milestone Report)

Energy Technology Data Exchange (ETDEWEB)

Deichert, Geoffrey G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Selby, Aaron P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Reches, Yonathan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

This document outlines the irradiation of concrete specimens in the Gamma Irradiation Facility in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Two gamma irradiation runs were performed in July of 2017 on 18 reference mortar bar specimens, 26 reference cement paste bar specimens, and 28 reference cement paste tab specimens to determine the dose and temperature response of the specimens in the gamma irradiation environment. Specimens from the first two gamma irradiations were surveyed and released to Vanderbilt University. The temperature and dose information obtained informs the test parameters of the final two gamma irradiations of nano-modified concrete planned for FY 2018.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Effects of neutron irradiation on thermal conductivity of SiC-based composites and monolithic ceramics

International Nuclear Information System (INIS)

Senor, D.J.; Youngblood, G.E.; Moore, C.E.; Trimble, D.J.; Woods, J.J.

1996-06-01

A variety of SiC-based composites and monolithic ceramics were characterized by measuring their thermal diffusivity in the unirradiated, thermal annealed, and irradiated conditions over the temperature range 400 to 1,000 C. The irradiation was conducted in the EBR-II to doses of 33 and 43 dpa-SiC (185 EFPD) at a nominal temperature of 1,000 C. The annealed specimens were held at 1,010 C for 165 days to approximately duplicate the thermal exposure of the irradiated specimens. Thermal diffusivity was measured using the laser flash method, and was converted to thermal conductivity using density data and calculated specific heat values. Exposure to the 165 day anneal did not appreciably degrade the conductivity of the monolithic or particulate-reinforced composites, but the conductivity of the fiber-reinforced composites was slightly degraded. The crystalline SiC-based materials tested in this study exhibited thermal conductivity degradation of irradiation, presumably caused by the presence of irradiation-induced defects. Irradiation-induced conductivity degradation was greater at lower temperatures, and was typically more pronounced for materials with higher unirradiated conductivity. Annealing the irradiated specimens for one hour at 150 C above the irradiation temperature produced an increase in thermal conductivity, which is likely the result of interstitial-vacancy pair recombination. Multiple post-irradiation anneals on CVD β-SiC indicated that a portion of the irradiation-induced damage was permanent. A possible explanation for this phenomenon was the formation of stable dislocation loops at the high irradiation temperature and/or high dose that prevented subsequent interstitial/vacancy recombination

Effects of neutron irradiation on thermal conductivity of SiC-based composites and monolithic ceramics

International Nuclear Information System (INIS)

Senor, D.J.; Youngblood, G.E.; Moore, C.E.; Trimble, D.J.; Woods, J.J.

1997-05-01

A variety of SiC-based composites and monolithic ceramics were characterized by measuring their thermal diffusivity in the unirradiated, thermal annealed, and irradiated conditions over the temperature range 400 to 1,000 C. The irradiation was conducted in the EBR-II to doses of 33 and 43 dpa-SiC (185 EFPD) at a nominal temperature of 1,000 C. The annealed specimens were held at 1,010 C for 165 days to approximately duplicate the thermal exposure of the irradiated specimens. Thermal diffusivity was measured using the laser flash method, and was converted to thermal conductivity using density data and calculated specific heat values. Exposure to the 165 day anneal did not appreciably degrade the conductivity of the monolithic or particulate-reinforced composites, but the conductivity of the fiber-reinforced composites was slightly degraded. The crystalline SiC-based materials tested in this study exhibited thermal conductivity degradation after irradiation, presumably caused by the presence of irradiation-induced defects. Irradiation-induced conductivity degradation was greater at lower temperatures, and was typically more pronounced for materials with higher unirradiated conductivity. Annealing the irradiated specimens for one hour at 150 C above the irradiation temperature produced an increase in thermal conductivity, which is likely the result of interstitial-vacancy pair recombination. Multiple post-irradiation anneals on CVD β-SiC indicated that a portion of the irradiation-induced damage was permanent. A possible explanation for this phenomenon was the formation of stable dislocation loops at the high irradiation temperature and/or high dose that prevented subsequent interstitial/vacancy recombination

Tensile properties of vanadium alloys irradiated at 390{degrees}C in EBR-II

Energy Technology Data Exchange (ETDEWEB)

Chung, H.M.; Tsai, H.C.; Nowicki, L.J. [Argonne National Lab., IL (United States)] [and others

1997-08-01

Vanadium alloys were irradiated in Li-bonded stainless steel capsules to {approx}390{degrees}C in the EBR-II X-530 experiment. This report presents results of postirradiation tests of tensile properties of two large-scale (100 and 500 kg) heats of V-4Cr-Ti and laboratory (15-30 kg) heats of boron-doped V-4Cr-4Ti, V-8Cr-6Ti, V-5Ti, and V-3Ti-1Si alloys. Tensile specimens, divided into two groups, were irradiated in two different capsules under nominally similar conditions. The 500-kg heat (No. 832665) and the 100-kg heat (VX-8) of V-4Cr-4Ti irradiated in one of the subcapsules exhibited complete loss of work-hardening capability, which was manifested by very low uniform plastic strain. In contrast, the 100-kg heat of V-4Cr-4Ti irradiated in another subcapsule exhibited good tensile properties (uniform plastic strain 2.8-4.0%). A laboratory heat of V-3Ti-1Si irradiated in the latter subcapsule also exhibited good tensile properties. These results indicate that work-hardening capability at low irradiation temperatures varies significantly from heat to heat and is influenced by nominally small differences in irradiation conditions.

Evaluation of fatigue properties of HFIR-irradiated nimonic PE-16 at 4300C

International Nuclear Information System (INIS)

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

1984-01-01

Nimonic PE-16 was irradiated in the HFIR to 6 to 9 dpa and 560 to 1000 at. ppM He at 430 0 C. Postirradiation fatigue tests revealed a reduction in fatigue life by about a factor of 10 at 430 0 C. In contrast with AISI type 316 stainless steel, no endurance limit was observed. All irradiated specimens exhibited some intergranular fracture with an increasing tendency toward cleavage like intragranular fracture for low strain ranges

Development of reconstitution method for surveillance specimens using surface activated joining

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Terumi; Kaihara, Shoichiro; Yoshida, Kazuo; Sato, Akira [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan); Onizawa, Kunio; Nishiyama, Yutaka; Fukaya, Kiyoshi; Suzuki, Masahide

1996-03-01

Evaluation of embrittlement of reactor vessel steel due to irradiation requires surveillance tests. However, many surveillance specimens are necessary for nuclear plants life extension. Therefore, a specimen reconstitution technique has become important to provide the many specimens for continued surveillance. A surface activated joining (SAJ) method has been developed to join various materials together at low temperatures with little deformation, and is useful to bond irradiated specimens. To assess the validity of this method, Charpy impact tests were carried out, and the characteristics caused by heating during joining were measured. The test results showed the Charpy impact values were almost the same as base materials, and surface activated joining reduced heat affected zone to less than 2 mm. (author).

Specimen Machining for the Study of the Effect of Swelling on CGR in PWR Environment.

Energy Technology Data Exchange (ETDEWEB)

Teysseyre, Sebastien Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-06-01

This report describes the preparation of ten specimens to be used for the study of the effect of swelling on the propagation of irradiation assisted stress corrosion cracking cracks. Four compact tension specimens, four microscopy plates and two tensile specimens were machined from a AISI 304 material that was irradiated up to 33 dpa. The specimens had been machined such as to represent the behavior of materials with 3.7%swelling and <2% swelling.

Technology development on production of test specimens from irradiated capsule outer-tube and mechanical evaluation test of stainless steel with high dose carried out by the technology

International Nuclear Information System (INIS)

Hayashi, Koji; Shibata, Akira; Iwamatsu, Shigemi; Sozawa, Shizuo; Takada, Fumiki; Ohmi, Masao; Nakagawa, Tetsuya

2008-03-01

The irradiation capsule 74M-52J was irradiated during total 136 cycles at reactor core of JMTR and the maximum neutron dose reached on 3.9x10 26 n/m 2 at the capsule outer-tube made of a type 304 stainless steel. In order to produce mechanical test specimens from the outer-tube, a punching technique was developed as a simple remote-handling method in a hot-cell. From comparison between the punching and the mechanical cutting methods, it was clarified that the punching technique was applicable to practical use. Moreover, an evaluation test of mechanical properties using specimens sampled from the 74M-52 was performed with in-water high temperature condition, less than 288degC. The result shows that the residual elongation is 18% at 150degC and 13% at 288degC. It was confirmed that the type 304 stainless steel irradiated up to such high dose shows enough ductility. (author)

Effect of irradiation spectrum on the microstructure of ion-irradiated Al2O3

International Nuclear Information System (INIS)

Zinkle, S.J.

1994-01-01

Polycrystalline samples of alpha-alumina have been irradiated with various ions ranging from 3.6 MeV Fe + to 1 MeV H + ions at 650 C. Cross-section transmission electron microscopy was used to investigate the depth-dependent microstructure of the irradiated specimens. The microstructure following irradiation was observed to be dependent on the irradiation spectrum. In particular, defect cluster nucleation was effectively suppressed in specimens irradiated with light ions such as 1 MeV H + ions. On the other hand, light ion irradiation tended to accelerate the growth rate of dislocation loops. The microstructural observations are discussed in terms of ionization enhanced diffusion processes

Development of piezoelectric ceramics driven fatigue testing machine for small specimens

International Nuclear Information System (INIS)

Saito, S.; Kikuchi, K.; Onishi, Y.; Nishino, T.

2002-01-01

A new fatigue testing machine with piezoelectric ceramics actuators was developed and a prototype was manufactured for high-cycle fatigue tests with small specimens. The machine has a simple mechanism and is compact. These features make it easy to set up and to maintain the machine in a hot cell. The excitation of the actuator can be transmitted to the specimen using a lever-type testing jig. More than 100 μm of displacement could be prescribed precisely to the specimen at a frequency of 50 Hz. This was sufficient performance for high-cycle bend fatigue tests on specimens irradiated at the SINQ target in Paul Scherrer Institute. The relationship of a displacement applied to the specimen and the strain of the necking part were obtained by experimental methods and by finite element method (FEM) calculations. Both results showed good agreement. This fact makes it possible to evaluate the strain of irradiated specimens by FEM simulations

Effect of neutron irradiation and post-irradiation annealing on microstructure and mechanical properties of OFHC-copper

International Nuclear Information System (INIS)

Singh, B.N.; Edwards, D.J.; Toft, P.

2001-01-01

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Risoe at 100 deg. C to doses in the range 0.01-0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 deg. C whereas others were given a post-irradiation annealing treatment at 300 deg. C for 50 h and subsequently tested at 100 deg. C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model

Effect of neutron irradiation and post-irradiation annealing on microstructure and mechanical properties of OFHC-copper

Energy Technology Data Exchange (ETDEWEB)

Singh, B.N. E-mail: bachu.singh@risoe.dk; Edwards, D.J.; Toft, P

2001-12-01

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Risoe at 100 deg. C to doses in the range 0.01-0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 deg. C whereas others were given a post-irradiation annealing treatment at 300 deg. C for 50 h and subsequently tested at 100 deg. C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model.

Effect of neutron irradiation and post-irradiation annealing on microstructure and mechanical properties of OFHC-copper

Science.gov (United States)

Singh, B. N.; Edwards, D. J.; Toft, P.

2001-12-01

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Risø at 100 °C to doses in the range 0.01-0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 °C whereas others were given a post-irradiation annealing treatment at 300 °C for 50 h and subsequently tested at 100 °C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model.

Evaluation of Ion Irradiation Behavior of ODS Alloys

Energy Technology Data Exchange (ETDEWEB)

Jang, Jin Sung; Kim, Min Chul; Hong, Jun Hwa; Han, Chang Hee; Chang, Young Mun; Bae, Chang Soo; Bae, Yoon Young; Chang, Moon Hee

2006-08-15

FM steel (Grade 92) and ODS alloy(MA956) specimens were ion irradiated with 122 MeV Ne ions. Irradiation temperatures were about 450 and 550 .deg. C and the peak dose was 1, 5, and 10 dpa. Cross-sectional TEM samples were prepared by the electrolytic Ni-plating after pre-treatment of the irradiated specimens. Irradiation cavities in FM steel and ODS alloy specimens were not much different in size; about 20 nm in diameter in both specimens irradiated at around 450 .deg. C. However, the size distribution of cavities in FM steel specimens was broader than that in ODS alloy specimen, indicating that the cavity growth probably via coalescence). It was noticeable that the location and the preferential growth of the cavities in FM steel specimens: cavities on the PAGB (prior austenite grain boundary) was significantly larger than those within the grains. This could be an important issue for the mechanical properties, especially high temperature creep, fracture toughness, and so on. The dependency of the dose threshold and swelling on the ratio of the inert gas concentration/dpa was analysed for the various irradiation source, including He, Ne, Fe/He, and fast neutron, and the empirical correlation was established.

Evaluation of Ion Irradiation Behavior of ODS Alloys

International Nuclear Information System (INIS)

Jang, Jin Sung; Kim, Min Chul; Hong, Jun Hwa; Han, Chang Hee; Chang, Young Mun; Bae, Chang Soo; Bae, Yoon Young; Chang, Moon Hee

2006-08-01

FM steel (Grade 92) and ODS alloy(MA956) specimens were ion irradiated with 122 MeV Ne ions. Irradiation temperatures were about 450 and 550 .deg. C and the peak dose was 1, 5, and 10 dpa. Cross-sectional TEM samples were prepared by the electrolytic Ni-plating after pre-treatment of the irradiated specimens. Irradiation cavities in FM steel and ODS alloy specimens were not much different in size; about 20 nm in diameter in both specimens irradiated at around 450 .deg. C. However, the size distribution of cavities in FM steel specimens was broader than that in ODS alloy specimen, indicating that the cavity growth probably via coalescence). It was noticeable that the location and the preferential growth of the cavities in FM steel specimens: cavities on the PAGB (prior austenite grain boundary) was significantly larger than those within the grains. This could be an important issue for the mechanical properties, especially high temperature creep, fracture toughness, and so on. The dependency of the dose threshold and swelling on the ratio of the inert gas concentration/dpa was analysed for the various irradiation source, including He, Ne, Fe/He, and fast neutron, and the empirical correlation was established

Study on creep-fatigue life of irradiated austenitic stainless steel

International Nuclear Information System (INIS)

Ioka, Ikuo; Miwa, Yukio; Tsuji, Hirokazu; Yonekawa, Minoru; Takada, Fumiki; Hoshiya, Taiji

2001-01-01

The low cycle creep-fatigue test with tensile strain hold of the austenitic stainless steel irradiated to 2 dpa was carried out at 823K in vacuum. The applicability of creep-fatigue life prediction methods to the irradiated specimen was examined. The fatigue life on the irradiated specimen without tensile strain hold time was reduced by a factor of 2-5 in comparison with the unirradiated specimen. The decline in fatigue life of the irradiated specimen with tensile strain hold was almost equal to that of the unirradiated specimen. The creep damage of both unirradiated and irradiated specimens was underestimated by the time fraction rule or the ductility exhaustion rule. The creep damage calculated by the time fraction rule or the ductility exhaustion rule increased by the irradiation. The predictions derived from the linear damage rule are unsafe as compared with the experimental fatigue lives. (author)

Tritium and helium retention and release from irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Anderl, R.A.; Longhurst, G.R.; Oates, M.A.; Pawelko, R.J. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States)

1998-01-01

This paper reports the results of an experimental effort to anneal irradiated beryllium specimens and characterize them for steam-chemical reactivity experiments. Fully-dense, consolidated powder metallurgy Be cylinders, irradiated in the EBR-II to a fast neutron (>0.1 MeV) fluence of {approx}6 x 10{sup 22} n/cm{sup 2}, were annealed at temperatures from 450degC to 1200degC. The releases of tritium and helium were measured during the heat-up phase and during the high-temperature anneals. These experiments revealed that, at 600degC and below, there was insignificant gas release. Tritium release at 700degC exhibited a delayed increase in the release rate, while the specimen was at 700degC. For anneal temperatures of 800degC and higher, tritium and helium release was concurrent and the release behavior was characterized by gas-burst peaks. Essentially all of the tritium and helium was released at temperatures of 1000degC and higher, whereas about 1/10 of the tritium was released during the anneals at 700degC and 800degC. Measurements were made to determine the bulk density, porosity and specific surface area for each specimen before and after annealing. These measurements indicated that annealing caused the irradiated Be to swell, by as much as 14% at 700degC and 56% at 1200degC. Kr gas adsorption measurements for samples annealed at 1000degC and 1200degC determined specific surface areas between 0.04 m{sup 2}/g and 0.1 m{sup 2}/g for these annealed specimens. The tritium and helium gas release measurements and the specific surface area measurements indicated that annealing of irradiated Be caused a porosity network to evolve and become surface-connected to relieve internal gas pressure. (author)

A novel hydrogel of poloxamer 407 and chitosan obtained by gamma irradiation exhibits physicochemical properties for wound management

Energy Technology Data Exchange (ETDEWEB)

Leyva-Gómez, Gerardo, E-mail: gerardoleyva@hotmail.com [Laboratory of Connective Tissue, Centro Nacional de Investigación y Atención de Quemados, Instituto Nacional de Rehabilitación Luis Guillermo Ibarra Ibarra, Mexico City (Mexico); Santillan-Reyes, Erika, E-mail: kikita5410@gmail.com [Laboratory of Connective Tissue, Centro Nacional de Investigación y Atención de Quemados, Instituto Nacional de Rehabilitación Luis Guillermo Ibarra Ibarra, Mexico City (Mexico); Lima, E, E-mail: lima@iim.unam.mx [Departamento de Materiales Metálicos y Cerámicos, Instituto de Investigaciones en Materiales, Universidad Nacional Autónoma de México, Mexico City (Mexico); Madrid-Martínez, Abigail, E-mail: abitzy@hotmail.com [Laboratory of Connective Tissue, Centro Nacional de Investigación y Atención de Quemados, Instituto Nacional de Rehabilitación Luis Guillermo Ibarra Ibarra, Mexico City (Mexico); Krötzsch, E, E-mail: kroted@yahoo.com.mx [Laboratory of Connective Tissue, Centro Nacional de Investigación y Atención de Quemados, Instituto Nacional de Rehabilitación Luis Guillermo Ibarra Ibarra, Mexico City (Mexico); and others

2017-05-01

Application of polymers cross-linked by gamma irradiation on cutaneous wounds has resulted in the improvement of healing. Chitosan (CH) and poloxamer 407 (P407)-based hydrogels confer different advantages in wound management. To combine the properties of both compounds, a gamma-irradiated mixture of 0.75/25% (w/w) CH and P407, respectively, was obtained (CH-P), and several physical, chemical, and biological analyses were performed. Notably, gamma radiation induced changes in the mixture's thermal behavior, viscosity, and swelling, and exhibited stability at neutral pH. The thermal reversibility provided by P407 and the bacteriostatic effect of CH were maintained. Mice full-thickness wounds treated with CH-P diminished the wound area during the first days. Consequently, with this treatment, increased levels of macrophages, α-SMA, and collagen deposition in wounds were observed, indicating a more mature scar tissue. In conclusion, the new hydrogel CH-P, at physiologic pH, combined the beneficial characteristics of both polymers and produced new properties for wound management. - Highlights: • ϒ-irradiation of chitosan + poloxamer 407 produced a hydrogel (CH-P) to wound care. • ϒ-irradiation allows chitosan (CH) solubility at physiological pH (CH-P 7). • CH-P 7 copolymer exhibits antimicrobial/antifungal features. • CH-P 7 hydrogel stimulates early wound-closure rate. • CH-P 7 increases collagen deposition and macrophage/fibroblasts recruitment.

Irradiation effects on organic insulators

International Nuclear Information System (INIS)

Kasen, M.B.

1986-01-01

The overall objective of this work is to contribute to development of organic insulators having the cryogenic neutron irradiation resistance required for MFE systems utilizing superconducting magnet confinement. The system for producing standard 3.2-mm (0.125-in) diameter rod specimens discussed in previous reports has been further refined to permit the fabrication of both fiber-reinforced and heat-resin specimens from hot-melt resin systems. The method has been successfully used to produce very high quality specimens duplicating the G-11CR system and specimens from a variant of that system eliminating a boron-containing additive. We have also produced specimens from an epoxy system suitable for impregnation or potting operations and from a bismaleimide polyimide system. These materials will be used in the first irradiation program in the National Low Temperature Neutron Irradiation Facility (NLTNIF) reactor at Oak Ridge. We have refined the 4-K torsional shear test method for evaluating radiation degradation of the fiber-matrix interface and have developed a method of quantitatively measuring changes in fracture energy as a function of radiation dose. Cooperative work with laboratories in Japan and England in this area is continuing and plans are being formulated for joint production, irradiation, and testing of specimens

Design of a radiation facility for very small specimens used in radiobiology studies

Science.gov (United States)

Rodriguez, Manuel; Jeraj, Robert

2008-06-01

A design of a radiation facility for very small specimens used in radiobiology is presented. This micro-irradiator has been primarily designed to irradiate partial bodies in zebrafish embryos 3-4 mm in length. A miniature x-ray, 50 kV photon beam, is used as a radiation source. The source is inserted in a cylindrical brass collimator that has a pinhole of 1.0 mm in diameter along the central axis to produce a pencil photon beam. The collimator with the source is attached underneath a computer-controlled movable table which holds the specimens. Using a 45° tilted mirror, a digital camera, connected to the computer, takes pictures of the specimen and the pinhole collimator. From the image provided by the camera, the relative distance from the specimen to the pinhole axis is calculated and coordinates are sent to the movable table to properly position the samples in the beam path. Due to its monitoring system, characteristic of the radiation beam, accuracy and precision of specimen positioning, and automatic image-based specimen recognition, this radiation facility is a suitable tool to irradiate partial bodies in zebrafish embryos, cell cultures or any other small specimen used in radiobiology research.

Transition Fracture Toughness Characterization of Eurofer 97 Steel using Pre-Cracked Miniature Multi-notch Bend Bar Specimens

Energy Technology Data Exchange (ETDEWEB)

Chen, Xiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clowers, Logan N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-11-01

In this report, we present the feasibility study of using pre-cracked miniature multi-notch bend bar specimens (M4CVN) with a dimension of 45mm (length) x 3.3mm (width) x 1.65mm (thickness) to characterize the transition fracture toughness of Eurofer97 based on the ASTM E1921 Master Curve method. From literature survey results, we did not find any obvious specimen size effects on the measured fracture toughness of unirradiated Eurofer97. Nonetheless, in order to exclude the specimen size effect on the measured fracture toughness of neutron irradiated Eurofer97, comparison of results obtained from larger size specimens with those from smaller size specimens after neutron irradiation is necessary, which is not practical and can be formidably expensive. However, limited literature results indicate that the transition fracture toughness of Eurofer97 obtained from different specimen sizes and geometries followed the similar irradiation embrittlement trend. We then described the newly designed experimental setup to be used for testing neutron irradiated Eurofer97 pre-cracked M4CVN bend bars in the hot cell. We recently used the same setup for testing neutron irradiated F82H pre-cracked miniature multi-notch bend bars with great success. Considering the similarity in materials, specimen types, and the nature of tests between Eurofer97 and F82H, we believe the newly designed experimental setup can be used successfully in fracture toughness testing of Eurofer97 pre-cracked M4CVN specimens.

Development of irradiation rig in HTTR and dosimetry method. I-I type irradiation equipment

Energy Technology Data Exchange (ETDEWEB)

Shibata, Taiju; Kikuchi, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi; Ogura, Kazutomo [Japan Atomic Power Co., Tokyo (Japan)

2002-12-01

The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium gas-cooled test reactor with a maximum power of 30 MW. The HTTR aims not only to establish and upgrade the technological basis for the HTGRs but also to perform the innovative basic research on high temperature engineering with high temperature irradiation fields. It is planned that the HTTR is used to perform various engineering tests such as the safety demonstration test, high temperature test operation and irradiation test with large irradiation fields at high temperatures. This paper describes the design of the I-I type irradiation equipment developed as the first irradiation rig for the HTTR and does the planned dosimetry method at the first irradiation test. It was developed to perform in-pile creep test on a stainless steel with large standard size specimens in the HTTR. It can give great loads on the specimens stably and can control the irradiation temperature precisely. The in-core creep properties on the specimens are measured by newly developed differential transformers and the irradiation condition in the core is monitored by thermocouples and self-powered neutron detectors (SPNDs), continuously. The irradiated neutron fluence is assessed by neutron fluence monitors of small metallic wires after the irradiation. The obtained data at the first irradiation test can strongly be contributed to upgrade the technological basis for the HTGRs, since it is the first direct measurement of the in-core irradiation environments of the HTTR. (author)

Specimen size effects in Charpy impact testing

International Nuclear Information System (INIS)

Alexander, D.J.; Klueh, R.L.

1989-01-01

Full-size , half-size, and third-size specimens from several different steels have been tested as part of an ongoing alloy development program. The smaller specimens permit more specimens to be made from small trail heats and are much more efficient for irradiation experiments. The results of several comparisons between the different specimen sizes have shown that the smaller specimens show qualitatively similar behavior to large specimens, although the upper-shelf energy level and ductile-to-ductile transition temperature are reduced. The upper-shelf energy levels from different specimen sizes can be compared by using a simple volume normalization method. The effect of specimen size and geometry on the ductile-to-ductile transition temperature is more difficult to predict, although the available data suggest a simple shift in the transition temperature due to specimen size changes.The relatively shallower notch used in smaller specimens alters the deformation pattern, and permits yielding to spread back to the notched surface as well as through to the back. This reduces the constraint and the peak stresses, and thus the initiation of cleavage is more difficult. A better understanding of the stress and strain distributions is needed. 19 refs., 3 figs., 3 tabs

Microstructure and mechanical properties of neutron irradiated OFHC-copper before and after post-irradiation annealing

International Nuclear Information System (INIS)

Singh, B.N.; Edwards, D.J.; Toft, P.

2001-02-01

Tensile specimens of OFHC-copper were irradiated with fission neutrons in the DR-3 reactor at Risoe National Laboratory at 100 deg. C to different displacement dose levels in the range of 0.01 to 0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 deg. C whereas other were given a post-irradiation annealing at 300 deg. C for 50 h and subsequently tested at 100 deg. C. Transmission electron microscopy was used to characterize the microstructure of specimens in the as-irradiation as well as irradiation and annealed conditions both before and after tensile deformation. The results show that while the interstitial loop microstructure coarsens with irradiation dose, no significant changes are observed in the population of stacking fault tetrahedra. The results also illustrates that the post-irradiation annealing leads to only a partial recovery and that the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the problem of yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade induced source hardening (CISH) and the dispersed barrier hardening (DBH) models. Both technological and scientific implications of these results are considered. (au)

Application of subsize specimens in nuclear plant life extension

International Nuclear Information System (INIS)

Rosinski, S.T.; Kumar, A.S.; Cannon, N.S.; Hamilton, M.L.

1993-01-01

The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missouri-Rolla to test a correlation for the upper shelf energy (USE) values obtained from the impact testing of subsize Charpy V-notch specimens to those obtained from the testing of full-size samples. The program involves the impact testing of unirradiated and irradiated full-, half-, and third-size Charpy V-notch specimens. To verify the applicability of the correlation on LWR materials, unirradiated and irradiated full-, half-, and third-size Charpy V-notch specimens of a commercial pressure vessel steel (ASTM A533 Grade B) will be tested. The correlation methodology is based on the partitioning of the USE into crack initiation and crack propagation energies. To accomplish this partition, both precracked and notched-only specimens will be used. Whereas the USE of notched-only specimens is the sum of both crack initiation and crack propagation energies, the USE of precracked specimens reflects only the crack propagation component. The difference in the USE of the two types of specimens represents a measure of the crack initiation energy. Normalizing the values of the crack initiation energy to the fracture volume of the sample produces similar values for the full-, half-, and third-size specimens. In addition, the ratios of the USE and the crack propagation energy are also in agreement for full-, half-, and third-size specimens. These two observations will be used to predict the USE of full-size specimens based on subsize USE data. This paper provides details of the program and presents results obtained from the application of the developed correlation methodology to the impact testing of the unirradiated full-, half-, and third-size A533 Grade B Charpy V-notch specimens

Crack-arrest tests on two irradiated high-copper welds

International Nuclear Information System (INIS)

Iskander, S.K.; Corwin, W.R.; Nanstad, R.K.

1994-03-01

The objective of the Heavy-Section Steel Irradiation Program Sixth Irradiation Series is to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest toughness data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288 degrees C to an average fluence of 1.9 x 10 19 neutrons/cm 2 (>1 MeV). This is the second report giving the results of the tests on irradiated duplex-type crack-arrest specimens. A previous report gave results of tests on irradiated weld-embrittled-type specimens. Charpy V-notch (CVN) specimens irradiated in the same capsules as the crack-arrest specimens were also tested, and a 41-J transition temperature shift was determined from these specimens. open-quotes Mean close-quote curves of the same form as the American Society of Mechanical Engineers (ASME) K la curve were fit to the data with only the open-quotes reference temperatureclose quotes as a parameter. The shift between the mean curves agrees well with the 41-J transition temperature shift obtained from the CVN specimen tests. Moreover, the four data points resulting from tests on the duplex crack-arrest specimens of the present study did not make a significant change to mean curve fits to either the previously obtained data or all the data combined

Neutron metrology in the HFR. Steel irradiation. R139-801 (SINAS)

International Nuclear Information System (INIS)

Ketema, D.J.

1999-02-01

The R139-80 series irradiation experiments is part of the NRG materials test programme to evaluate the irradiation behaviour of several types of austenitic stainless steel. Within this programme five R139-80 specimen holders were irradiated in the HFR Petten to different dose levels. This report presents the final metrology results obtained from activation monitors in a specimen holder, coded as R139-801, containing 12 Compact Tension (CT-10 mm) specimens made from the austenitic stainless steel types 308LSXB/TIG and 304-SXB. The R139-801 assembly was irradiated in channel 1 of a TRIO type facility placed in HFR core-position F8. The aim of this irradiation of specimen holder R139-801 was to reach a minimum target damage level of 7.5 dpa for the specimens at a temperature of 335C. The monitor sets are used to calculate the thermal and fast neutron fluences, displacements per atom and the generated helium content. Additionally detailed information concerning an estimation of the fluence and damage doses received by each specimen and its temperature during irradiation are presented. The main results of the thermal and fast neutron fluence measurements are presented. The results indicate that the obtained damage levels in the steel specimens loaded in this specimen holder vary from 5.8 to 7.9 dpa. The temperatures of the specimens during irradiation varied between 304 and 337C. 14 refs

Post Irradiation Mechanical Behaviour of Three EUROFER Joints

International Nuclear Information System (INIS)

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

2006-01-01

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

Isolation of chlamydia in irradiated and non-irradiated McCoy cells

International Nuclear Information System (INIS)

Johnson, L.; Harper, I.A.

1975-01-01

Specimens from eye and genital tract were cultured in parallel in irradiated and non-irradiated McCoy cells and the frequency of isolation of chlamydia using these culture methods was compared. There was a significant difference between the frequencies of isolation; irradiated McCoy cells produced a greater number of positive results. (author)

Preparation of TEM specimen by cross-section technique

International Nuclear Information System (INIS)

Hamada, Shozo

1986-01-01

Transmission electron microscopy (TEM) is applied to the direct observation of the depth dependent damage structure in ion-irradiated stainless steel by using the cross-section technique; obtaining the TEM specimen from a slice of the irradiated stainless steel with thick Ni plating. Here has been developed the specimen preparation method of cross-section technique without heat treatment, which was necessary in the conventional method to strengthen the bonding between Ni and stainless steel. Nickel plating with good bonding to stainless steel is enabled by the following manner. First, the irradiated stainless steel is immersed in the Wood's nickel solution at room temperature for 60s to activate the surface, followed by the stricking for 300s at a current density of 300 A/m 2 in the solution to make fine and homogeneous nucleation of Ni on the stainless steel. Then, the sample is plated with Ni in the Watt's nickel plating solution at 333 K with current density of 900 ∼ 1,000 A/m 2 . The TEM disc is obtained by mechanical slicing from the specimen with Ni plating of more than 3 mm thickness. Electropolishing is accomplished by using both Ballmann method and jet electropolishing to perforate the disc accurately at the aimed point for the observation of the damage structure. (author)

Influence of side-groove root radius on the ductile fracture toughness of miniature C(T) specimens

Energy Technology Data Exchange (ETDEWEB)

Lucon, E.; Scibetta, M.

2009-05-15

The use of miniature C(T) specimens, MC(T), for fracture toughness measurements in the upper shelf regime has been investigated at SCK-CEN since 2004, in the framework of the Electrabel/Tractebel SCK-CEN Convention (now General Framework Agreement SUEZ-SCK-CEN). This geometry has been used and validated on both unirradiated (2004-05) and irradiated (2006) materials, mainly reactor pressure vessel (RPV) steels. While side-grooved MC(T) specimens have shown in all conditions a systematically lower tearing resistance and ductile crack initiation toughness as compared to standard-size 1TC(T) samples, the only plain-sided MC(T) specimen tested in 2005 exhibited very high ductile fracture toughness, thus pointing at a strong influence of side-grooving on the upper shelf properties of MC(T) specimens. This study investigates the influence of side-grooving on the initiation toughness and tearing resistance of MC(T) specimens, as a function of the root radius of the side-groove (ranging from 0.1 to 1 mm) and in comparison with plain-sided MC(T) and reference 1TC(T) samples. The material used is the well characterized DIN 22NiMoCr37 RPV steel, which had been used in the European project which generated the famous EURO fracture toughness data set.

Creep test under irradiation with thermal gradient for the cylindrical carbon fiber reinforced carbon composite. Interim report on irradiation examinations: 03M-47AS

International Nuclear Information System (INIS)

Baba, Shin-ichi; Sawa, Kazuhiro; Yamaji, Masatoshi; Matsui, Yoshinori; Ishihara, Masahiro

2007-03-01

The creep test under irradiation with thermal gradient for the cylindrical carbon fiber reinforced carbon composites (c/c composite) are carried out in the Japan Material Testing Reactor (JMTR). This report described 4-items; first item is design/evaluation of the capsule for the irradiation test, second is before irradiation measurements for the residual strain due to manufactured cylindrical c/c composite, and third is also before irradiation measurements of the distance between 2-holes of predrilled in the specimen and last item is examination of analysis for the irradiation creep with thermal gradient by VIENUS Code. The normal creep test is static mechanical load on the specimen in thermal condition, but this creep test under irradiation capsule is thermal stress due to thermal gradient at inside specimen in the thermal condition. Consequently, it is necessary as large as possible thermal gradient in the narrow space of the capsule inside volume. In which the tungsten rod (W-rod) was inserted to the cylindrical c/c composite specimen, for γ-ray heat generation density occurred highly and so maximize the difference temperatures of surface wall between inside and outside wall of the specimen. The measurement method of the deflection due to the irradiation creep of cylindrical c/c composite was adopted as way of ruptured the specimen among the predrilled distance of 2-holes before/after irradiation. Accordingly as the laser dimensional apparatus used to measure the distance between the 2-holes of specimen exactly, easy and untouchable. And also before irradiation measurement of the residual stress due to the manufactured process was estimated by neutron diffraction used Residual Stress Analyzer (RESA) at JRR-3M in JAEA. The irradiation test was finished as total irradiation time, average temperature and neutron dose showed 4189 hours, 873 K and 8.2x10 24 (E>1.0MeV:m -2 ) respectively. The thermal stress was estimated by the difference temperatures of 4

Study of irradiation creep of vanadium alloys

Energy Technology Data Exchange (ETDEWEB)

Tsai, H.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

1997-08-01

Thin-wall tubing was produced from the 832665 (500 kg) heat of V-4 wt.% Cr-4 wt.% Ti to study its irradiation creep behavior. The specimens, in the form of pressurized capsules, were irradiated in Advanced Test Reactor and High Flux Isotope Reactor experiments (ATR-A1 and HFIR RB-12J, respectively). The ATR-A1 irradiation has been completed and specimens from it will soon be available for postirradiation examination. The RB-12J irradiation is not yet complete.

Accelerated irradiation test of Gundremmingen reactor vessel trepan material

Energy Technology Data Exchange (ETDEWEB)

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

1992-08-01

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

Accelerated irradiation test of gundremmingen reactor vessel trepan material

International Nuclear Information System (INIS)

Hawthorne, J.R.

1992-08-01

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

Investigation of irradiation embrittlement and annealing behaviour of JRQ pressure vessel steel by instrumented impact tests

Energy Technology Data Exchange (ETDEWEB)

Valo, M; Rintamaa, R; Nevalainen, M; Wallin, K; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Tipping, P [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1994-12-31

Seven series of A533-B type pressure vessel steel specimens irradiated as well as irradiated - annealed - re-irradiated to different fast neutron fluences (up to 5.10{sup 19}/cm{sup 2}) have been tested with a new type of instrumented impact test machine. The radiation embrittlement and the effect of the intermediate annealing was assessed by using the ductile and cleavage fracture initiation toughness. Although the ductile fracture initiation toughness exhibited scatter, the transition temperature shift corresponding to the dynamic cleavage fracture initiation agreed well with the 41 J Charpy-V shift. The results indicate that annealing is beneficial in restoring mechanical properties in an irradiated nuclear pressure vessel steel. (authors). 8 refs., 11 figs., 1 tab.

Effects of heat treatments and neutron irradiation on the physical and mechanical properties of copper alloys at 100 deg. C

International Nuclear Information System (INIS)

Singh, B.N.; Eldrup, M.; Toft, P.; Edwards, D.J.

1998-05-01

The final irradiation experiment in a series of screening experiments aimed at investigating the effects of bonding and bakeout thermal cycles on irradiated copper alloys is described herein. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment. Additional specimens were reaged and given a reactor bakeout treatment at 350 deg. C for 100 h. GlidCop TM CuAl-15 (previously referred to as CuAl-25) was given a heat treatment corresponding to a bonding thermal cycle only. Specimens were neutron irradiated at 100 deg. C to a dose level of ∼0.3 dpa. Post-irradiation tensile tests at (100 deg. C), electrical resistivity measurements (at 23 deg. C), and microstructural examinations were performed. The post-irradiation tests at 100 deg. C revealed that the greatest loss of ductility occurred in the CuCrZr alloys irradiated at 100 deg. C, irrespective of the pre-irradiation heat treatment, with the uniform elongation dropping to levels of less than 1.5%. The yield and ultimate strengths for all of the individual heat treated samples increased substantially after irradiation. The same trend was observed for the CuNiBe alloys, which exhibited much higher uniform elongation and strength after irradiation than that observed in the case of CuCrZr alloys. In both alloys irradiation-induced precipitation lead to a large increase in the strength of the solution annealed specimens with a noticeable decrease in uniform elongation. The CuAl-25 alloy also experienced an increase in the overall strength of the alloy after irradiation, accompanied by approximately a 50% decrease in the uniform and total elongation. The additional bakeout treatments given to the CuCrZr and CuNiBe before irradiation served to increase the strength, but in terms of the ductility no improvement or degradation resulted from the additional thermal exposure

Post-deformation examination of specimens subjected to SCC testing

Energy Technology Data Exchange (ETDEWEB)

Gussev, Maxim N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Busby, Jeremy T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Leonard, Keith J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-01

This report details the results of post-radiation and post-deformation characterizations performed during FY 2015–FY 2016 on a subset of specimens that had previously been irradiated at high displacement per atom (dpa) damage doses. The specimens, made of commercial austenitic stainless steels and alloys, were subjected to stress-corrosion cracking tests (constant extension rate testing and crack growth testing) at the University of Michigan under conditions typical of nuclear power plants. After testing, the specimens were returned to Oak Ridge National Laboratory (ORNL) for further analysis and evaluation.

Mechanical and irradiation properties of zirconium alloys irradiated in HANARO

International Nuclear Information System (INIS)

Kwon, Oh Hyun; Eom, Kyong Bo; Kim, Jae Ik; Suh, Jung Min; Jeon, Kyeong Lak

2011-01-01

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.1 10 21 n/cm 2 ). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed

Fracture toughness measurements with subsize disk compact specimens

International Nuclear Information System (INIS)

Alexander, D.J.

1992-01-01

Special fixtures and test methods have been developed for testing small disk compact specimens (12.5 mm diam by 4.6 mm thick). Both unloading compliance and potential drop methods have been used to monitor crack extension during the J-integral resistance (J-R) curve testing. Provisions have been made to allow the necessary probes and instrumentation to be installed remotely using manipulators for testing of irradiated specimens in a hat cell. Laboratory trials showed that both unloading compliance and potential drop gave useful results. Both techniques gave similar data, and predicted the final crack extension within allowable limits. The results from the small disk compact specimens were similar to results from conventional compact specimens 12.7-mm thick. However, the slopes of the J-R curves from the larger specimens were lower, suggesting that the smaller disk compact specimens may have lost some constraint due to their size. The testing shows that it should be possible to generate useful J-R curve fracture toughness data from the small disk compact specimens

Use of miniaturized compact tension specimens for fracture toughness measurements in the upper shelf regime. Electrabel/Tractebel-SCK-CEN Convention 2004 Task 1.1.4/2

International Nuclear Information System (INIS)

Lucon, E.; Scibetta, M.; Chaouadi, R.; Walle, E. van

2005-04-01

In the nuclear field, the importance of direct fracture toughness measurements on RPV materials has been nowadays widely recognized, as opposed to Charpy-based estimations. However, sample dimensions have to be kept small in order to optimize the use of available material (often in the form of previously broken Charpy specimens) or, in the case of new irradiations, make effective use of the limited space available inside irradiation facilities. One of the most appealing geometries for fracture toughness measurements is the miniature Compact Tension specimen, MC(T), which has the following dimensions: B = 4.15 mm, W = 8.3 mm, cross section 10 x 10 mm 2 . Four MC(T) specimens can be machined out of a broken half Charpy, and in the case of irradiation ten MC(T) samples occupy approximately the same volume as a full-size Charpy specimen. The MC(T) geometry was already successfully applied and qualified for fracture toughness assessments in the ductile-to-brittle transition regime, using the Master Curve method (ASTM E1921-03). A further, comprehensive investigation is presented in this report, aimed at assessing the applicability of MC(T) specimens to measure fracture toughness in fully ductile (upper shelf) conditions. In this study, 18 1TC(T) and 20 MC(T) specimens have been tested at different temperatures from three RPV steels and one low-alloy C-Mn steel. The results obtained clearly show that MC(T) samples exhibit lower fracture toughness properties, both in terms of initiation of ductile tearing (according to various test standards) and resistance to ductile crack propagation (J-R curve). The reduction of tearing resistance might be attributed to work hardening prevailing over loss of constraint in the uncracked ligament in a side-grooved specimen, or to the inadequacy of J-integral to represent ductile crack extension in very small specimens. Both arguments will have to be verified with further investigations. (author)

Investigation on the effects of gamma irradiation on bitumen

International Nuclear Information System (INIS)

Mello, M.S.; Braz, D.; Motta, L.M.G.

2011-01-01

Brazil has more than 218,000 km of asphalt-paved highways. Bitumen is a generic term for natural or manufactured black or dark-colored solid, semisolid, or viscous cementitious materials that are composed mainly of high molecular weight hydrocarbons (90-95%). Several papers have shown that the irradiation process has changed the mechanical behavior in some polymers. This work aims to analyze the behavior of Brazilian irradiated Bitumen (CAP 50-70). In order to provide a preliminary evaluation, bitumen samples and cylindrical specimens of asphaltic mixture were tested. The bitumen samples were irradiated 0.1 to 300 kGy, and asphaltic mixture specimen was irradiated 5 to 300 kGy. The cylindrical asphaltic mixture specimen of 10.16 cm diameter used in this study was molded using an asphalt-aggregate mixture. The specimens were irradiated in LIN/UFRJ/Brazil using a Gamma cell Co 60 source of gamma irradiation with an applied dose rate of 29.7 Gy/min. After irradiated, the bitumen samples were subjected to penetration test and the asphaltic mixtures were subjected to indirect tensile strength test (diametral compression) for determination of the resilient modulus, according to ASTM method D 4123. The results of these experiments for each dose were compared with the control (nonirradiated). As expected, the penetration results showed that the ratio (irradiated/non-irradiated) decreases with increasing of irradiation dose for bitumen samples and the resilient modulus results showed that the ratio (irradiated/non-irradiated) increases with increasing of irradiation dose for asphaltic mixture. (author)

Investigations of X-ray irradiation of marine fish aboard

International Nuclear Information System (INIS)

Karnop, G.; Reinacher, E.; Antonacopoulos, N.; Meyer, V.

1976-01-01

Studies on X-ray irradiation of ocean perch, cod and coley (at doses of 50-150 krad) are described. The results show that irradiation within this dose range has no significant effect on the shelf-life of fish stored in ice. Although irradiation positively influenced bacteriological and chemical characteristics (e.g. reduction of total aerobic count, and inhibition of decomposition of N-containing compounds), the organoleptically-limited shelf-life of irradiated specimens was similar to that of non-irradiated specimens. Organoleptic changes in irradiated and in non-irradiated samples differed; this is attributed to the abnormal spoilage flora (mainly radiation-resistant Moraxella spp.) in the irradiated samples. (orig./HP) [de

Multispecimen dual-beam irradiation damage chamber

International Nuclear Information System (INIS)

Packan, N.H.; Buhl, R.A.

1980-06-01

An irradiation damage chamber that can be used to rapidly simulate fast neutron damage in fission or fusion materials has been designed and constructed. The chamber operates in conjunction with dual Van de Graaff accelerators at ORNL to simulate a wide range of irradiation conditions, including pulsed irradiation. Up to six experiments, each with up to nine 3-mm disk specimens, can be loaded into the ultrahigh vacuum chamber. Specimen holders are heated with individual electron guns, and the temperature of each specimen can be monitored during bombardment by an infrared pyrometer. Three different dose levels may be obtained during any single bombardment, and the heavy-ion flux on each of the nine specimens can be measured independently with only a brief interruption of the beam. The chamber has been in service for nearly three years, during which time approximately 250 bombardments have been successfully carried out. An appendix contains detailed procedures for operating the chamber

Irradiation effects on fracture toughness of two high-copper submerged-arc welds, HSSI Series 5

International Nuclear Information System (INIS)

Nanstad, R.K.; Haggag, F.M.; McCabe, D.E.; Iskander, S.K.; Bowman, K.O.; Menke, B.H.

1992-10-01

The Fifth Irradiation Series in the Heavy-Section Steel Irradiation Program obtained a statistically significant fracture toughness data base on two high-copper (0.23 and 0.31 wt %) submerged-arc welds to determine the shift and shape of the K Ic curve as a consequence of irradiation. Compact specimens with thicknesses to 101.6 mm (4 in) in the irradiated condition and 203.2 mm (8 in) in the unirradiated condition were tested, in addition to Charpy impact, tensile, and drop-weight specimens. Irradiations were conducted at a nominal temperature of 288 degree C and an average fluence of 1.5 x 10 19 neutrons/cm 2 (>l MeV). The Charpy 41-J temperature shifts are about the same as the corresponding drop-weight NDT temperature shifts. The irradiated welds exhibited substantial numbers of cleavage pop-ins. Mean curve fits using two-parameter (with fixed intercept) nonlinear and linearized exponential regression analysis revealed that the fracture toughness 100 MPa lg-bullet √m shifts exceeded the Charpy 41-J shifts for both welds. Analyses of curve shape changes indicated decreases in the slopes of the fracture toughness curves, especially for the higher copper weld. Weibull analyses were performed to investigate development of lower bound curves to the data, including the use of a variable K min parameter which affects the curve shape

Irradiation hardening and localized deformation of neutron-irradiated α-iron single crystals

International Nuclear Information System (INIS)

Mughrabi, H.; Stroehle, D.; Wilkens, M.

1981-01-01

The early yielding behaviour of neutron-irradiated α iron single crystals orientated for single slip was investigated as a function of neutron dose. In the range of neutron doses between approx. equal to 10 18 and approx. equal to 10 19 n/cm 2 , the irradiation hardening increment was found to be almost constant. Qualitative modifications of this behaviour were observed in the case of predeformed specimens. The localized deformation of the neutron-irradiated specimens by dislocation channelling was investigated by slip-line observations, transmission electron microscopy and X-ray topography. A model of localized deformation is proposed in order to explain the development of the observed asymmetric dislocation double layers which bound the channels and transmit characteristic misorientations. (orig.)

Elemental microanalysis of biological and medical specimens with a scanning proton microprobe

International Nuclear Information System (INIS)

Legge, G.J.F.; Mazzolini, A.P.

1979-01-01

The scanning proton microprobe is shown to be a sensitive instrument for elemental microanalysis of cells and tissues in biological and medical specimens. The preparation of specimens and their behaviour under irradiation are crucial and the application of quantitative scanning analysis to the monitoring of such problems is illustrated

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

Science.gov (United States)

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

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

Determination of the mechanical characteristics of irradiated metals from the results of microhardness tests

International Nuclear Information System (INIS)

Hofman, A.

1999-01-01

To predict the possibilities of using structural materials in nuclear and thermonuclear reactors, it is important to have data on changes of the mechanical characteristics and irradiation obtained from full-scale or simulation tests. Materials are irradiated in nuclear reactors with fast neutrons, the sources of high-energy neutrons with an energy of 14 MeV and the accelerators of charged particles. The restricted volumes for irradiation of these specimens in the systems and also the need to test large numbers of specimens under the same conditions make it necessary to reduce the size of irradiated specimens. To solve this problem, work is being carried out to develop various methods of testing miniature specimens, including tension extrusion of disc-shaped micro-specimens, microhardness, and the Charpy Method. In examination of the irradiation hardening of the materials, the main advantage of the microhardness method is that it makes it possible to examine small specimens. In single microhardness tests, only a small area of the irradiated specimens is examined. This makes it possible to increase the radiation dose and carry out subsequent tests of microhardness on the same specimens. The aim of this work was to determine the possibilities of using the microhardness measurement method for evaluating the mechanical characteristics of metallic materials. The comparison of the data, obtained in microhardness tests and in tensile loading specimens of 0Kh18N10Tsteel, irradiated with neutrons, shows the efficiency of the microhardness method as a tool for investigating the irradiation hardening of reactor materials

Irradiation Test in HANARO of the Parts of an X-Gen Nuclear Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

2008-08-15

An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of an X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens requested by Westinghouse Co. and Hanyang university were also inserted. 389 KNF specimens such as bucking and spring test specimens of 1x1 cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718 were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of Ni-Ti-Fe (2 sets contain additional Nb-Ag) neutron fluence monitors installed in the capsule. The capsule was irradiated for 59.19days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 300 {approx} 420 .deg. C(for KNF specimens) up to a fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1MeV). After an irradiation test, the main body of the capsule was cut off at the bottom of the protection tube with a cutting system and it was transported to the IMEF (Irradiated Materials Examination Facility). The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell.

Remote replacement of materials open-test assembly specimens at the FFTF/IEM cell

International Nuclear Information System (INIS)

Gibbons, P.W.; Ramsey, E.B.

1990-01-01

The Fast Flux Test Facility (FFTF) interim examination and maintenance (IEM) cell is used for the remote disassembly of irradiated fuel and materials experiments. The materials open-test assembly (MOTA) is brought to the IEM cell for materials test specimen removal. The specimens are shipped to the materials laboratory for sorting and installation in new specimen holders and then returned within 10 days to the IEM cell where they are installed in a new MOTA vehicle for further irradiation. Reconstituting a MOTA is a challenging remote operation involving dozens of steps and two separate facilities. Handling and disassembling sodium-wetted components pose interesting handling, cleaning, and disposal challenges. The success of this system is evidenced by its timely completion in the critical path of FFTF outage schedules

Some recent innovations in small specimen testing

International Nuclear Information System (INIS)

Odette, G.R.; He, M.; Gragg, D.; Klingensmith, D.; Lucas, G.E.

2002-01-01

New innovative small specimen test techniques are described. Finite element simulations show that combinations of cone indentation pile-up geometry and load-penetration depth relations can be used to determine both the yield stress and strain-hardening behavior of a material. Techniques for pre-cracking and testing sub-miniaturized fracture toughness bend bars, with dimensions of 1.65x1.65x9 mm 3 , or less, are described. The corresponding toughness-temperature curves have a very steep transition slope, primarily due to rapid loss of constraint, which has advantages in some experiments to characterize the effects of specified irradiation variables. As one example of using composite specimens, an approach to evaluating helium effects is proposed, involving diffusion bonding small wires of a 54 Fe-based ferritic-martensitic alloy to a surrounding fracture specimen composed of an elemental Fe-based alloy. Finally, we briefly outline some potential approaches to multipurpose specimens and test automation

KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; Morris, Robert N.

2016-11-01

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of

Effects of residual stress on irradiation hardening in stainless steels

Energy Technology Data Exchange (ETDEWEB)

Okubo, N.; Kondo, K.; Kaji, Y. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Miwa, Y. [Nuclear Energy and Science Directorate, Japan Atomic Energy Agency, Tokai-mura, Ibaraki-ken (Japan)

2007-07-01

Full text of publication follows: Structural materials in fusion reactor with water cooling system will undergo corrosion in aqueous environment and heavier irradiation than that in LWR. Irradiation assisted stress corrosion (IASCC) may be induced in stainless steels exposed in these environment for a long term of reactor operation. The IASCC is considered to be caused in a welding zone. It is difficult to predict and estimate the IASCC, because several irradiation effects (irradiation hardening, swelling, irradiation induced stress relaxation, etc) work intricately. Firstly, effects of residual stress on irradiation hardening were investigated in stainless steels. Specimens used in this study were SUS316 and SUS316L. By bending deformation, the specimens with several % plastic strain, which corresponds to weld residual stress, were prepared. Ion irradiations of 12 MeV Ni{sup 3+} were performed at 330, 400 and 550 deg. C to 45 dpa in TIARA facility at JAEA. No bent specimen was simultaneously irradiated with the bent specimen. The residual stress was estimated by X-ray residual stress measurements before and after the irradiation. The micro-hardness was measured by using nano-indenter. The irradiation hardening and the stress relaxation were changed by irradiation under bending deformation. The residual stress did not relax even for the case of the higher temperature aging at 500 deg. C for the same time of irradiation. The residual stress after ion irradiation, however, relaxed at these experimental temperatures in SUS316L. The hardness was obviously suppressed in bent SUS316L irradiated at 300 deg. C to 6 or 12 dpa. It was evident that irradiation induced stress relaxation occasionally suppressed the irradiation hardening in SUS316L. (authors)

Evaluation of the mechanical properties of carbon fiber after electron beam irradiation

International Nuclear Information System (INIS)

Giovedi, Claudia; Diva Brocardo Machado, Luci; Augusto, Marcos; Segura Pino, Eddy; Radino, Patricia

2005-01-01

Carbon fibers are used as reinforcement material in epoxy matrix in advanced composites. An important aspect of the mechanical properties of composites is associated to the adhesion between the surface of the carbon fiber and the epoxy matrix. This paper aimed to the evaluation of the effects of EB irradiation on the tensile properties of two different carbon fibers prepared as resin-impregnated specimens. The fibers were EB irradiated before the preparation of the resin-impregnated specimens for mechanical tests. Observations of the specimens after breakage have shown that EB irradiation promoted significant changes in the failure mode. Furthermore, the tensile strength data obtained for resin-impregnated specimens prepared with carbons fibers previously irradiated presented a slight tendency to be higher than those obtained from non-irradiated carbon fibers

A Study on Mechanical behavior of Tensile Specimen Fabricated by Laser Cutting

Energy Technology Data Exchange (ETDEWEB)

Jin, Y. G.; Kim, G. S.; Baik, S. J.; Baek, S. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The mechanical testing data are required for the assessment of dry storage of the spent nuclear fuel. Laser cutting system could be useful tools for material processing such as cutting in radioactive environment due to non-contact nature, ease in handling and the laser cutting process is most advantageous, offering the narrow kerf width and heat affected zone by using small beam spot diameter. The feasibility of the laser cutting system was demonstrated for the fabrication of various types of the unirradiated cladding with and without oxide layer on the specimens. In the present study, the dimensional measurement and tensile test were conducted to investigate the mechanical behavior of the axial tensile test specimens depending on the material processing methods in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser cutting system was used to fabricate the tensile test specimens, and the mechanical behavior was investigated using the dimensional measurement and tensile test. It was shown that the laser beam machining could be a useful tool to fabricate the specimens and this technique will be developed for the fabrication of various types of irradiated specimens in a hotcell.

A Study on Mechanical behavior of Tensile Specimen Fabricated by Laser Cutting

International Nuclear Information System (INIS)

Jin, Y. G.; Kim, G. S.; Baik, S. J.; Baek, S. Y.

2016-01-01

The mechanical testing data are required for the assessment of dry storage of the spent nuclear fuel. Laser cutting system could be useful tools for material processing such as cutting in radioactive environment due to non-contact nature, ease in handling and the laser cutting process is most advantageous, offering the narrow kerf width and heat affected zone by using small beam spot diameter. The feasibility of the laser cutting system was demonstrated for the fabrication of various types of the unirradiated cladding with and without oxide layer on the specimens. In the present study, the dimensional measurement and tensile test were conducted to investigate the mechanical behavior of the axial tensile test specimens depending on the material processing methods in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser cutting system was used to fabricate the tensile test specimens, and the mechanical behavior was investigated using the dimensional measurement and tensile test. It was shown that the laser beam machining could be a useful tool to fabricate the specimens and this technique will be developed for the fabrication of various types of irradiated specimens in a hotcell

Defects annihilation behavior of neutron-irradiated SiC ceramics densified by liquid-phase-assisted method after post-irradiation annealing

Directory of Open Access Journals (Sweden)

Mohd Idzat Idris

2016-12-01

Full Text Available Numerous studies on the recovery behavior of neutron-irradiated high-purity SiC have shown that most of the defects present in it are annihilated by post-irradiation annealing, if the neutron fluence is less than 1×1026 n/m2 (>0.1MeV and the irradiation is performed at temperatures lower than 973K. However, the recovery behavior of SiC fabricated by the nanoinfiltrated and transient eutectic phase (NITE process is not well understood. In this study, the effects of secondary phases on the irradiation-related swelling and recovery behavior of monolithic NITE-SiC after post-irradiation annealing were studied. The NITE-SiC specimens were irradiated in the BR2 reactor at fluences of up to 2.0–2.5×1024 n/m2 (E>0.1MeV at 333–363K. This resulted in the specimens swelling up ∼1.3%, which is 0.1% higher than the increase seen in concurrently irradiated high-purity SiC. The recovery behaviors of the specimens after post-irradiation thermal annealing were examined using a precision dilatometer; the specimens were heated at temperatures of up to 1673K using a step-heating method. The recovery curves were analyzed using a first-order model, and the rate constants for each annealing step were obtained to determine the activation energy for volume recovery. The NITE-A specimen (containing 12 wt% sintering additives recovered completely after annealing at ∼1573K; however, it shrank because of the volatilization of the oxide phases at 1673K. The NITE-B specimen (containing 18wt% sintering additives did not recover fully, since the secondary phase (YAG was crystallized during the annealing process. The recovery mechanism of NITE-A SiC was based on the recombination of the C and Si Frenkel pairs, which were very closely sited or only slightly separated at temperatures lower than 1223K, as well as the recombination of the slightly separated C Frenkel pairs and the migration of C and Si interstitials at temperatures of 1223–1573K. That is to say, the

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

International Nuclear Information System (INIS)

Allen, C.W.

1992-01-01

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

TEM characterization of irradiated microstructure of Fe-9%Cr ODS and ferritic-martensitic alloys

Science.gov (United States)

Swenson, M. J.; Wharry, J. P.

2018-04-01

The objective of this study is to evaluate the effects of irradiation dose and dose rate on defect cluster (i.e. dislocation loops and voids) evolution in a model Fe-9%Cr oxide dispersion strengthened steel and commercial ferritic-martensitic steels HCM12A and HT9. Complimentary irradiations using Fe2+ ions, protons, or neutrons to doses ranging from 1 to 100 displacements per atom (dpa) at 500 °C are conducted on each alloy. The irradiated microstructures are characterized using transmission electron microscopy (TEM). Dislocation loops exhibit limited growth after 1 dpa upon Fe2+ and proton irradiation, while any voids observed are small and sparse. The average size and number density of loops are statistically invariant between Fe2+, proton, and neutron irradiated specimens at otherwise fixed irradiation conditions of ∼3 dpa, 500 °C. Therefore, we conclude that higher dose rate charged particle irradiations can reproduce the neutron irradiated loop microstructure with temperature shift governed by the invariance theory; this temperature shift is ∼0 °C for the high sink strength alloys studied herein.

Specimen preparation of irradiated materials for examination in the atom probe field ion microscope

International Nuclear Information System (INIS)

Russell, K.F.; Miller, M.K.

1994-01-01

The atom probe field ion microscope (APFIM) requires specimens in the form of ultrasharp needles. Basic protective measures used to reduce exposure druing specimen preparation are discussed. The low-level radioactive specimen blanks may be made using a two-stage electropolishing process using a thin layer of electrolyte floating on a denser inert liquid; this produces a necked region and eventually two specimens from each single blank. The amount of material handled may also be reduced using a micropolishing technique to repolish blunt or fractured specimens. Control of contamination and possible spills is discussed

Elemental microanalysis of botanical specimens using the Melbourne Proton Microprobe

International Nuclear Information System (INIS)

Mazzolini, A.P.J.; Legge, G.J.F.

1978-01-01

A proton microprobe has been used to obtain the distribution of elements of various botanical specimens. This paper presents preliminary results obtained by the irradiation of certain organs of the wheat plant

In-situ SCC observation on thermally-sensitized type 304 stainless steel irradiated to 1 x 10{sup 25} n/m{sup 2}

Energy Technology Data Exchange (ETDEWEB)

Nakano, J.; Nemoto, Y.; Tsukada, T.; Usami, K. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hide, K. [Central Research Institute of Electric Power Industry, Yokosuka-shi, Kanagawa-ken (Japan)

2007-07-01

Full text of publication follows: Irradiation assisted stress corrosion cracking (IASCC) is concerned as being one of the specific problems for water-cooled first wall/blanket components in the design activity of international thermonuclear experimental reactor (ITER). To examine the crack initiation and growth behaviors of IASCC, in-situ observation on gage length of specimens was conducted during slow strain rate tests (SSRT) in high temperature water. Results from in-situ observation on Type 304 stainless steel (SS) irradiated to 1.0 x 10{sup 26} n/m{sup 2} have been reported already. Type 304 SS was subjected to a solution annealing (SA), a thermally sensitization (TS) or a cold working (CW, 20%) and irradiated to 1.0 x 10{sup 25} n/m{sup 2} in the Japan Materials Testing Reactor (JMTR). After neutron irradiation, SSRT for the specimens was conducted in oxygenated high purity water at 561 K. The gage length of the specimen was observed through a window equipped on an autoclave during the SSRT. Subsequently, fracture surface examination was performed using a scanning electron microscope (SEM). In fracture surface examination of the specimens irradiated to 1.0 x 10{sup 25} n/m{sup 2}, almost entire intergranular stress corrosion cracking (IGSCC) was exhibited for the TS material while mixtures of transgranular stress corrosion cracking (TGSCC) and ductile dimple fracture were observed for the SA and the CW materials. Although crack initiation was observed immediately after maximum stress for the CW irradiated to 1.0 x 10{sup 26} n/m{sup 2}, crack initiation was observed immediately before maximum stress (99% of maximum stress) for the CW irradiated to 1.0 x 10{sup 25} n/m{sup 2} in in-situ observation. (authors)

The two types of a loss of sight (blindness) exhibited by cats and dogs after local irradiation of heat

International Nuclear Information System (INIS)

Ushakov, I.B.; Razgovorov, B.L.

1985-01-01

After local irradiation of heads with doses of 50 to 100 Gy cats and dogs exhibited two types of a loss of sight: early blindness (during the first two hours) noted only in cats after a dose of 100 Gy, and delayed blindness in cats after a dose of 50 Gy, and in dogs after all doses under study

Influence of irradiance on Knoop hardness, degree of conversion, and polymerization shrinkage of nanofilled and microhybrid composite resins.

Science.gov (United States)

Fugolin, Ana Paula Piovezan; Correr-Sobrinho, Lourenço; Correr, Américo Bortolazzo; Sinhoreti, Mário Alexandre Coelho; Guiraldo, Ricardo Danil; Consani, Simonides

2016-01-01

The purpose of this study was to investigate the influence of the irradiance emitted by a light-curing unit on microhardness, degree of conversion (DC), and gaps resulting from shrinkage of 2 dental composite resins. Cylinders of nanofilled and microhybrid composites were fabricated and light cured. After 24 hours, the tops and bottoms of the specimens were evaluated via indentation testing and Fourier transform infrared spectroscopy to determine Knoop hardness number (KHN) and DC, respectively. Gap width (representing polymerization shrinkage) was measured under a scanning electron microscope. The nanofilled composite specimens presented significantly greater KHNs than did the microhybrid specimens (P composite resin exhibited significantly greater DC and gap width than the nanofilled material (P composite resins.

Fracture toughness evaluation of Eurofer'97 by testing small specimens

International Nuclear Information System (INIS)

Serrano, M.; Fernandez, P.; Lapena, J.

2006-01-01

The Eurofer'97 is the structural reference material that will be tested in the ITER modules. Its metallurgical properties have been well characterized during the last years. However, more investigations related with the fracture toughness of this material are necessary because this property is one of the most important to design structural components and to study their integrity assessment. In the case of structural materials for fusion reactor the small specimen technology (SSTT) are being actively developed to investigate the fracture toughness among other mechanical properties. The use of small specimens is due to the small available irradiation volume of IFMIF and also due to the high fluence expected in the fusion reactor. The aim of this paper is to determine the fracture toughness of the Eurofer'97 steel by testing small specimens of different geometry in the ductile to brittle transition region, with the application of the Master Curve methodology, and to evaluate this method to assess the decrease in fracture toughness due to neutron irradiation. The tests and data analysis have been performed following the Master Curve approach included in the ASTM Standard E1921-05. Specimen size effect and comparison of the fracture toughness results with data available in the literature are also considered. (author)

Temper embrittlement, irradiation induced phosphorus segregation and implications for post-irradiation annealing of reactor pressure vessels

International Nuclear Information System (INIS)

McElroy, R.J.; English, C.A.; Foreman, A.J.; Gage, G.; Hyde, J.M.; Ray, P.H.N.; Vatter, I.A.

1999-01-01

Three steels designated JPB, JPC and JPG from the IAEA Phase 3 Programme containing two copper and phosphorus levels were pre- and post-irradiation Charpy and hardness tested in the as-received (AR), 1200 C/0.5h heat treated (HT) and heat treated and 450 C/2000h aged (HTA) conditions. The HT condition was designed to simulate coarse grained heat-affected zones (HAZ's) and showed a marked sensitivity to thermal ageing in all three alloys. Embrittlement after thermal ageing was greater in the higher phosphorus alloys JPB and JPG. Charpy shifts due to thermal ageing of between 118 and 209 C were observed and accompanied by pronounced intergranular fracture, due to phosphorus segregation. The irradiation embrittlement response was complex. The low copper alloys, JPC and JPB, in the HT and HTA condition exhibited significant irradiation induced Charpy shift but very low or even negative hardness changes indicating non-hardening embrittlement. The higher copper alloy, JPG, also exhibited irradiation hardening in line with its copper content. Fractographic and microchemical studies indicated irradiation induced phosphorus segregation and a transition from cleavage to intergranular failure at grain boundary phosphorus concentrations above a critical level. The enhanced grain boundary phosphorus level increased with dose in agreement with a kinetic segregation model developed at Harwell. The relevance of the thermal ageing studies to RPV Annealing for Plant-Life Extension was identified early in the program. It is of concern that annealing of RPV's has been performed, or is proposed, at temperatures in the range 425--475 C for periods of about 1 week (168h). Much attention has been given to the use of in-situ hardness measurements and machining miniature Charpy and tensile specimens from belt-line plate and weld materials. However, HAZ's, often containing higher phosphorus levels than the present materials, have largely been ignored. A post-irradiation annealing (PIA

Irradiation Effects at 160-240 deg C in Some Swedish Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

Grounes, M [AB Atomenergi, Nykoeping (Sweden); Myers, H P [Chalmers Institute of Technology, Goeteborg (Sweden); Hannerz, N E [Motala Verkstads AB, Motala (Sweden)

1967-09-15

Tensile specimens, Charpy impact specimens and miniature impact specimens of six steels in different conditions were irradiated to 2.8 x 10{sup 18} and 5.6 x 10{sup 18} n/cm{sup 2} (> 1 MeV) at 160-240 deg C. The steels investigated were SIS 142103, 2103/R3, NO 345, Fortiweld, Fortiweld HS and OK 54 P. There is no correlation between the increase in transition temperature and initial transition temperature. However, changes in strength and ductility can be correlated to the initial yield strength. The increases in transition temperature due to strain aging and irradiation are approximately additive. The irradiation-induced changes in 2103/R3 and Fortiweld HS steels do not vary with position in the thickness of the plate. Different tempering treatments in Fortiweld HS steel do not change the extent of irradiation effects. Normal Charpy V-notch impact specimens and miniature specimens give the same irradiation-induced increase in transition temperature.

Irradiation Effects at 160-240 deg C in Some Swedish Pressure Vessel Steels

International Nuclear Information System (INIS)

Grounes, M.; Myers, H.P.; Hannerz, N.E.

1967-09-01

Tensile specimens, Charpy impact specimens and miniature impact specimens of six steels in different conditions were irradiated to 2.8 x 10 18 and 5.6 x 10 18 n/cm 2 (> 1 MeV) at 160-240 deg C. The steels investigated were SIS 142103, 2103/R3, NO 345, Fortiweld, Fortiweld HS and OK 54 P. There is no correlation between the increase in transition temperature and initial transition temperature. However, changes in strength and ductility can be correlated to the initial yield strength. The increases in transition temperature due to strain aging and irradiation are approximately additive. The irradiation-induced changes in 2103/R3 and Fortiweld HS steels do not vary with position in the thickness of the plate. Different tempering treatments in Fortiweld HS steel do not change the extent of irradiation effects. Normal Charpy V-notch impact specimens and miniature specimens give the same irradiation-induced increase in transition temperature

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Production of iodine-123 radiobiological specimen on 25 MeV electron beam

International Nuclear Information System (INIS)

Oganesyan, Yu.Ts.; Starodub, G.Ya.; Buklanov, G.V.; Korotkin, Yu.S.; Belov, A.G.

1988-01-01

The technique is described and experimental results are presented for production of radioactive specimen-iodine-123 for medical biological investigations. It is shown that in ten hour irradiation of 124 Xe enriched target of 10 g weight by the 25 MeV electron beam at MT-25 microtron short lived 123 I with activity of about 200 mCl can be accumulated. The procedure was developed for extraction of radioactive atoms and preparing the solution that permits to obtain during 1-1.5 h after the end of irradiation the specimen which satisfies all pharmacopeia requirements. It follows from the results that using small-size electron accelerators with the beam energy up to 25 MeV permits to organize economical and large-scale production of high quality radioactive specimen of 123 I for servicing a large region of this country. 14 refs.; 4 figs.; 1 tab

Irradiation induced tensile property change of SA 508 Cl.3 reactor pressure vessel steels

International Nuclear Information System (INIS)

Chi, Se-Hwan; Hong, Jun-Hwa; Kuk, Il-Hiun

1998-01-01

Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the unirradiated and irradiated microstructure. Microvickers hardness, indentation, and miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were 2 irradiated to a neutron fluence of 2.7x10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg. C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Band-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural. state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation(VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by

Structural changes induced by electron irradiation

International Nuclear Information System (INIS)

Koike, J.; Pedraza, D.F.

1993-01-01

Highly oriented pyrolytic graphite was irradiated at room temperature with 300 kV electrons. Transmission electron microscopy and electron energy loss spectroscopy were employed to study the structural changes produced by irradiation. The occurrence of a continuous ring intensity in the selected area diffraction (SAD) pattern obtained on a specimen irradiated with the electron beam parallel to the c-crystallographic axis indicated that microstructural changes had occurred. However, from the SAD pattern obtained for the specimens tilted relative to the irradiation direction, it was found that up to a fluence of 1.1x10 27 e/m 2 graphite remained crystalline. An SAD pattern of a specimen irradiated with the electron beam perpendicular to the c-axis confirmed the persistence of crystalline order. High resolution electron microscopy showed that ordering along the c-axis direction remained. A density reduction of 8.9% due to irradiation was determined from the plasmon frequency shift. A qualitative model is proposed to explain these observations. A new determination of the threshold displacement energy, Ed, of carbon atoms in graphite was done by examining the appearance of a continuous ring in the SAD pattern at various electron energies. A value of 30 eV was obtained whether the incident electron beam was parallel or perpendicular to the c-axis, demonstrating that Ed is independent of the displacement direction

Significance of endoscopic biopsy after preoperative irradiation therapy for rectal cancer

Energy Technology Data Exchange (ETDEWEB)

Takiguchi, Nobuhiro; Sarashina, Hiromi; Saito, Norio; Nunomura, Masao; Kohda, Keishi; Nakajima, Nobuyuki (Chiba Univ. (Japan). School of Medicine)

1994-05-01

To evaluate the utility of endoscopic biopsy before and after preoperative irradiation therapy for rectal cancer, we examined histologically both biopsy specimens and resected materials of forty-three patients. Two pieces of biopsy materials were taken both before and after irradiation therapy (total dose 42.6 Gy) from the marginal wall of the tumor, cavity and transitional mucosa, respectively. In biopsy specimens, according to the degree of degeneration of cancer cells, cases with remarkable changes of nucleus, nucleolus, and cytoplasm due to irradiation were classified into the severely degenerated group. According to the histological examinations of resected materials, twenty-four cases were under Grade 1b (Gr I), and nineteen cases were over Grade 2 (Gr II). The rates of cancer cells found in biopsy materials after irradiation were 91.7% in Gr I and were 47.4% in Gr II, respectively (p<0.01). Among the cases, 54.5% in Gr I and 100% in Gr II belonged to the severely degenerated group (p<0.05). Transitional mucosas were not greatly damaged by irradiation. As a result, the greater the irradiation effect was, the fewer cancer cells were found and the more degenerated cancer cells were found in biopsy specimens. But the rate of severely degenerated cells found in the biopsy specimens of little effect cases was high. So it was thought to be too difficult to predict the histological radiation effect of resected specimens from only biopsy specimens. (author).

A new materials irradiation facility at the Kyoto university reactor

International Nuclear Information System (INIS)

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

2003-01-01

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

Development of an End-plug Welding Technology for an Instrumented Fuel Irradiation Test

International Nuclear Information System (INIS)

Kim, Soo Sung; Lee, Chul Yong; Shin, Yoon Taek; Choo, Kee Nam

2010-01-01

The irradiation test of end-plug specimens was planned for the evaluation of nuclear fuels performance. To establish the fabrication process, and for satisfying the requirements of the irradiation test, an orbital-GTA weld machine for the specimens of the dual rods was developed, and the preliminary welding experiments for optimizing the process conditions of the specimens of the dual rods were performed. Dual rods with a 9.5mm diameter and a 0.6mm wall thickness of the cladding tubes and end-plugs have been used and the optimum conditions of the pin-hole welding have also been selected. This paper describes the experimental results of the GTA welds of the specimens of the dual rods and the metallography examinations of the GTA welded specimens for various welding conditions for the instrumented fuel irradiation test. These investigations satisfied the requirements of the instrumented irradiation test and the GTA welds for the specimens of the dual rods at the HANARO research reactor

Characterisation of gamma-irradiated crystalline polymer: Pt. 3

International Nuclear Information System (INIS)

Jinhua Feng; Lihua Zhang; Donglin Chen

1991-01-01

Thermal behaviour of γ-irradiated plain PA1010 containing different amounts of difunctional cross-linking agent BMI was investigated. In DSC endo- and exotherm, it was found that during irradiation the presence of BMI markedly changes the melting and crystallisation characteristics of PA1010. A supposition that the network of BMI-containing specimens is rather loose in structure was proposed to explain the discrepancy in thermal behaviour between these two kinds of specimens. The supposition was further ascertained by the less brittleness in mechanical property of specimens containing BMI. Besides, the complexity of the thermal behaviour of γ-irradiated PA1010 was discussed and attributed mainly to the increase in σ e , the fold surface free energy of chain fold crystals. (author)

Further Charpy impact test results of low activation ferritic alloys, irradiated at 430 degrees C to 67 dpa

International Nuclear Information System (INIS)

Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

1997-01-01

Miniature CVN specimens of four ferritic alloys, GA3X, F82H, GA4X and HT9, have been impact tested following irradiation at 430 degrees C to 67 dpa. Comparison of the results with those of the previously tested lower dose irradiation condition indicates that the GA3X and F82H alloys, two primary candidate low activation alloys, exhibit virtually identical behavior following irradiation at 430 degrees C to ∼67 dpa and at 370 degrees C to ∼15 dpa. Very little shift is observed in either DBTT or USE relative to the unirradiated condition. The shifts in DBTT and USE observed in both GA4X and HT9 were smaller after irradiation at 430 degrees C to ∼67 dpa than after irradiation at 370 degrees C to ∼15 dpa

Further Charpy impact test results of low activation ferritic alloys, irradiated at 430{degrees}C to 67 dpa

Energy Technology Data Exchange (ETDEWEB)

Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

1997-04-01

Miniature CVN specimens of four ferritic alloys, GA3X, F82H, GA4X and HT9, have been impact tested following irradiation at 430{degrees}C to 67 dpa. Comparison of the results with those of the previously tested lower dose irradiation condition indicates that the GA3X and F82H alloys, two primary candidate low activation alloys, exhibit virtually identical behavior following irradiation at 430{degrees}C to {approximately}67 dpa and at 370{degrees}C to {approximately}15 dpa. Very little shift is observed in either DBTT or USE relative to the unirradiated condition. The shifts in DBTT and USE observed in both GA4X and HT9 were smaller after irradiation at 430{degrees}C to {approximately}67 dpa than after irradiation at 370{degrees}C to {approximately}15 dpa.

Irradiation effects in strain aged pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Grounes, M; Myers, H P

1962-02-15

Tensile specimens, Charpy-V notch and subsize impact specimens of an aluminium killed carbon manganese steel, have been irradiated at 160 - 190 deg C in the reactor G1. The total neutron dose received was 2.4 x 10{sup 18} n/cm{sup 2} (> 1 MeV). Specimens were prepared from normalized plate and from strain aged material from the same plate. It was found that the changes in brittle ductile transition temperature due to neutron irradiation and those due to strain ageing must be considered additive.

Irradiation proctitis

International Nuclear Information System (INIS)

Minami, Akira

1977-01-01

Literatures on late rectal injuries are discussed, referring to two patients with uterine cervical cancer in whom irradiation proctitis occurred after telecobalt irradiation following uterine extirpation. To one patients, a total of 5000 rads was irradiated, dividing into 250 rads at one time, and after 3 months, irradiation with a total of 2000 rads, dividing into 200 rads at one time, was further given. In another one patient, two parallel opposing portal irradiation with a total of 6000 rads was given. About a year after the irradiation, rectal injuries and cystitis, accompanying with hemorrhage, were found in both of the patients. Rectal amputation and proctotoreusis were performed. Cystitis was treated by cystic irradiation in the urological department. Pathohistological studies of the rectal specimen revealed atrophic mucosa, and dilatation of the blood vessels and edema in the colonic submucosa. Incidence of this disease, term when the disease occurs, irradiation dose, type of the disease, treatment and prevention are described on the basis of the literatures. (Kanao, N.)

Irradiation proctitis

Energy Technology Data Exchange (ETDEWEB)

Minami, A [Osaka Kita Tsishin Hospital (Japan)

1977-06-01

Literatures on late rectal injuries are discussed, referring to two patients with uterine cervical cancer in whom irradiation proctitis occurred after telecobalt irradiation following uterine extirpation. To one patients, a total of 5000 rads was irradiated, dividing into 250 rads at one time, and after 3 months, irradiation with a total of 2000 rads, dividing into 200 rads at one time, was further given. In another one patient, two parallel opposing portal irradiation with a total of 6000 rads was given. About a year after the irradiation, rectal injuries and cystitis, accompanying with hemorrhage, were found in both of the patients. Rectal amputation and proctotoreusis were performed. Cystitis was treated by cystic irradiation in the urological department. Pathohistological studies of the rectal specimen revealed atrophic mucosa, and dilatation of the blood vessels and edema in the colonic submucosa. Incidence of this disease, term when the disease occurs, irradiation dose, type of the disease, treatment and prevention are described on the basis of the literatures.

Specimen rotation system of the WSU TRIGA-fueled reactor

International Nuclear Information System (INIS)

Lovas, Thomas A.

1976-01-01

The specimen rotation system presently in use at the WSU reactor has been designed to provide maximum utilization of the irradiation capabilities achieved through use of TRIGA-type fuel. This paper describes the system with particular emphasis on characteristics which are advantageous to experimenters. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-09-01

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

Design, Fabrication, and Initial Operation of a Reusable Irradiation Facility

International Nuclear Information System (INIS)

Heatherly, D.W.; Thoms, K.R.; Siman-Tov, I.I.; Hurst, M.T.

1999-01-01

A Heavy-Section Steel Irradiation (HSSI) Program project, funded by the US Nuclear Regulatory Commission, was initiated at Oak Ridge National Laboratory to develop reusable materials irradiation facilities in which metallurgical specimens of reactor pressure vessel steels could be irradiated. As a consequence, two new, identical, reusable materials irradiation facilities have been designed, fabricated, installed, and are now operating at the Ford Nuclear Reactor at the University of Michigan. The facilities are referred to as the HSSI-IAR facilities with the individual facilities being designated as IAR-1 and IAR-2. This new and unique facility design requires no cutting or grinding operations to retrieve irradiated specimens, all capsule hardware is totally reusable, and materials transported from site to site are limited to specimens only. At the time of this letter report, the facilities have operated successfully for approximately 2500 effective full-power hours

Final report on development and operation of instrumented irradiation capsules for creep experiments on nuclear fuels at FR2

International Nuclear Information System (INIS)

Haefner, H.E.; Philipp, K.; Blumhofer, M.

1980-02-01

The capsule test rig No. 154 removed from FR2 in April 1979 was the last irradiation rig in a long series of creep experiments. The target of the irradiation tests, started exactly ten years ago, was to investigate the creep behaviour of various ceramic nuclear fuels under different in-pile irradiation conditions. An irradiation test rig had been developed for this purpose which allowed the continuous measurement of changes in length of fuel specimens. A total of 28 capsule test rigs each containing two packages of creep specimens have been irradiated in FR2 during this decade. They included 23 specimen stacks of UO 2 , 16 specimen stacks of UO 2 -PuO 2 , 4 specimen stacks of UN, 10 specimen stacks of (U,Pu) C, and 13 reference specimens of molybdenum. Besides the description of the test facility, the report provides above all a survey of the operation data applicable to the specimens and of the operating experience gathered as well as of the findings obtained in post-irradiation examinations. (orig.) [de

Fabrication and operation of HFIR-MFE RB* spectrally tailored irradiation capsules

International Nuclear Information System (INIS)

Longest, A.W.; Pawel, J.E.; Heatherly, D.W.; Sitterson, R.G.; Wallace, R.L.

1993-01-01

Fabrication and operation of four HFIR-MFE RB * capsules (60, 200, 330, and 400 degrees C) to accommodate MFE specimens previously irradiated in spectrally tailored experiments in the ORR are proceeding satisfactorily. With the exception of the 60 degrees C capsule, where the test specimens were in direct contact with the reactor cooling water, specimen temperatures (monitored by 21 thermocouples) are controlled by varying the thermal conductance of a thin gap region between the specimen holder outer sleeve and containment tube. Irradiation of the 60 and 330 degrees C capsules, which started on July 17, 1990, was completed on November 14, 1992, after 24 cycles of irradiation to an incremental damage level of approximately 10.9 displacements per atom (dpa). Assembly of the follow-up 200 and 400 degrees C capsules was completed in November 1992, and their planned 20-cycle irradiation to approximately 9.1 incremental dpa was started on November 21, 1992. As of February 11, 1993, the 200 and 400 degrees C capsules had successfully completed three cycles of irradiation to approximately 1.4 incremental dpa

Electron-beam irradiation effects on mechanical properties of PEEK/CF composite

International Nuclear Information System (INIS)

Sasuga, Tsuneo; Seguchi, Tadao

1989-01-01

Carbon fibre-reinforced composite (PEEK/CF) using polyarylether-ether-ketone (PEEK) as a matrix material was prepared and electron-beam irradiation effects on the mechanical properties at low and high temperatures were studied. The flexural strength and modulus of the unirradiated PEEK/CF were almost the same as those of carbon fibre-reinforced composites with epoxide resin. The mechanical properties at room temperature were little affected by irradiation up to 180 MGy, but in the test at 77K the strength of the specimens irradiated over 100 MGy was slightly decreased. The mechanical properties of the unirradiated specimen decreased with increasing testing temperature, but the high-temperature properties were improved by irradiation, i.e. the strength measured at 413K for the specimen irradiated with 120 MGy almost reached the value for the unirradiated specimen measured at room temperature. It was apparent from the viscoelastic measurement that the improvement of mechanical properties at high temperature resulted from the high-temperature shift of the glass transition of the matrix PEEK caused by radiation-induced cross-linking. (author)

Helium release from neutron-irradiated Li{sub 2}O single crystals

Energy Technology Data Exchange (ETDEWEB)

Yamaki, Daiju; Tanifuji, Takaaki; Noda, Kenji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Helium release behavior in post-irradiation heating tests was investigated for Li{sub 2}O single crystals which had been irradiated with thermal neutrons in JRR-4 and JRR-2, and fast neutrons in FFTF. It is clarified that the helium release curves from JRR-4 and JRR-2 specimens consists of only one broad peak. From the dependence of the peak temperatures on the neutron fluence and the crystal diameter, and the comparison with the results obtained for sintered pellets, it is considered that the helium generated in the specimen is released through the process of bulk diffusion with trapping by irradiation defects such as some defect clusters. For the helium release from FFTF specimens, two broad peaks were observed in the release curves. It is considered to suggest that two different diffusion paths exist for helium migration in the specimen, that is, bulk diffusion and diffusion through the micro-crack due to the heavy irradiation. In addition, helium bubble formation after irradiation due to the high temperature over 800K is suggested. (J.P.N.)

Elevated-temperature tensile properties of 2 1/4 Cr-1 Mo steel irradiated in the EBR-II, AD-2 experiment

International Nuclear Information System (INIS)

Klueh, R.L.; Vitek, J.M.

1984-01-01

The effect of irradiated on the tensile properties of 2 1/4 Cr-1 Mo steel was determined for specimens irradiation in EBR-II at 390 to 550 0 C. Unirradiated control specimens and specimens aged for 5000 h at the irradiation temperatures were also tested. Irradiation to approximately 9 dpa at 390 0 C increased the strength and decreased the ductility compared with the unirradiated and aged specimens. Softening occurred in samples irradiated and tested at 450, 500, and 550 0 C

Irradiation induced tensile property change of SA 508 Cl. 3 reactor pressure vessel steels

International Nuclear Information System (INIS)

Chi, Se Hwan; Hong, Jun Hwa; Kuk, Il Hiun

1998-01-01

Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were irradiated to a neutron fluence of 2.7 x 10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Ban-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation (VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by conventional TEM. (author)

Miniature specimen technology for postirradiation fatigue crack growth testing

International Nuclear Information System (INIS)

Mervyn, D.A.; Ermi, A.M.

1979-01-01

Current magnetic fusion reactor design concepts require that the fatigue behavior of candidate first wall materials be characterized. Fatigue crack growth may, in fact, be the design limiting factor in these cyclic reactor concepts given the inevitable presence of crack-like flaws in fabricated sheet structures. Miniature specimen technology has been developed to provide the large data base necessary to characterize irradiation effects on the fatigue crack growth behavior. An electrical potential method of measuring crack growth rates is employed on miniature center-cracked-tension specimens (1.27 cm x 2.54 cm x 0.061 cm). Results of a baseline study on 20% cold-worked 316 stainless steel, which was tested in an in-cell prototypic fatigue machine, are presented. The miniature fatigue machine is designed for low cost, on-line, real time testing of irradiated fusion candidate alloys. It will enable large scale characterization and development of candidate first wall alloys

Study on silk yellowing induced by gamma-irradiation

International Nuclear Information System (INIS)

Tsukada, Masuhiro; Aoki, Akira

1985-01-01

The changes in the yellow color of silk threads with total dose of irradiation applied were described and studied by a colorimetric method and by monochrome photography. The change into a yellow color of the specimen in the course of irradiation was clearly detected in photographs using filters, 2B and SC 56 under light conditions at the wavelength of 366 nm. The b/L value measured by colorimetry in undegummed and degummed silk fibers sharply increased in the early stage of irradiation. Yellow color indices (b/L) of the specimen subjected to gamma-irradiation continued to increase and the yellow color of the silk threads became more pronounced above a total dose of irradiation of 21 Mrad. The b/L value of the undegummed silk fiber which had deen irradiated was about 2 times that of the degummed silk fiber. (author)

Irradiation effects on fracture toughness of two high-copper submerged-arc welds, HSSI series 5

International Nuclear Information System (INIS)

Nanstad, R.K.; Haggag, F.M.; McCabe, D.E.; Iskander, S.K.; Bowman, K.O.; Menke, B.H.

1992-10-01

The Fifth Irradiation Series in the Heavy-Section Steel irradiation (HSSI) Program was aimed at obtaining a statistically significant fracture toughness data base on two weldments with high-copper contents to determine the shift and shape of the K lc curve as a consequence of irradiation. The program included irradiated Charpy V-notch impact, tensile, and drop-weight specimens in addition to compact fracture toughness specimens. Compact specimens with thicknesses of 25.4, 50.8, and 101.6 mm [1T C(T), 2T C(T), and 4T C(T), respectively] were irradiated. Additionally, unirradiated 6T C(T) and 8T C(T) specimens with the same K lc measuring capacity as the irradiated specimens were tested. The materials for this irradiation series were two weldments fabricated from special heats of weld wire with copper added to the melt. One lot of Linde 0124 flux was used for all the welds. Copper levels for the two welds are 0.23 and 0.31 wt %, while the nickel contents for both welds are 0.60 wt %. Twelve capsules of specimens were irradiated in the pool-side facility of the Oak Ridge Research Reactor at a nominal temperature of 288 degree C and an average fluence of about 1.5 x 10 19 neutrons/cm 2 (> 1 MeV). This volume, Appendices E and F, contains the load-displacement curves and photographs of the fracture toughness specimens from the 72W weld (0.23 wt % Cu) and the 73 W weld (0.31 wt % Cu), respectively

Tensile and fracture toughness test results of neutron irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Chaouadi, R.; Moons, F.; Puzzolante, J.L. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium)

1998-01-01

Tensile and fracture toughness test results of four Beryllium grades are reported here. The flow and fracture properties are investigated by using small size tensile and round compact tension specimens. Irradiation was performed at the BR2 material testing reactor which allows various temperature and irradiation conditions. The fast neutron fluence (>1 MeV) ranges between 0.65 and 2.45 10{sup 21} n/cm{sup 2}. In the meantime, un-irradiated specimens were aged at the irradiation temperatures to separate if any the effect of temperature from irradiation damage. Test results are analyzed and discussed, in particular in terms of the effects of material grade, test temperature, thermal ageing and neutron irradiation. (author)

An ESR study of radicals induced in irradiated fresh mango

International Nuclear Information System (INIS)

Kikuchi, Masahiro; Hussain, Mohammed S.; Morishita, Norio; Kobayashi, Yasuhiko; Ukai, Mitsuko; Shimoyama, Yuhei

2009-01-01

An electron spin resonance (ESR) spectroscopic study was performed on the radicals induced irradiated fresh mangoes. Fresh Philippine mangoes were irradiated by the γ-rays, lyophilized and powdered. The ESR spectrum of the dry specimen showed a strong main peak at g=2.004 and a pair of peaks at both magnetic fields of the main peak. The main peak detected from flesh and skin specimens faded away in a few days after the irradiation. On the other hand, the side peaks showed a well-defined dose response even 9 days after the irradiation. The side-peak is a useful mean to define the irradiation on fresh mangoes. (author)

Irradiation creep in zirconium single crystals

International Nuclear Information System (INIS)

MacEwen, S.R.; Fidleris, V.

1976-07-01

Two identical single crystals of crystal bar zirconium have been creep tested in reactor. Both specimens were preirradiated at low stress to a dose of about 4 x 10 23 n/m 2 (E > 1 MeV), and were then loaded to 25 MPa. The first specimen was loaded with reactor at full power, the second during a shutdown. The loading strain for both crystals was more than an order of magnitude smaller than that observed when an identical unirradiated crystal was loaded to the same stress. Both crystals exhibited periods of primary creep, after which their creep rates reached nearly constant values when the reactor was at power. During shutdowns the creep rates decreased rapidly with time. Electron microscopy revealed that the irradiation damage consisted of prismatic dislocation loops, approximately 13.5 nm in diameter. Cleared channels, identified as lying on (1010) planes, were also observed. The results are discussed in terms of the current theories for flux enhanced creep in the light of the microstructures observed. (author)

Plastic zone size for nanoindentation of irradiated Fe–9%Cr ODS

Energy Technology Data Exchange (ETDEWEB)

Dolph, Corey K. [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Silva, Douglas J. da [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Federal University of São Carlos, Rodovia Washington Luís, km 235 - SP-310, São Carlos, São Paulo (Brazil); Swenson, Matthew J. [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Wharry, Janelle P., E-mail: jwharry@purdue.edu [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States)

2016-12-01

The objective of this study is to determine irradiation effects on the nanoindentation plastic zone morphology in a model Fe–9%Cr ODS alloy. Specimens are irradiated to 50 displacements per atom at 400°C with Fe{sup ++} self-ions or to 3 dpa at 500°C with neutrons. The as-received specimen is also studied as a control. The nanoindentation plastic zone size is calculated using two approaches: (1) an analytical model based on the expanding spherical cavity analogy, and (2) finite element modeling (FEM). Plastic zones in all specimen conditions extend radially outward from the indenter, ∼4–5 times the tip radius, indicative of fully plastic contact. Non-negligible plastic flow in the radial direction requires the experimentalist to consider the plastic zone morphology when nanoindenting ion-irradiated specimens; a single nanoindent may sample non-uniform irradiation damage, regardless of whether the indent is made top-down or in cross-section. Finally, true stress-strain curves are generated.

Cracking behavior of thermally aged and irradiated CF-8 cast austenitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Chen, Y., E-mail: Yiren_Chen@anl.gov [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Alexandreanu, B.; Chen, W.-Y.; Natesan, K. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Li, Z.; Yang, Y. [University of Florida, Gainesville, FL 32611 (United States); Rao, A.S. [US Nuclear Regulatory Commission, 11545 Rockville Pike, Rockville, MD 20852 (United States)

2015-11-15

To assess the combined effect of thermal aging and neutron irradiation on the cracking behavior of CF-8 cast austenitic stainless steel, crack growth rate (CGR) and fracture toughness J-R curve tests were carried out on compact-tension specimens in high-purity water with low dissolved oxygen. Both unaged and thermally aged specimens were irradiated at ∼320 °C to 0.08 dpa. Thermal aging at 400 °C for 10,000 h apparently had no effect on the corrosion fatigue and stress corrosion cracking behavior in the test environment. The cracking susceptibility of CF-8 was not elevated significantly by neutron irradiation at 0.08 dpa. Transgranular cleavage-like cracking was the main fracture mode during the CGR tests, and a brittle morphology of delta ferrite was often seen on the fracture surfaces at the end of CGR tests. The fracture toughness J-R curve tests showed that both thermal aging and neutron irradiation can induce significant embrittlement. The loss of fracture toughness due to neutron irradiation was more pronounced in the unaged than aged specimens. After neutron irradiation, the fracture toughness values of the unaged and aged specimens were reduced to a similar level. G-phase precipitates were observed in the aged and irradiated specimens with or without prior aging. The similar microstructural changes resulting from thermal aging and irradiation suggest a common microstructural mechanism of inducing embrittlement in CF-8.

Irradiation embrittlement and optimisation of annealing

International Nuclear Information System (INIS)

1993-01-01

This conference is composed of 30 papers grouped in 6 sessions related to the following themes: neutron irradiation effects in pressure vessel steels and weldments used in PWR, WWER and BWR nuclear plants; results from surveillance programmes (irradiation induced damage and annealing processes); studies on the influence of variations in irradiation conditions and mechanisms, and modelling; mitigation of irradiation effects, especially through thermal annealing; mechanical test procedures and specimen size effects

Irradiation embrittlement and optimisation of annealing

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-31

This conference is composed of 30 papers grouped in 6 sessions related to the following themes: neutron irradiation effects in pressure vessel steels and weldments used in PWR, WWER and BWR nuclear plants; results from surveillance programmes (irradiation induced damage and annealing processes); studies on the influence of variations in irradiation conditions and mechanisms, and modelling; mitigation of irradiation effects, especially through thermal annealing; mechanical test procedures and specimen size effects.

Effect of fast-neutron irradiation on plastic deformation of Type 304 stainless steel

International Nuclear Information System (INIS)

Yamada, H.

1978-01-01

Plastic deformation of EBR-II-irradiated Type 304 stainless steel was investigated by a stress-relaxation method. The stress-strain-rate relationships for the irradiated specimens at room temperature are concave upward, which are similar to those for the unirradiated specimens. However, concave downward behavior in the stress-strain-rate relationships were observed at much lower temperatures for the irradiated specimens in contrast to the unirradiated specimens. These results were analyzed succccessfully using Hart's mechanical equation-of-state concept. It was found that the hardness sigma*, which is the minimum stress necessary for the dislocation to overcome obstacles without thermal activation, increases linearly with fast-neutron fluence. This increase in sigma* is consistent with so-called ''irradiation hardening.'' In addition, resistance to dislocation glide, which is quantitatively measured in terms of sigma 0 , was observed to decrease linearly with fast-neutron fluence. The decrease in sigma 0 can be attributed to a decrease of solute drag due to irradiation-induced solute segregation

Temperature and dose dependencies of microstructure and hardness of neutron irradiated OFHC copper

International Nuclear Information System (INIS)

Singh, B.N.; Horsewell, A.; Toft, P.; Edwards, D.J.

1995-01-01

Tensile specimens of pure oxygen free high conductivity (OFHC) copper were irradiated with fission neutrons between 320 and 723 K to fluences in the range 5x10 21 to 1.5x10 24 n/m 2 (E>1 MeV) with a flux of 2.5x10 17 n/m 2 s. Irradiated specimens were investigated by transmission electron microscopy (TEM) and quantitative determinations were made of defect clusters and cavities. The dose dependence of tensile properties of specimens irradiated at 320 K was determined at 295 K. Hardness measurements were made at 295 K on specimens irradiated at different temperatures and doses. Microstructures of tensile tested specimens were also investigated by TEM. Results show that the increase in cluster density and hardening nearly saturate at a dose of similar 0.3 dpa. Irradiations at 320 K cause a drastic decrease in the uniform elongation already at ∼ =0.1 dpa. It is suggested that the irradiation-induced increase in the initial yield stress and a drastic decrease in the ability of copper to deform plastically in a homogeneous fashion are caused by a substantial reduction in the ability of grown-in dislocations to act as efficient dislocation sources. ((orig.))

Tensile properties of helium-injected V-15Cr-5Ti after irradiation in EBR-II

International Nuclear Information System (INIS)

Grossbeck, M.L.; Horak, J.A.

1985-01-01

Miniature specimens of V-15Cr-5Ti were prepared in the annealed condition and with 10, 20, and 30% cold work. The annealed specimens were cyclotron injected with helium and irradiated in sodium in EBR-II. The cold-worked specimens were irradiated in EBR-II but not helium injected. The specimens were irradiated at 400, 525, 625, and 700 0 C and received a fluence of 4.1 to 5.5 x 10 26 neutrons/m 2 (E > 0.1 meV). Tensile testing revealed very significant embrittlement as a result of the neutron irradiation but a much smaller change, mostly at 400 0 C, resulting from helium injection. 5 references, 9 figures, 2 tables

Tensile properties of irradiated TZM and tungsten

International Nuclear Information System (INIS)

Steichen, J.M.

1975-04-01

The effect of neutron irradiation on the elevated temperature tensile properties of TZM and tungsten has been experimentally determined. Specimens were irradiated at a temperature of approximately 720 0 F to fluences of 0.4 and 0.9 x 10 22 n/cm 2 (E greater than 0.1 MeV). Test parameters for both control and irradiated specimens included strain rates from 3 x 10 -4 to 1 s -1 and temperatures from 72 to 1700 0 F. The results of these tests were correlated with a rate-temperature parameter (T ln A/epsilon) to provide a concise description of material behavior over the range of deformation conditions of this study. The yield strength of the subject materials was significantly increased by decreasing temperature, increasing strain rate, and increasing fluence. Ductility was significantly reduced at any temperature or strain rate by increasing fluence. Cleavage fractures occurred in both unirradiated and irradiated specimens when the yield strength was elevated to the effective cleavage stress by temperature and/or strain rate. Neutron irradiation for the conditions of this study increased the ductile-to-brittle transition temperature of tungsten by approximately 300 0 F and TZM by approximately 420 0 F. (U.S.)

Effect of irradiation temperature and strain rate on the mechanical properties of V-4Cr-4Ti irradiated to low doses in fission reactors

International Nuclear Information System (INIS)

Zinkle, S.J.; Snead, L.L.; Rowcliffe, A.F.; Alexander, D.J.; Gibson, L.T.

1998-01-01

Tensile tests performed on irradiated V-(3-6%)Cr-(3-6%)Ti alloys indicate that pronounced hardening and loss of strain hardening capacity occurs for doses of 0.1--20 dpa at irradiation temperatures below ∼330 C. The amount of radiation hardening decreases rapidly for irradiation temperatures above 400 C, with a concomitant increase in strain hardening capacity. Low-dose (0.1--0.5 dpa) irradiation shifts the dynamic strain aging regime to higher temperatures and lower strain rates compared to unirradiated specimens. Very low fracture toughness values were observed in miniature disk compact specimens irradiated at 200--320 C to ∼1.5--15 dpa and tested at 200 C

Irradiation creep of dispersion strengthened copper alloy

Energy Technology Data Exchange (ETDEWEB)

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A. [and others

1997-04-01

Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.

Irradiation creep of dispersion strengthened copper alloy

International Nuclear Information System (INIS)

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

1997-01-01

Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al 2 O 3 , is very similar to the GlidCop trademark alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10 21 n/cm 2 (E>0.1 MeV), which corresponds to ∼3-5 dpa. The irradiation temperature ranged from 60-90 degrees C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of ±0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as ∼2 x 10 -9 s -1 . These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys

Comparison of irradiated and hydrogen implanted German RPV steels using PAS technique

Energy Technology Data Exchange (ETDEWEB)

Pecko, Stanislav, E-mail: stanislav.pecko@stuba.sk; Sojak, Stanislav; Slugeň, Vladimír

2015-12-15

Highlights: • German RPV steels were originally studied by positron annihilation spectroscopy. • Neutron irradiated and hydrogen ion implanted specimens were studied. • Both irradiation ways caused to increase of defect size. • We determined that the defect size was higher in implanted specimens. - Abstract: Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This spectroscopic method is a really effective tool for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to irradiation. German commercial reactor pressure vessel steels, originally from CARISMA program, were used in our study. The German experimental reactor VAK was selected as the proper irradiation facility in the 1980s. A specimen in as-received state and 2 different irradiated cuts from the same material were measured by PALS and size of defects with their intensity was indentified. Afterwards there was prepared an experiment with concern in simulation of neutron irradiation by hydrogen ion implantation on a linear accelerator with energy of 100 keV. Results are concerning on comparison between defects caused by neutron irradiation and hydrogen implantation. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to hydrogen ions implantation.

Fabrication of irradiation capsule for IASCC irradiation tests (2). Irradiation capsule for crack propagation test (Joint research)

International Nuclear Information System (INIS)

Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; Saito, Takashi; Ishitsuka, Etsuo; Kawamura, Hiroshi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi

2008-03-01

It is known that irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, it is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack propagation test is reported. (author)

Fabrication of irradiation capsule for IASCC irradiation tests (1). Irradiation capsule for crack growth test (Joint research)

International Nuclear Information System (INIS)

Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; Saito, Takashi; Ishitsuka, Etsuo; Kawamura, Hiroshi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi

2008-03-01

It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, it is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack growth test is reported. (author)

Rows of Dislocation Loops in Aluminium Irradiated by Aluminium Ions

DEFF Research Database (Denmark)

Henriksen, L.; Johansen, A.; Koch, J.

1967-01-01

Single-crystal aluminium specimens, irradiated with 50-keV aluminium ions, contain dislocation loops that are arranged in regular rows along <110 > directions. ©1967 The American Institute of Physics......Single-crystal aluminium specimens, irradiated with 50-keV aluminium ions, contain dislocation loops that are arranged in regular rows along directions. ©1967 The American Institute of Physics...

Charpy impact test results of four low activation ferritic alloys irradiated at 370{degrees}C to 15 DPA

Energy Technology Data Exchange (ETDEWEB)

Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

1996-10-01

Miniature CVN specimens of four low activation ferritic alloys have been impact tested following irradiation at 370{degrees}C to 15 dpa. Comparison of the results with those of control specimens indicates that degradation in the impact behavior occurs in each of these four alloys. The 9Cr-2W alloy referred to as GA3X and the similar alloy F82H with 7.8Cr-2W appear most promising for further consideration as candidate structural materials in fusion energy system applications. These two alloys exhibit a small DBTT shift to higher temperatures but show increased absorbed energy on the upper shelf.

Mechanical properties of Mo and TZM alloy neutron-irradiated at high temperatures

International Nuclear Information System (INIS)

Ueda, Kazukiyo; Satou, Manabu; Hasegawa, Akira; Abe, Katsunori

1997-01-01

This work reports the mechanical properties of irradiated molybdenum (Mo) and its alloy, TZM. Recrystallized and stress-relieved specimens were irradiated at five temperatures between 373 and 800degC in FFTF/MOTA to fluence levels of 6.8 to 34 dpa. Irradiation embrittlement and hardening were evaluated by three-point bend test and Vickers hardness test, respectively. Stress-relieved materials showed the enough ductility even after high fluence irradiation. The role of layered structure of stress-relieved specimen was discussed. (author)

Tensile properties of a titanium modified austenitic stainless steel and the weld joints after neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Shiba, K.; Ioka, I.; Jitsukawa, S.; Hamada, A.; Hishinuma, A. [and others

1996-10-01

Tensile specimens of a titanium modified austenitic stainless steel and its weldments fabricated with Tungsten Inert Gas (TIG) and Electron Beam (EB) welding techniques were irradiated to a peak dose of 19 dpa and a peak helium level of 250 appm in the temperature range between 200 and 400{degrees}C in spectrally tailored capsules in the Oak Ridge Research Reactor (ORR) and the High Flux Isotope Reactor (HFIR). The He/dpa ratio of about 13 appm/dpa is similar to the typical helium/dpa ratio of a fusion reactor environment. The tensile tests were carried out at the irradiation temperature in vacuum. The irradiation caused an increase in yield stress to levels between 670 and 800 MPa depending on the irradiation temperature. Total elongation was reduced to less than 10%, however the specimens failed in a ductile manner. The results were compared with those of the specimens irradiated using irradiation capsules producing larger amount of He. Although the He/dpa ratio affected the microstructural change, the impact on the post irradiation tensile behavior was rather small for not only base metal specimens but also for the weld joint and the weld metal specimens.

Response of reptilian live to external gamma irradiation

International Nuclear Information System (INIS)

Gupta, M.L.

1990-01-01

Adult healthy specimens of Uromastix hardwickii were exposed to three doses (i.e. 2.25, 4.50 and 9.00 Gy) of gamma radiation from a 60 Co source (experimental group). Five animals were sacrificed at each post-irradiation intervals of 1, 2, 3, 7 and 14 days. The liver was fixed in Bouin's fluid and after processing in a routine way, it was examined histologically. Five sham-irradiated animals (control group) were also sacrificed to compare the results. Low dose (i.e. 2.25 Gy) did not produce any apparent radiolesions in the liver. Changes in the form of cytoplasmic degranulation, swollen hepatocytes, pycnosis, increases in bile pigmentation were noticed after 4.50 and 9.00 Gy gamma ray exposure. Hyperaemia, widening of sinusoids and cytoplasmic vacuolation were also noticed in 9.00 Gy group. The liver exhibited normal picture on day 14 after exposure to both the doses. The radiolesions were found dose dependent. (author)

Dose Distribution of Gamma Irradiators

International Nuclear Information System (INIS)

Park, Seung Woo; Shin, Sang Hun; Son, Ki Hong; Lee, Chang Yeol; Kim, Kum Bae; Jung, Hai Jo; Ji, Young Hoon

2010-01-01

Gamma irradiator using Cs-137 have been widely utilized to the irradiation of cell, blood, and animal, and the dose measurement and education. The Gamma cell 3000 Elan (Nordion International, Kanata, Ontario, Canada) irradiator was installed in 2003 with Cs-137 and dose rate of 3.2 Gy/min. And the BioBeam 8000 (Gamma-Service Medical GmbH, Leipzig, Germany) irradiator was installed in 2008 with Cs-137 and dose rate of 3.5 Gy/min. Our purpose was to evaluate the practical dosimetric problems associated with inhomogeneous dose distribution within the irradiated volume in open air state using glass dosimeter and Gafchromic EBT film dosimeter for routine Gamma irradiator dosimetry applications at the KIRAMS and the measurements were compared with each other. In addition, an user guideline for useful utilization of the device based on practical dosimetry will be prepared. The measurement results of uniformity of delivered dose within the device showed variation more than 14% between middle point and the lowest position at central axis. Therefore, to maintain dose variation within 10%, the criteria of useful dose distribution, for research radiation effects, the irradiated specimen located at central axis of the container should be placed within 30 mm from top and bottom surface, respectively. In addition, for measurements using the film, the variations of dose distribution were more then 50% for the case of less than 10 second irradiation, mostly within 20% for the case of more than 20 second irradiation, respectively. Therefore, the irradiation experiments using the BioBeam 8000 irradiator are recommended to be used for specimen required at least more than 20 second irradiation time.

Fundamental irradiation studies on vanadium alloys

International Nuclear Information System (INIS)

Loomis, B.A.; Garner, F.A.; Ermi, A.M.

1985-05-01

A joint experiment on the irradiation response of simple vanadium alloys has been initiated under the auspices of the DAFS and BES progams. Specimen fabrication is nearly complete and the alloys are expected to be irradiated in lithium in FFTF-MOTA Cycles 7 and 8

Fracture toughness of ferritic alloys irradiated at FFTF

International Nuclear Information System (INIS)

Huang, F.H.

1986-05-01

Ferritic compact tension specimens loaded in the Material Open Test Assembly (MOTA) for irradiation during FFTF Cycle 4 were tested at temperatures ranging from room temperature to 428/degree/C. The electrical potential single specimen method was used to measure the fracture toughness of the specimens. Results showed that the fracture toughness of both HT-9 and 9Cr-1Mo decreases with increasing test temperature and that the toughness of HT-9 was about 30% higher than that of 9Cr-1Mo. In addition, increasing irradiation temperature resulted in an increase in tearing modulus for both alloys. 4 refs., 5 figs., 1 tab

Comparison of damage microstructures in neutron-irradiated vanadium and iron

International Nuclear Information System (INIS)

Horton, L.L.; Farrell, K.

1983-01-01

The cavity morphology and dislocation loop geometry in bcc vanadium are compared with the previously reported observations for neutron-irradiated iron. The specimens were vanadium (V) with 100 wppM of interstitial impurities and vanadium with boron carbide additions (V-B 4 C) which were irradiated to approx. 1 dpa in the same Oak Ridge Research Reactor capsules as the iron specimens

Saturation behavior of irradiation hardening in F82H irradiated in the HFIR

Energy Technology Data Exchange (ETDEWEB)

Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Shiba, K.; Tanigawa, H.; Ando, M. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Klueh, R.L. [Oak Ridge National Laboratory, TN (United States); Stoller, R. [ORNL - Oak Ridge National Laboratory, Materials Science and Technology Div., Oak Ridge, AK TN (United States)

2007-07-01

Full text of publication follows: Post irradiation tensile tests on reduced activation ferritic/martensitic steel, F82H have been conducted over the past two decades using Japan Materials Testing Reactor (JMTR) of JAEA, and Fast Flux Testing Facility (FFTF) of PNNL and High Flux Isotope Reactor (HFIR) of ORNL, USA, under Japan/US collaboration programs. According to these results, F82H does not demonstrate irradiation hardening above 673 K up to 60 dpa. The current study has been concentrated on hardening behavior at temperature around 573 K. A series of low temperature irradiation experiment has been conducted at the HFIR under the international collaborative research between JAEA/US-DOE. In this collaboration, the irradiation condition is precisely controlled by the well matured capsule designing and instrumentation. This paper summarizes recent results of the irradiation experiments focused on F82H and its modified steels compared with the irradiation properties database on F82H. Post irradiation tensile tests have been conducted on the F82H and its modified steels irradiated at 573 K and the dose level was up to 25 dpa. According to these results, irradiation hardening of F82H is saturated by 9 dpa and the as-irradiated 0.2 % proof stress is less than 1 GPa at ambient temperature. The deterioration of total elongation was also saturated by 9 dpa irradiation. The ductility of some modified steels which showed larger total elongation than that of F82H before irradiation become the same level as that of standard F82H steel after irradiation, even though its magnitude of irradiation hardening is smaller than that of F82H. This suggests that the more ductile steel demonstrates the more ductility loss at this temperature, regardless to the hardening level. The difference in ductility loss behavior between various tensile specimens will be discussed as the ductility could depend on the specimen dimension. (authors)

Effect of bonding and bakeout thermal cycles on the properties of copper alloys irradiated at 350 deg. C

International Nuclear Information System (INIS)

Singh, B.N.; Eldrup, M.; Toft, P.; Edwards, D.J.

1997-02-01

Screening experiments were carried out to investigate the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties and electrical resistivity of the oxide dispersion strengthened (GlidCop, CuAl-25) and the precipitation hardened (CuCrZr, CuNiBe) cooper alloys. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing, and bonding thermal treatment followed by re-ageing and the reactor bakeout treatment at 350 deg. C for 100 h. Tensile specimens of CuAl-25 were given the heat treatment corresponding to the bonding thermal cycle. A number of heat treated specimens were neutron irradiated at 350 deg. C to a dose level of ∼ 0.3 dpa in the DR-3 reactor at Risoe. Both unirradiated and irradiated specimens with various heat treatments were tensile tested at 350 deg. C. The microstructure and electrical resistivity of these specimens were determined in the unirradiated as well as irradiated conditions. The post-deformation microstructure of the irradiated specimens was also investigated. The fracture surfaces of both unirradiated and irradiated specimens were examined. Results of these investigations are reported in the present report. The results are briefly discussed in terms of thermal and irradiation stability of precipitates and particles and irradiation-induced segregation, precipitation and recovery of dislocation microstructure. (au) 6 tabs., 24 ills., 9 refs

Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

Science.gov (United States)

Shibata, A.; Nakano, J.; Ohmi, M.; Kawamata, K.; Nakagawa, T.; Tsukada, T.

2012-03-01

To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0-3.9 × 1026 n/m2 (E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03-1.0 × 1026 n/m2. Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

International Nuclear Information System (INIS)

Shibata, A.; Nakano, J.; Ohmi, M.; Kawamata, K.; Nakagawa, T.; Tsukada, T.

2012-01-01

To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0–3.9 × 10 26 n/m 2 (E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03–1.0 × 10 26 n/m 2 . Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-02-16

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

Initial decay process of radicals induced in irradiated food

International Nuclear Information System (INIS)

Kaimori, Yoshihiko; Sakamoto, Yuki; Nakamura, Hideo; Ukai, Mitsuko; Kikuchi, Masahiro; Shimoyama, Yuhei; Kobayashi, Yasuhiko; Kameya, Hiromi

2011-01-01

In order to determine radial decay behaviors of γ-irradiated food, we analyzed radicals in the food using ESR. We detected the ESR signal of specimens just several minutes after irradiation. The singlet signal intensity at g=2.0, originated from organic free radicals was increased as followed by the increasing radiation dose. Singlet signal intensity that increased by γ-irradiation was decreased with time. The phenomena of decay of the ESR singlet signal showed two phase that are rapid decay and slow decay. It was suggested that those two phase decay is due to at least the two radical species. Also we concluded that after three hours of radiation treatment long life radical as ESR signal intensity was detected in irradiated specimen; black pepper, green coffee bean and ginseng, showed the same decay phenomena. But the signal intensity of irradiated black pepper was three times larger than that of irradiated green coffee bean and irradiated ginseng. (author)

Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

International Nuclear Information System (INIS)

Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio

2002-10-01

An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

Energy Technology Data Exchange (ETDEWEB)

Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

2002-10-01

An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

Property change of advanced tungsten alloys due to neutron irradiation

International Nuclear Information System (INIS)

Fukuda, Makoto; Hasegawa, Akira; Tanno, Takashi; Nogami, Shuhei; Kurishita, Hiroaki

2013-01-01

This study investigates the effect of neutron irradiation on the functional properties of pure tungsten (W) and advanced tungsten alloys (e.g., lanthanum (La)-doped W, potassium (K)-doped W, and ultra-fine-grained (UFG) W–TiC alloys) tested in the Japan Materials Testing Reactor (JMTR) or experimental fast reactor Joyo. The irradiation temperature and damage were in the range 804–1073 K and 0.15–0.47 dpa, respectively. TEM images of all samples after 0.42 dpa irradiation at 1023 K showed voids, black dots, and dislocation loops, indicating that similar damage structures were formed in pure W, La-doped W, K-doped W, and UFG W–0.5 wt% TiC. The electrical resistivity of all specimens increased following neutron irradiation. Nearly identical electrical resistivity and irradiation hardening were observed in pure W, La-doped W, and K-doped W. The electrical resistivity of UFG W–TiC was higher than that of other specimens before and after irradiation, which may be attributed to its ultra-fine-grain structure, as well as the presence of impurities introduced during the alloying process. Compared to the other specimens, the UFG W–TiC was more resistant to irradiation hardening

Final report on graphite irradiation test OG-3

International Nuclear Information System (INIS)

Price, R.J.; Beavan, L.A.

1977-01-01

The results of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on graphite specimens irradiated in capsule OG-3 are presented. The graphite grades investigated included near-isotropic H-451 (three different preproduction lots), TS-1240, and SO818; needle coke H-327; and European coal tar pitch coke grades P 3 JHA 2 N, P 3 JHAN, and ASI2-500. Data were obtained in the temperature range 823 0 K to 1673 0 K. The peak fast neutron fluence in the experiment was 3 x 10 25 n/m 3 (E greater than 29 fJ)/sub HTGR/; the total accumulated fluence exceeded 9 x 10 25 n/m 2 on some H-451 specimens and 6 x 10 25 n/m 2 on some TS-1240 specimens. Irradiation-induced dimensional changes on H-451 graphite differed slightly from earlier predictions. For an irradiation temperature of about 1225 0 K, axial shrinkage rates at high fluences were somewhat higher than predicted, and the fluence at which radial expansion started (about 9 x 10 25 n/m 2 at 1275 0 K) was lower. TS-1240 graphite underwent smaller dimensional changes than H-451 graphite, while limited data on SO818 and ASI2-500 graphites showed similar behavior to H-451. P 3 JHAN and P 3 JHA 2 N graphites displayed anisotropic behavior with rapid axial shrinkage. Comparison of dimensional changes between specimens from three logs of H-451 and of TS-1240 graphites showed no significant log-to-log variations for H-451, and small but significant log-to-log variations for TS-1240. The thermal expansivity of the near-isotropic graphites irradiated at 865-1045 0 K first increased by 5 percent to 10 percent and then decreased. At higher irradiation temperatures the thermal expansivity decreased by up to 50 percent. Changes in thermal conductivity were consistent with previously established curves. Specimens which were successively irradiated at two different temperatures took on the saturation conductivity for the new temperature

Effects of postpolymerization microwave irradiation on provisional dental acrylics: physical and mechanical properties.

Science.gov (United States)

Ozkomur, Ahmet; Fortes, Carmen Beatriz Borges

2016-07-26

This study aimed to evaluate the effects of microwave irradiation on the physical and mechanical properties of poly(methyl methacrylate) (PMMA) provisional resins. Twenty bars and 20 disc-shaped specimens were fabricated for each selected provisional restorative material (Dencor and Duralay). Test groups were subjected to microwave irradiation (3 minutes at 600 W) after polymerization. Bar specimens were subjected to a flexural strength test. Disc-shaped specimens were used to evaluate microhardness. Backscattered Raman spectroscopy was employed for each group to define the degree of conversion of the monomer/polymer. The frequency bands corresponding to C = C and C = O groups were used to determine the conversion of methyl methacrylate (MMA) monomers into polymers. Glass transition temperature was determined using a differential scanning calorimeter. Microwave irradiation of both tested autopolymerizing PMMA provisional materials resulted in a statistically significant increase in microhardness, degree of conversion and glass transition temperature values. Also, the results demonstrated a significant increase in flexural strength after postpolymerization microwave irradiation for the Dencor specimens. It is concluded that mechanical and physical properties are positively influenced by microwave irradiation.

Synchrotron radiation microprobe quantitative analysis method for biomedical specimens

International Nuclear Information System (INIS)

Xu Qing; Shao Hanru

1994-01-01

Relative changes of trace elemental content in biomedical specimens are obtained easily by means of synchrotron radiation X-ray fluorescence microprobe analysis (SXRFM). However, the accurate assignment of concentration on a g/g basis is difficult. Because it is necessary to know both the trace elemental content and the specimen mass in the irradiated volume simultaneously. the specimen mass is a function of the spatial position and can not be weighed. It is possible to measure the specimen mass indirectly by measuring the intensity of Compton scattered peak for normal XRF analysis using a X-ray tube with Mo anode, if the matrix was consisted of light elements and the specimen was a thin sample. The Compton peak is not presented in fluorescence spectrum for white light SXRFM analysis. The continuous background in the spectrum was resulted from the Compton scattering with a linear polarization X-ray source. Biomedical specimens for SXRFM analysis, for example biological section and human hair, are always a thin sample for high energy X-ray, and they consist of H,C,N and O etc. light elements, which implies a linear relationship between the specimen mass and the Compton scattering background in the high energy region of spectrum. By this way , it is possible to carry out measurement of concentration for SXRFM analysis

Influence of irradiation spectrum and implanted ions on the amorphization of ceramics

International Nuclear Information System (INIS)

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

1995-01-01

Polycrystalline Al2O3, magnesium aluminate spinel (MgAl2O4), MgO, Si3N4, and SiC were irradiated with various ions at 200-450 K, and microstructures were examined following irradiation using cross-section TEM. Amorphization was not observed in any of the irradiated oxide ceramics, despsite damage energy densities up to ∼7 keV/atom (70 displacements per atom). On the other hand, SiC readily amorphized after damage levels of ∼0.4 dpa at room temperature (RT). Si3N4 exhibited intermediate behavior; irradiation with Fe 2+ ions at RT produced amorphization in the implanted ion region after damage levels of ∼1 dpa. However, irradiated regions outside the implanted ion region did not amorphize even after damage levels > 5 dpa. The amorphous layer in the Fe-implanted region of Si3N4 did not appear if the specimen was simultaneoulsy irradiated with 1-MeV He + ions at RT. By comparison with published results, it is concluded that the implantation of certain chemical species has a pronounced effect on the amorphization threshold dose of all five materials. Intense ionizing radiation inhibits amorphization in Si3N4, but does not appear to significantly influence the amorphization of SiC

Neutron irradiation damage in Al2O3 and Y2O3

International Nuclear Information System (INIS)

Clinard, F.W. Jr.; Bunch, J.M.; Ranken, W.A.

1975-01-01

Two ceramics under consideration for use in fusion reactors, Al 2 O 3 and Y 2 O 3 , were irradiated in the EBR-II fission reactor at 650, 875, and 1025 0 K to fluences between 2 and 6 x 10 21 n/cm 2 (E greater than 0.1 MeV). Samples evaluated include sapphire, Lucalox, alumina, Y 2 O 3 , and Y 2 O 3 -10 percent ZrO 2 (Yttralox). All Al 2 O 3 specimens swelled significantly (1 to 3 percent), with most of the growth observed in sapphire along the c-axis at the higher temperatures. Al 2 O 3 samples irradiated at 875 to 1025 0 K contained a high density of small aligned ''pores''. Irradiated Y 2 O 3 -based ceramics exhibited dimensional stability and a defect content consisting primarily of unresolved damage and/or dislocation loops. The behavior of these ceramics under irradiation is discussed, and the relevance of fission neutron damage studies to fusion reactor applications is considered. (auth)

On impact testing of subsize Charpy V-notch type specimens

International Nuclear Information System (INIS)

Mikhail, A.S.; Nanstad, R.K.

1994-01-01

The potential for using subsize specimens to determine the actual properties of reactor pressure vessel steels is receiving increasing attention for improved vessel condition monitoring that could be beneficial for light-water reactor plant-life extension. This potential is made conditional upon, on the one hand, by the possibility of cutting samples of small volume from the internal surface of the pressure vessel for determination of actual properties of the operating pressure vessel. The plant-life extension will require supplemental surveillance data that cannot be provided by the existing surveillance programs. Testing of subsize specimens manufactured from broken halves of previously tested surveillance Charpy V-notch (CVN) specimens offers an attractive means of extending existing surveillance programs. Using subsize CVN type specimens requires the establishment of a specimen geometry that is adequate to obtain a ductile-to-brittle transition curve similar to that obtained from full-size specimens. This requires the development of a correlation of transition temperature and upper-shelf toughness between subsize and full-size specimens. The present study was conducted under the Heavy-Section Steel Irradiation Program. Different published approaches to the use of subsize specimens were analyzed and five different geometries of subsize specimens were selected for testing and evaluation. The specimens were made from several types of pressure vessel steels with a wide range of yield strengths, transition temperatures, and upper-shelf energies (USEs). Effects of specimen dimensions, including depth, angle, and radius of notch have been studied. The correlation of transition temperature determined from different types of subsize specimens and the full-size specimen is presented. A new procedure for transforming data from subsize specimens was developed and is presented

The results of the surveillance specimen program performed in the RPVs NPP V-2 in Jaslovske Bohunice

Energy Technology Data Exchange (ETDEWEB)

Kupca, L; Beno, P [Vyskumny Ustav Jadrovych Elektrarni, Trnava (Slovakia); Cepeek, S [Atomova Elektraren Bohunice, Jaslovske Bohunice (Slovakia); Tomasich, M [Slovak Nuclear Society, Bratislava (Slovakia)

1994-12-31

After a description of the mechanical and chemical characteristics of the materials (steels, welded joints) used in the pressure vessels of the WWER-440 V-213 type, the present status of the Bohunice NPP Unit 3 and 4 pressure vessel embrittlement assessment programme is presented: neutron flux monitoring and calculations, detector accuracy, irradiation temperature monitoring, reactor core fuel loading calculation, materials, number and types of surveillance specimens, specimen testing. Results are given for 5 years of irradiation: mechanical properties, transition temperatures, lifetime evaluation. 4 refs., 13 figs., 6 tabs.

Hardness distribution and effect of irradiation in FSW-ODS ferritic steels

International Nuclear Information System (INIS)

Noh, Sanghoon; Kasada, Ryuta; Kimura, Akihiko; Nagasaka, Takuya; Sokolov, M.A.; Yamamoto, T.

2014-01-01

Oxide dispersion strengthened ferritic steels (ODS-FS) have been considered as one of the most promising structural materials for advanced nuclear systems such as fusion reactors and next generation fission reactors, because of its excellent elevated temperature strength, corrosion and radiation resistance. Especially, irradiation resistance is a critical issue for the high performance of ODS-FS. In this study, effects of the irradiation on hardness properties of friction stri processed (FSP) ODS-FS were investigated. FSP technique was employed on ODS-FS. A plate specimen was cut out from the cross section and irradiated to 1.2 dpa at 573K in the High Flux Isotope Reactor (HFIR). To investigate the effect of neutron irradiation on processed area, the hardness distributions were evaluated on the cross section. Hardness of FSP ODS-FS was various with each microstructure after irradiation to 1.2 dpa at 573K. The increase of Vickers hardness was significant in the stirred zone and heat affected zone. Base material exhibited the lowest hardening about 38HV. Since nano-oxide particles in stirred zone showed identical mean diameter and number density, it is considered that hardening differences between stirred zone and base material is due to differences in initial dislocation density. (author)

EFFECT OF GAMMA RAY IRRADIATION ON INTERLAMINAR SHEAR STRENGTH OF GLASS FIBER REINFORCED PLASTICS AT 77 K

International Nuclear Information System (INIS)

Nishimura, A.; Nishijima, S.; Izumi, Y.

2008-01-01

It is known that an organic material is damaged by gamma ray irradiation, and the strength after irradiation has dependence on the gamma ray dose. These issues are important not only to make global understanding of electric insulating performance of glass fiber reinforced plastics (GFRP) under irradiation condition but also to develop new insulation materials. This paper presents the dependence of fracture mode and interlaminar shear strength (ILSS) on the material and the gamma ray irradiation effect on the fracture mode and the ILSS. 6 mm radius loading nose and supports were used to prompt ILS fracture for a short beam test. A 2.5 mm thick small specimen machined out of a 13 mm thick G-10CR GFRP plate (sliced specimen) showed lower ILSS and translaminar shear (TLS) fracture, although the same size specimen prepared from a 2.5 mm G-10CR GFRP plate (non-sliced specimen) showed ILS fracture and the higher ILSS. Both type of specimens showed the degradation of ILSS after gamma ray irradiation. The fracture mode of the non-sliced specimen changed from ILS to TLS fracture and no bending fracture was observed. The resistance to shear deformation of glass cloth/epoxy laminate structure would be damaged by the irradiation

Borosilicate glass for gamma irradiation fields

Science.gov (United States)

Baydogan, N.; Tugrul, A. B.

2012-11-01

Four different types of silicate glass specimens were irradiated with gamma radiation using a Co-60 radioisotope. Glass specimens, with four different chemical compositions, were exposed to neutron and mixed neutron/gamma doses in the central thimble and tangential beam tube of the nuclear research reactor. Optical variations were determined in accordance with standardisation concept. Changes in the direct solar absorbance (αe) of borosilicate glass were examined using the increase in gamma absorbed dose, and results were compared with the changes in the direct solar absorbance of the three different type silicate glass specimens. Solar absorption decreased due to decrease of penetration with absorbed dose. αe of borosilicate increased considerably when compared with other glass types. Changes in optical density were evaluated as an approach to create dose estimation. Mixed/thermal neutron irradiation on glass caused to increse αe.

Argon laser irradiation of the otolithic organ

Energy Technology Data Exchange (ETDEWEB)

Okuno, T.; Nomura, Y.; Young, Y.H.; Hara, M. (Univ. of Tokyo (Japan))

1990-12-01

An argon laser was used to irradiate the otolithic organs of guinea pigs and cynomolgus monkeys. After stapedectomy, the argon laser (1.5 W x 0.5 sec/shot) irradiated the utricle or saccule without touching the sensory organs. The stapes was replaced over the oval window after irradiation. The animals used for acute observation were killed immediately for morphologic studies; those used for long-term observation were kept alive for 2, 4, or 10 weeks. Acute observation revealed that sensory and supporting cells were elevated from the basement membrane only in the irradiated area. No rupture of the membranous labyrinth was observed. Long-term observation revealed that the otolith of the macula utriculi had disappeared in 2-week specimens. The entire macula utricili had disappeared in 10-week specimens. No morphologic changes were observed in cochlea, semicircular canals, or membranous labyrinth. The saccule showed similar changes.

Application of photostimulated luminescence (PSL) to detect irradiated molluscs

International Nuclear Information System (INIS)

Marchesani, G.; Chiaravalle, A.E.; Chiesa, L.M.

2011-01-01

Complete text of publication follows. In contrast to thermally processed foods, irradiation is a cold treatment both to reduce microbiological contamination and to increase the shelf-life of raw seafood. According to the list of States' authorizations molluscs can be irradiated in a range of 0.5 / 3 kGy only in authorized countries (e.g. UK, Belgium and Czech Republic). Therefore the aim of this study is to identify, at different dose levels (0.5, 1, 1.5, 2, 3 kGy), irradiated oysters, clams and mussels using luminescence materials from different sites (shells and pulps) and to determine sample sensitivity for previous screening result confirmation. A total number of 10 samples for each species were analyzed by both procedures: screening and calibrated PSL. Samples were irradiated using a low energy X-ray irradiator (RS-2400, Radsource Inc.) with the following operational settings: 150 kV and 45 mA. Whole pulps were simply dispensed into a clean Petri-dish whereas shells powder required to be fixed as a thick layer with silicone grease. Results obtained showed that screening analysis can be used to identify correctly all irradiated and non irradiated samples. Particularly untreated sample exhibited a sensitivity index from 2 to 4 order of magnitude greater than the exposed sample one, while for exposed specimen calibrated PSL signals, after re-irradiation at defined dose, were of the same order of the first measurement (initial PSL counts). In conclusion mineral debris contaminating pulps and biocarbonates from shells can be considered reliable radioinduced markers and PSL techniques can be easily applied for rapid and simple analysis to identify irradiated molluscs in official controls.

Determination of irradiation temperature using SiC temperature monitors

International Nuclear Information System (INIS)

Maruyama, Tadashi; Onose, Shoji

1999-01-01

This paper describes a method for detecting the change in length of SiC temperature monitors and a discussion is made on the relationship between irradiation temperature and the recovery in length of SiC temperature monitors. The SiC specimens were irradiated in the experimental fast reactor JOYO' at the irradiation temperatures around 417 to 645degC (design temperature). The change in length of irradiated specimens was detected using a dilatometer with SiO 2 glass push rod in an infrared image furnace. The temperature at which recovery in macroscopic length begins was obtained from the annealing intersection temperature. The results of measurements indicated that a difference between annealing intersection temperature and the design temperature sometimes reached well over ±100degC. A calibration method to obtain accurate irradiation temperature was presented and compared with the design temperature. (author)

A new shape specimens determined the J1c value of nuclear pressure vessel steel

International Nuclear Information System (INIS)

Xu, W.Q.

1989-01-01

The J integral has two basic definitions, a two-dimensional energy line integral definition and an energy rate definition. The line integral definition cannot be used for this experimental determination. The energy rate definition can be used but the procedure is somewhat laborious. Methods were developed for more easily determining J by approximation formulas. The first of these were where J could be estimated with reasonable accuracy for a deeply cracked bend-type specimen. This method is slightly inaccurate. This paper is concerned with a new shape specimen. It is called the W-shape specimen. The W-shape specimens are smaller volume than the compact specimens. It is convenient to operate the W-shape specimens in hot cell. It can be put into surveillance capsules and can also do specimen irradiation in engineering test reactor

PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

Energy Technology Data Exchange (ETDEWEB)

Davenport, Michael; Petti, D. A.

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

Vacancy defects in electron irradiated RPV steels studied by positron lifetime

Energy Technology Data Exchange (ETDEWEB)

Moser, P; Li, X H [CEA Centre d` Etudes de Grenoble, 38 (France). Dept. de Recherche Fondamentale sur la Matiere Condensee; Akamatsu, M; Van Duysen, J C [Electricite de France (EDF), 77 - Ecuelles (France)

1994-12-31

Specimens of French RPV (reactor pressure vessels) steels at different rates of segregation have been irradiated at 150 and 288 deg C with 3 MeV electrons (irradiation dose: 4*10{sup 19} e-/cm{sup 2}). Vacancy defects are studied by positron lifetime measurements before and after irradiation and at each step of isochronal annealing. After 150 deg C irradiation, a recovery step is observed in both specimens, for annealing treatments in the range 220-370 deg C and is attributed to the dissociation of vacancy-impurity complexes. The size of vacancy clusters never overcome 10 empty atomic volumes. If ``fresh`` dislocations are created just before irradiation, big vacancy clusters could be formed. After 288 deg C irradiation, small vacancy cluster of 4-10 empty atomic volumes are observed. (authors). 3 figs., 7 refs.

Infrared absorption studies of the annealing of irradiated diamonds

International Nuclear Information System (INIS)

Woods, G.S.

1984-01-01

Natural (types Ia and IIa) and synthetic (type Ib) diamonds have been irradiated with energetic electrons and neutrons and then heated at temperatures up to 1400 deg C. Attendant changes in the infrared absorption spectra, especially above the Raman frequency (1332 cm -1 ), have been monitored. The most prominent absorption to develop in the infrared region proper, on annealing both type Ia and type Ib specimens, whether electron- or neutron-irradiated is the H1a line at 1450 cm -1 . Measurements taken of neutron-irradiated type Ia specimens show that the strength of this line is specimen-dependent, and that it is a linear function of radiation dose. Isochronal annealing studies show that the onset of the line occurs during heating at 250 deg C for type Ia specimens and at 650 deg C for type Ib specimens. The absorption begins to weaken during heating at 1100 deg C, but it is very persistent, surviving an anneal of 4 hours at 1400 deg C, albeit with diminished intensity. Three other weaker lines at 1438, 1358 and 1355 cm -1 develop with the 1450 cm -1 line, but differ from it and from each other in subsequent annealing behaviour. Other lines were observed; these are reported and discussed. (author)

Irradiation conditions for fiber laser bonding of HAp-glass ceramics with bovine cortical bone.

Science.gov (United States)

Tadano, Shigeru; Yamada, Satoshi; Kanaoka, Masaru

2014-01-01

Orthopedic implants are widely used to repair bones and to replace articulating joint surfaces. It is important to develop an instantaneous technique for the direct bonding of bone and implant materials. The aim of this study was to develop a technique for the laser bonding of bone with an implant material like ceramics. Ceramic specimens (10 mm diameter and 1 mm thickness) were sintered with hydroxyapatite and MgO-Al2O3-SiO2 glass powders mixed in 40:60 wt% proportions. A small hole was bored at the center of a ceramic specimen. The ceramic specimen was positioned onto a bovine bone specimen and a 5 mm diameter area of the ceramic specimen was irradiated using a fiber laser beam (1070-1080 nm wavelength). As a result, the bone and the ceramic specimens bonded strongly under the irradiation conditions of a 400 W laser power and a 1.0 s exposure time. The maximum shear strength was 5.3 ± 2.3 N. A bonding substance that penetrated deeply into the bone specimen was generated around the hole in the ceramic specimen. On using the fiber laser, the ceramic specimen instantaneously bonded to the bone specimen. Further, the irradiation conditions required for the bonding were investigated.

A study of defect cluster formation in vanadium by heavy ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Sekimura, Naoto; Shirao, Yasuyuki; Morishita, Kazunori [Tokyo Univ. (Japan)

1996-10-01

Formation of defect clusters in thin foils of vanadium was investigated by heavy ion irradiation. In the very thin region of the specimens less than 20 nm, vacancy clusters were formed under gold ion irradiation, while very few clusters were detected in the specimens irradiated with 200 and 400 keV self-ions up to 1 x 10{sup 16} ions/m{sup 2}. The density of vacancy clusters were found to be strongly dependent on ion energy. Only above the critical value of kinetic energy transfer density in vanadium, vacancy clusters are considered to be formed in the cascade damage from which interstitials can escape to the specimen surface in the very thin region. (author)

Tensile and charpy impact properties of irradiated reduced-activation ferritic steels

Energy Technology Data Exchange (ETDEWEB)

Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

1996-10-01

Tensile tests were conducted on eight reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on the steels irradiated to 26-29 dpa. Irradiation was in the Fast Flux Test Facility at 365{degrees}C on steels containing 2.25-12% Cr, varying amounts of W, V, and Ta, and 0.1%C. Previously, tensile specimens were irradiated to 6-8 dpa and Charpy specimens to 6-8, 15-17, and 20-24 dpa. Tensile and Charpy specimens were also thermally aged to 20000 h at 365{degrees}C. Thermal aging had little effect on the tensile behavior or the ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in the upper-shelf energy (USE). After {approx}7 dpa, the strength of the steels increased and then remained relatively unchanged through 26-29 dpa (i.e., the strength saturated with fluence). Post-irradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness, as measured by an increase in DBTT and a decrease in the USE, remained relatively unchanged from the values after 20-24 dpa, which had been relatively unchanged from the earlier irradiations. As before, the two 9Cr steels were the most irradiation resistant.

Microstructural comparison of HT-9 irradiated in HFIR and EBR-II

International Nuclear Information System (INIS)

Gelles, D.S.

1985-05-01

A series of specimens of HT-9 heat 91354 have been examined following irradiation in HFIR to 39 dpa at 300, 400, 500 and 600 0 C and following irradiation in EBR-II to 29 dpa at 390 and 500 0 C. HFIR irradiation was found to have promoted helium bubble formation at all temperatures and voids at 400 0 C. Cavitation had not been observed at lower fluence, nor was it found in EBR-II irradiated specimens. The onset of void swelling in HFIR is attributed to helium generation. The observations provide an explanation for saturation of ductile-brittle transition temperature shifts with increasing fluence

Evolution of microstructure after irradiation creep in several austenitic steels irradiated up to 120 dpa at 320 °C

Energy Technology Data Exchange (ETDEWEB)

Renault-Laborne, A., E-mail: alexandra.renault@cea.fr [DEN-Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Garnier, J.; Malaplate, J. [DEN-Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Gavoille, P. [DEN-Service d' Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Sefta, F. [EDF R& D, MMC, Site des Renardières, F-77818, Morêt-sur-Loing Cedex (France); Tanguy, B. [DEN-Service d' Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

2016-07-15

Irradiation creep was investigated in different austenitic steels. Pressurized tubes with stresses of 127–220 MPa were irradiated in BOR-60 at 320 °C to 120 dpa. Creep behavior was dependent on both chemical composition and metallurgical state of steels. Different steels irradiated with and without stress were examined by TEM. Without stress, the irradiation produced high densities of dislocation lines and Frank loops and, depending on the type of steels, precipitates. Stress induced an increase of the precipitate mean size and density and, for some grades, an increase of the mean loop size and a decrease of their density. An anisotropy of Frank loop density or size induced by stress was not observed systematically. Dislocation line microstructure seems not to be different between the stressed and unstressed specimens. No cavities were detectable in these specimens. By comparing with the data from this work, the main irradiation creep models are discussed.

Cytologic studies on irradiated gestric cancer cells

Energy Technology Data Exchange (ETDEWEB)

Isono, S; Takeda, T; Amakasu, H; Asakawa, H; Yamada, S [Miyagi Prefectural Adult Disease Center, Natori (Japan)

1981-06-01

The smears of the biopsy and resected specimens obtained from 74 cases of irradiated gastric cancer were cytologically analyzed for effects of irradiation. Irradiation increased the amount of both necrotic materials and neutrophils in the smears. Cancer cells were decreased in number almost in inverse proportion to irradiation dose. Clusters of cancer cells shrank in size and cells were less stratified after irradiation. Irradiated cytoplasms were swollen, vacuolated and stained abnormally. Irradiation with less than 3,000 rads gave rise to swelling of cytoplasms in almost all cases. Nuclei became enlarged, multiple, pyknotic and/or stained pale after irradiation. Nuclear swelling was more remarkable in cancer cells of differentiated adenocarcinomas.

Development of out-of-pile version of instrumented irradiation capsule for determination of online creep deformation

International Nuclear Information System (INIS)

Venkatesu, Sadu; Saxena, Rajesh; Chaurasia, P.K.; Muthuganesh, M.; Murugan, S.; Venugopal, S.

2016-01-01

Materials used for fuel cladding and structural components in fast reactors can undergo significant dimensional and physical changes due to exposure to high energy neutrons. At high temperatures in nuclear environment, material undergoes considerable deformation due to thermal and irradiation creep. Diametral increase of fuel pin due to thermal and irradiation creep, apart from irradiation swelling, reduces the coolant flow area around the fuel pins affecting the effective removal of heat generated in the fuel pins. The changes due to creep can be determined by two types of material irradiation tests in reactor. The first type includes non-instrumented irradiation tests with specimen dimensional evaluations carried out in post-irradiation examinations. The second type includes instrumented irradiation tests with online monitoring and/or controlling of test conditions and real time measurement of changes in dimensions of the specimen. During instrumented irradiation tests, parameters such as specimen temperature, the load exerted on the specimen, specimen elongation, etc. can be monitored and/or controlled using suitable components such as linear variable differential transformers (LVDTs), bellows, thermocouples, etc. Instrumented irradiation experiments in reactors are relatively complex in design but can provide full information on the experimental parameters. Such benefits provide motivation for development of instrumented irradiation capsule to measure creep behavior online during in-pile instrumented irradiation tests. Out-of-pile version of the instrumented irradiation capsule for determination of online creep deformation has been developed and tested in the furnace by raising the temperature gradually up to 330 °C. This paper discusses the details of the design, assembly of experimental set up and experimental results of the out-of-pile version of instrumented capsule developed in our laboratory for determination of online creep deformation. (author)

Temperature dependence of the damage microstructures in neutron-irradiated vanadium

International Nuclear Information System (INIS)

Horton, L.L.; Farrell, K.

1983-01-01

Vanadium and vanadium with boron carbide additions (V-B 4 C) were irradiated to approx. 1 dpa in the Oak Ridge Research Reactor at controlled temperatures ranging from 455 to 925 K. The V-B 4 C alloy was enriched in 10 B, which produced approx. 3900 at. ppM helium. In the vanadium specimens, the dislocation microstructures varied from clusters of small ( . The V-B 4 C specimens contained only tangled dislocation segments. Cavities were observed in all specimens. The cavity concentration decrease and the average diameter increased with increasing irradiation temperature. At 725 K, the maximum swelling was observed in both the vanadium (0.1%) and V-B 4 C (1.4%). At comparable temperatures the cavities in the V-B 4 C specimens were smaller and more numerous than those in the vanadium specimens. Helium bubbles were found on the grain boundaries in all of the V-B 4 specimens

Mechanical properties of 1950's vintage 304 stainless steel weldment components after low temperature neutron irradiation

International Nuclear Information System (INIS)

Sindelar, R.L.; Caskey, G.R. Jr.; Thomas, J.K.; Hawthorne, J.R.; Hiser, A.L.; Lott, R.A.; Begley, J.A.; Shogan, R.P.

1991-01-01

The reactor vessels of the nuclear production reactors at the Savannah River Site (SRS) were constructed in the 1950's from Type 304 stainless steel plates welded with Type 308 stainless steel filler using the multipass metal inert gas process. An irradiated mechanical properties database has been developed for the vessel with materials from archival primary coolant system piping irradiated at low temperatures (75 to 150 degrees C) in the State University of New York at Buffalo reactor (UBR) and the High Flux Isotope Reactor (HFIR) to doses of 0.065 to 2.1 dpa. Fracture toughness, tensile, and Charpy-V impact properties of the weldment components (base, weld, and weld heat-affected-zone (HAZ)) have been measured at temperatures of 25 degrees C and 125 degrees C in the L-C and C-L orientations for materials in both the irradiated and unirradiated conditions for companion specimens. Fracture toughness and tensile properties of specimens cut from an SRS reactor vessel sidewall with doses of 0.1 and 0.5 dpa were also measured at temperatures of 25 and 125 degrees C. The irradiated materials exhibit hardening with loss of work hardenability and a reduction in toughness relative to the unirradiated materials. The HFIR-irradiated materials show an increase in yield strength between about 20% and 190% with a concomitant tensile strength increase between about 15% to 30%. The elastic-plastic fracture toughness parameters and Charpy-V energy absorption both decrease and show only a slight sensitivity to dose. The irradiation-induced decrease in the elastic-plastic fracture toughness (J def at 1 mm crack extension) is between 20% to 65%; the range of J 1C values are 72.8 to 366 kJ/m 2 for the irradiated materials. Similarly, Charpy V-notch results show a 40% to 60% decrease in impact energies

The growth of intra-granular bubbles in post-irradiation annealed UO2 fuel

International Nuclear Information System (INIS)

White, R.J.

2001-01-01

Post-irradiation examinations of low temperature irradiated UO 2 reveal large numbers of very small intra-granular bubbles, typically of around 1 nm diameter. During high temperature reactor transients these bubbles act as sinks for fission gas atoms and vacancies and can give rise to large volumetric swellings, sometimes of the order of 10%. Under irradiation conditions, the nucleation and growth of these bubbles is determined by a balance between irradiation-induced nucleation, diffusional growth and an irradiation induced re-solution mechanism. This conceptual picture is, however, incomplete because in the absence of irradiation the model predicts that the bubble population present from the pre-irradiation would act as the dominant sink for fission gas atoms resulting in large intra-granular swellings and little or no fission gas release. In practice, large fission gas releases are observed from post-irradiation annealed fuel. A recent series of experiments addressed the issue of fission gas release and swelling in post-irradiation annealed UO 2 originating from Advanced Gas Cooled Reactor (AGR) fuel which had been ramp tested in the Halden Test reactor. Specimens of fuel were subjected to transient heating at ramp rates of 0.5 deg. C/s and 20 deg. C/s to target temperatures between 1600 deg. C and 1900 deg. C. The release of fission gas was monitored during the tests. Subsequently, the fuel was subjected to post-irradiation examination involving detailed Scanning Electron Microscopy (SEM) analysis. Bubble-size distributions were obtained from seventeen specimens, which entailed the measurement of nearly 26,000 intra-granular bubbles. The analysis reveals that the bubble densities remain approximately invariant during the anneals and the bubble-size distributions exhibit long exponential tails in which the largest bubbles are present in concentrations of 10 4 or 10 5 lower than the concentrations of the average sized bubbles. Detailed modelling of the bubble

Irradiation sensitivity of human and porcine mesenchymal stem cells

International Nuclear Information System (INIS)

Singh, S.

2009-01-01

Surgical resection, chemotherapy, radiotherapy, and combinations thereof are a plethora of possible treatment modalities of head and neck malignancies. Treatment regimens including radiotherapy however put jaws at risk of subsequent osteoradionecrosis. Besides cancer cells, irradiation impacts on all tissue-inherent cells, including mesenchymal stem cells (MSCs). Since it is the bone and bone marrow MSC, which contributes to bone regeneration through proliferation and osteogenic differentiation of its progeny, the influence of irradiation on MSC viability and the respective differentiation capacity appears to be critical. However to date, only a few reports picked MSCs role out as a pivotal topic. As a first attempt, we irradiated human bone derived MSC in vitro. With increasing doses the cells self-renewal capabilities were greatly reduced. Notably however, the mitotically stalled cells were still capable of differentiating into osteoblasts and preadipocytes. Next, the mandibles of Sus scrofa domestica were irradiated with a total dose of 18 Gy. At different time points post radiatio, MSCs were isolated from bone autopsies. In comparison between irradiated and non- irradiated samples, no significant differences regarding the proliferation and osteogenic differentiation potential of tissue specific MSC became apparent Therefore, pig mandibles were irradiated with doses of 9 and 18 Gy, and MSCs were isolated immediately afterwards. No significant differences between the untreated and bone irradiated with 9 Gy with respect of proliferation and osteogenic differentiation were observed. Cells isolated from 18 Gy irradiated specimens exhibited a greatly reduced osteogenic differentiation capacity, and during the first two weeks proliferation rates of explanted cells were greatly diminished. Thereafter, cells recovered and showed proliferation behaviour comparable to control samples. These results imply that MSCs can cope with irradiation up to relatively high doses

Effects of irradiation on strength and toughness of commercial LWR vessel cladding

International Nuclear Information System (INIS)

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

1987-01-01

The potential for stainless steel cladding to improve the fracture behavior of an operating nuclear reactor pressure vessel, particularly during certain overcooling transients, may depend greatly on the properties of the irradiated cladding. Therefore, weld overlay cladding irradiated at temperatures and to fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the three-wire series-arc commercial method. Cladding was applied in three layers to provide adequate thickness for the fabrication of test specimens. The three-wire series-arc procedure, developed by Combustion Engineering, Inc., Chattanooga, Tennessee, produced a highly controlled weld chemistry, microstructure, and fracture properties in all three layers of the weld. Charpy V-notch and tensile specimens were irradiated at 288 0 C to fluence levels of 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV). Postirradiation testing results show that, in the test temperature range from -125 to 288 0 C, the yield strength increased by 8 to 30%, ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing, due to the dominance of delta-ferrite failures at low temperatures. On the upper shelf, energy was reduced, due to irradiation exposure, 15 and 20%, while the lateral expansion was reduced 43 and 41%, at 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV), respectively. In addition, radiation damage resulted in 13 and 28 0 C shifts of the Charpy impact transition temperature at the 41-J level for the low and high fluences, respectively

Neutron metrology in the HFR. Steel irradiation R139-805 (SINAS)

International Nuclear Information System (INIS)

Baard, J.H.

1996-10-01

The experiment R139-80 is part of a material program to test austenitic stainless steel of different types and has been irradiated in the HFR Petten. This report presents the final metrology results obtained from activation monitors in the specimen holder, coded as R139-805. Data about the helium production as well as the number of displacements per atom are included. The irradiation circumstances for this experiment, carried out in a TRIO type capsule in HFR position F2, represent the conditions at the first wall of NET (Next European Torus). The aim of this irradiation for specimen holder R139-805 was to reach a damage level of 2.4 dpa at a temperature of 325 C. However, the specimens have been irradiated up to a damage level of 1.7-2.0 dpa. The main results of the thermal and fast neutron fluence measurements are presented in tables 2 and 3 as well as in the figure 2. (orig.)

Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

International Nuclear Information System (INIS)

Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

1998-01-01

This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

Anticorrosion ion implantation of fragments of zirconium fuel can specimens

International Nuclear Information System (INIS)

Kalin, B.A.; Osipov, V.V.; Volkov, N.V.; Khernov, V.Yu.

2001-01-01

Aimed at the study of specific features of oxide film formation in the initial stage of Eh110 and Eh635 alloy fuel can oxidation the modification of tubular specimen surfaces is performed using an ion mixing technique, and the structure of oxide films produced in a steam-water environment is investigated. Using the method of vacuum vapor deposition the outer surface of specimens is coated with alloying element films irradiated by a polyenergetic Ar + ion beam with a 10 keV mean energy up to radiation doses of (7-10) x 10 17 ion/cm 2 . Monatomic (Al, Fe, Cu, Cr, Mo, Sn) or diatomic (Al-Fe, Al-Mo, Al-Sn, Fe-Cu, Fe-Mo, Fe-Sn, Cr-Mo, Cr-Sn) implantation into a zirconium cladding occurs under irradiation effect. The positive influence of combined intrusion of Al and other elements is revealed. The presence of Al atoms enhances the oxide film structure. The least ZeO 2 film thickness is observed when alloying with molybdenum, Al-Fe, Al-Mo and Al-Sn [ru

Splenic injury caused by therapeutic irradiation

International Nuclear Information System (INIS)

Dailey, M.O.; Coleman, C.N.; Fajardo, L.F.

1981-01-01

Splenic irradiation in the course of therapy for lymphoma can result in functional deficit, sometimes as severe as that caused by splenectomy, placing the patient at risk for fatal infection. We examined 33 spleens obtained at necropsy from patients irradiated for lymphomas (mainly Hodgkin's disease) and compared them with 18 nonirradiated spleens from similar patients. One to 8 years after a mean radiation dose of 3899 rads, fractionated over 5-6 weeks, most irradiated spleens were small (average weight 75 g) and had thick, wrinkled capsules, often with focal hemorrhage. There was collapse of the parenchyma, with close apposition of trabeculae and mild to severe diffuse fibrosis of the red pulp. Lymphocyte depletion was obvious in more than 50% of the specimens. The most consistent alteration was myointimal proliferation of arteries. Significant intimal thickening was seen only in the irradiated specimens. Similar myointimal changes were found in the veins of three cases. While none of these changes is specific, their combination appears to be characteristic of delayed radiation injury to the spleen

High Temperature Tensile Properties of Unirradiated and Neutron Irradiated 20 Cr-35 Ni Austenitic Steel

Energy Technology Data Exchange (ETDEWEB)

Roy, R B; Solly, B

1966-12-15

The tensile properties of an unirradiated and neutron irradiated (at 40 deg C) 20 % Cr, 35 % Ni austenitic steel have been studied at 650 deg C, 750 deg C and 820 deg C. The tensile elongation and mode of fracture (transgranular) of unirradiated specimens tested at room temperature and 650 deg C are almost identical. At 750 deg C and 820 deg C the elongation decreases considerably and a large part of the total elongation is non-uniform. Furthermore, the mode of fracture at these temperatures is intergranular and microscopic evidence suggests that fracture is caused by formation and linkup of grain boundary cavities. YS and UTS decrease monotonically with temperature. Irradiated specimens show a further decrease in ductility and an increase in the tendency to grain boundary cracking. Irradiation has no significant effect on the YS, but the UTS are reduced. The embrittlement of the irradiated specimens is attributed to the presence of He and Li atoms produced during irradiation and the possible mechanisms are discussed. Prolonged annealing of irradiated and unirradiated specimens at 650 deg C appears to have no significant effect on tensile properties.

Effect of specimen size on the upper shelf energy of ferritic steels

International Nuclear Information System (INIS)

Kumar, A.S.

1990-01-01

A methodology is proposed that can be used to predict the upper shelf energy (USE) of ferritic steels based on subsize specimen data. The proposed methodology utilizes the partitioning of the USE into energies required for crack initiation and crack propagation. Notched-only Charpy specimens are used in conjunction with precracked specimens to separate the two components. An unirradiated ferritic steel, HT-9, was used to demonstrate the validity of the methodology. Unlike previous correlations that were limited in their applicability to either highly ductile or brittle material, the proposed methodology is expected to be applicable over a wide range of ductility and to be particularly useful for materials that harden significantly during irradiation

Pneumatic capsule with a measuring system for in-pile irradiation

International Nuclear Information System (INIS)

Oshima, Keiichi; Yamazaki, Yasaburo; Hirata, Mitsuho; Ishii, Toshio; Shimozawa, Ryohei.

1967-01-01

A prior-art in-pile irradiation apparatus wherein a rabbit containing an irradiation specimen therein is inserted into and removed from a pile by a pneumatic system does not include means for measuring the temperature and pressure of the specimen under irradiation. When the rabbit is deformed during irradiation, it cannot be reliably recovered. A pneumatic capsule assembly with a measuring system according to this invention has a double structure which consists of an inner capsule containing the specimen therein and an outer capsule evacuated or filled with a gas. A thermocouple lace wire and a strain gauge are welded on the outside surface of the inner capsule as detection terminals for measuring the temperature and pressure. A rupture plate which bursts when the pressure in the inner capsule reaches a predetermined value is provided at a part of the inner capsule, and a fin for heat transmission is provided between the inner and outer capsules to thus prevent the deformation of the pneumatic capsule assembly as a whole. (Takasuka, S.)

Physico Mechanical Properties of Irradiated Waste Rubber Cement Mortar

International Nuclear Information System (INIS)

Younes, M.M.

2010-01-01

In the present study a partial replacement of aggregate with two different ratios of waste rubber (5%, 10%) with the addition of a constant ratio of rice husk ash (RHA), 5% was carried out. The hardened cement mortar used the optimum water of consistency. The specimens were molded into 1 inch cubic moulds .The specimens were first cured for 24 hours, at 100% relative humidity and then cured under tap water for 3, 7 and 28 days followed by irradiation at different doses of gamma irradiation namely 5 and 10 kGy. The physico-chemical and mechanical properties such as compressive strength, total porosity and bulk density were studied for the three types of specimens. The results showed that the values of the compressive strength, bulk density and chemically combined water of the blended cement mortar paste (OPC-RHA) increase ,while blended cement mortar paste with 5% RHA and 5, 10% waste rubber decrease. The results were confirmed by scanning electron microscopy and thermal behavior of the specimens. Also, it was observed that the irradiated sample was thermally more stable than the unirradiated one

Features of structural response of mechanically loaded crystallites to irradiation

Energy Technology Data Exchange (ETDEWEB)

Korchuganov, Aleksandr V., E-mail: avkor@ispms.ru [Institute of Strength Physics and Materials Science SB RAS, Tomsk, 634055 (Russian Federation); National Research Tomsk State University, Tomsk, 634050 (Russian Federation)

2015-10-27

A molecular dynamics method is employed to investigate the origin and evolution of plastic deformation in elastically deformed iron and vanadium crystallites due to atomic displacement cascades. Elastic stress states of crystallites result from different degrees of specimen deformation. Crystallites are deformed under constant-volume conditions. Atomic displacement cascades with the primary knock-on atom energy up to 50 keV are generated in loaded specimens. It is shown that irradiation may cause not only the Frenkel pair formation but also large-scale structural rearrangements outside the irradiated area, which prove to be similar to rearrangements proceeding by the twinning mechanism in mechanically loaded specimens.

Moessbauer study of defects in molybdenum and chromium irradiated with ions

International Nuclear Information System (INIS)

Troyan, V.A.; Bogdanov, V.V.; Ivanyushkin, E.M.; Pen'kov, Yu.P.

1980-01-01

Effects of ion irradiation of monocrystalline molybdenum and polycrystalline chromium with Co-57 impurity were studied by Moessbauer effect. Molybdenum specimens were irradiated by He + ions at accelerators with 40 keV energy. Chromium specimens were irradiated by hydrogen ions with 1.2 MeV energy up to integral 2x10 17 -2x10 19 ion/cm 2 doses. It is shown, that defect introduction into the source matrix by irradiation results in change of gamma-resonance line form and effect value. The observed effects of defect influence on spectrum parameters are discussed. It is concluded, that study of Moessbauer spectra parameters of diluted Co-57 solutions in matrices of different metals permits to determine dynamics of movement of impurity atoms and defects in metals irradiated with ions [ru

Irradiation-enhanced and-induced mass transport

International Nuclear Information System (INIS)

Rehn, L.E.

1989-01-01

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

Lymph node retrieval in abdominoperineal surgical specimen is radiation time-dependent

Directory of Open Access Journals (Sweden)

Allal Abdelkarim S

2006-06-01

Full Text Available Abstract Background A low yield of lymph nodes (LN in abdominoperineal resection (APR specimen has been associated with preoperative radiation therapy (XRT in population-based studies, which may preclude adequate staging of anorectal carcinomas. We hypothesized that the number of LN retrieved in APR specimen was correlated with the dose and the timing of pelvic irradiation. Patients and methods We performed a retrospective study of 102 patients who underwent APR in a single institution between 1980 and 2004. Pathological reports were reviewed and the number of lymph nodes retrieved in APR specimens was correlated with: 1 Preoperative radiation; 2 Dose of pelvic irradiation; and 3 Time interval between the end of XRT and surgery. Results There were 61 men and 41 women, with a median age of 66 (range 25–89 years. There were 12 patients operated for squamous cell carcinoma of the anal canal (SCCA and 90 for rectal cancer. 83% and 46% of patients with anal and rectal cancer respectively underwent radical/neoadjuvant radiotherapy. The mean ± SD number of LN in APR specimen was 9.2 ± 5.9. The mean number of LN in APR specimen was significantly lower in patients who underwent preoperative XRT (8 ± 5.5 vs. 10.5 ± 6.1, Mann-Whitney U test, p = 0.02. The mean number of LN was not significantly different after XRT in patients with SCCA than in patients with rectal cancer (6.2 ± 5.3 vs. 7.8 ± 5.3, p = 0.33. Finally, there was an inverse correlation between the yield of LN and the time elapsed between XRT and surgery (linear regression coefficient r = -0.32, p = 0.03. Conclusion Our data indicate that: 1 radiation therapy affects the yield of LN retrieval in APR specimen; 2 this impact is time-dependent. These findings have important implications with regard to anatomic-pathological staging of anal and rectal cancers and subsequent decision-making regarding adjuvant chemotherapy.

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-08-31

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Device for investigating subcritical crack growth of RPV steel specimens under BWR conditions

International Nuclear Information System (INIS)

Anders, D.; Ahlf, J.

1983-01-01

An experiment is being prepared to investigate the subcritical crack growth of RPV steel specimens under cyclic load and under the environmental conditions of a BWR with regard to primary water and irradiation. The experiment will be carried out in the VAK reactor Kahl which is a boiling water reactor operating at 71 bar, 286 0 C and generating 16 MW/sub e/. The experimental setup is composed of an open frame to which a string consisting of five compact tension speciments (40 mm thickness) and connecting links is fixed. The specimen chain is set under cyclic load by a pneumatically actuated bellows unit which is attached to the frame top. Specimen strain and crack opening are measured by linear differential transformers; for temperature distribution measurements in the specimens thermocouples are applied

In-service irradiated and aged material evaluations

International Nuclear Information System (INIS)

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

1995-01-01

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

The natural aging of austenitic stainless steels irradiated with fast neutrons

Science.gov (United States)

Rofman, O. V.; Maksimkin, O. P.; Tsay, K. V.; Koyanbayev, Ye. T.; Short, M. P.

2018-02-01

Much of today's research in nuclear materials relies heavily on archived, historical specimens, as neutron irradiation facilities become ever more scarce. These materials are subject to many processes of stress- and irradiation-induced microstructural evolution, including those during and after irradiation. The latter of these, referring to specimens "naturally aged" in ambient laboratory conditions, receives far less attention. The long and slow set of rare defect migration and interaction events during natural aging can significantly change material properties over decadal timescales. This paper presents the results of natural aging carried out over 15 years on austenitic stainless steels from a BN-350 fast breeder reactor, each with its own irradiation, stress state, and natural aging history. Natural aging is shown to significantly reduce hardness in these steels by 10-25% and partially alleviate stress-induced hardening over this timescale, showing that materials evolve back towards equilibrium even at such a low temperature. The results in this study have significant implications to any nuclear materials research program which uses historical specimens from previous irradiations, challenging the commonly held assumption that materials "on the shelf" do not evolve.

Positron annihilation and thermally stimulated current of electron beam irradiated polyetheretherketone

Energy Technology Data Exchange (ETDEWEB)

Fujita, Shigetaka; Shinyama, Katsuyoshi; Baba, Makoto [Hachinohe Inst. of Tech., Hachinohe, Aomori (Japan); Suzuki, Takenori

1997-03-01

Positron lifetime measurements were applied to electron beam irradiated poly(ether-ether-ketone). The lifetime, {tau}{sub 3}, of the ortho-positronium of unirradiated and 5 MGy irradiated specimen became rapidly longer above about 150degC. {tau}{sub 3} of 50 MGy and 100 MGy irradiated specimen was shorter than that of unirradiated one. Thermally stimulated current (TSC) decreased with increasing the dose before voltage application. In the case of voltage application, a TSC peak appeared and the peak value decreased with increased the dose. The correlation between the results of positron annihilation and TSC was investigated. (author)

Beam-induced temperature changes in HVEM irradiations

International Nuclear Information System (INIS)

Garner, F.A.; Thomas, L.E.; Gelles, D.S.

1975-01-01

The peak value of the temperature distribution induced by energy loss of 1.0 MeV electrons in traversing a typical HVEM irradiation specimen can be very substantial. The origin and various features of this distribution were analyzed for a variety of specimen geometries. The major parametric dependencies are shown to be relatively independent of specimen geometry, however, and allow the definition of a scaling relationship that can be employed to predict temperature rises in materials that cannot be measured directly. The use of this scaling relationship requires that the experimenter minimize perturbations of the heat flow due to proximity of the central hole in the specimen. An experimental method of determining directly the magnitude and distribution of beam-induced temperature profiles was developed which utilizes the order-disorder transformation in Fe 3 Al and Cu 3 Au. Scaling of experimentally determined temperature changes leads to more realistic estimates of the total temperature rise than are currently available in various literature tabulations. The factors which determine the optimum selection of irradiation parameters for a given experiment are also discussed

Post-irradiation mechanical tests on F82H EB and TIG welds

International Nuclear Information System (INIS)

Rensman, J.; Osch, E.V. van; Horsten, M.G.; D'Hulst, D.S.

2000-01-01

The irradiation behaviour of electron beam (EB) and tungsten inert gas (TIG) welded joints of the reduced-activation martensitic steel IEA heat F82H-mod. was investigated by neutron irradiation experiments in the high flux reactor (HFR) in Petten. Mechanical test specimens, such as tensile specimens and KLST-type Charpy impact specimens, were neutron irradiated up to a dose level of 2-3 dpa at a temperature of 300 deg. C in the HFR reactor in Petten. The tensile results for TIG and EB welds are as expected with practically no strain hardening capacity left. Considering impact properties, there is a large variation in impact properties for the TIG weld. The irradiation tends to shift the DBTT of particularly the EB welds to very high values, some cases even above +250 deg. C. PWHT of EB-welded material gives a significant improvement of the DBTT and USE compared to the as-welded condition

Small specimen test technology of fracture toughness in structural material F82H steel for fusion nuclear reactors

International Nuclear Information System (INIS)

Wakai, Eiichi; Ohtsuka, Hideo; Jitsukawa, Shiro; Matsukawa, Shingo; Ando, Masami

2006-03-01

Small specimen test technology (SSTT) has been developed to investigate mechanical properties of nuclear materials. SSTT has been driven by limited availability of effective irradiation volumes in test reactors and accelerator-based neutron and charged particle sources, and it is very useful for the reduction of waste materials produced in nuclear engineering. In this study new bend test machines have been developed to obtain fracture behaviors of F82H steel for very small bend specimens of pre-cracked t/2-1/3CVN (Charpy V-notch) with 20 mm-length and DFMB (deformation and fracture mini bend specimen) with 9 mm-length and disk compact tension of 0.18DCT type, and fracture behaviors were examined to evaluate DBTT (ductile-brittle transition temperature) at temperature from -180 to 25degC. The effect of specimen size on DBTT of F82H steel was also examined by using Charpy type specimens such as 1/2t-CVN, 1/3CVN and t/2-1/3CVN. In this paper, it also provides the information of the specimens irradiated at 250degC and 350degC to about 2 dpa in the capsule of 04M-67A and 04M-68A of JMTR experiments. (author)

Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

Energy Technology Data Exchange (ETDEWEB)

Singh, B N; Edwards, D J; Horsewell, A; Toft, P

1995-09-01

The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al{sub 2}O{sub 3}, CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10{sup 17} n/m{sup 2}s (E > 1 MeV, i.e. a dose rate of {approx}5 x 10{sup -8} dpa/s) to fluences of 5 x 10{sup 22}, 5 x 10{sup 23} and 1 x 10{sup 24} n/m{sup 2} (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al{sub 2}O{sub 3} (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al{sub 2}O{sub 3}, (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al{sub 2}O{sub 3} alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs.

Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

International Nuclear Information System (INIS)

Singh, B.N.; Edwards, D.J.; Horsewell, A.; Toft, P.

1995-09-01

The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al 2 O 3 , CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10 17 n/m 2 s (E > 1 MeV, i.e. a dose rate of ∼5 x 10 -8 dpa/s) to fluences of 5 x 10 22 , 5 x 10 23 and 1 x 10 24 n/m 2 (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al 2 O 3 (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al 2 O 3 , (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al 2 O 3 alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs

Evolution of cleared channels in neutron-irradiated pure copper as a function of tensile strain

DEFF Research Database (Denmark)

Edwards, D.J.; Singh, B.N.

2004-01-01

Tensile specimens of pure copper were neutron irradiated at similar to323 K to a displacement dose of 0.3 dpa (displacement per atom). Five irradiated specimens were tensile tested at 300 K, but four of the specimens were stopped at specific strains -just before the yield point at similar to90......% of the macroscopic yield, at 1.5% and 5% elongation, and near the ultimate tensile strength at 14.5% elongation, with the 5th specimen tested to failure (e(T) = 22%). SEM and TEM characterization of the deformed specimens revealed that the plastic strain was confined primarily to the 'cleared' channels only...

Fracture toughness of Charpy-size compound specimens and its application in engineering

International Nuclear Information System (INIS)

Zhang, X.P.; Shi, Y.W.

1994-01-01

The use of a pre-cracked Charpy-size specimen with a side-groove to evaluate the fracture toughness of materials has been researched and considered. This method not only satisfies the demand for small-size specimens in surveillance tests of fracture toughness but also avoids using complicated physical methods to monitor the initial conditions of crack propagation. For most materials this method has solved the problem in which the small-size specimen did not satisfy the valid conditions of a fracture toughness measurement. In order to obtain more information from neutron-irradiated sample specimens and raise the reliability of fracture toughness surveillance tests, it has been considered more important to repeatedly exploit the broken Charpy-size specimen tested in the surveillance test, and to make it renewable. In this work, on the renewing design and utilization of Charpy-size specimens, nine data on fracture toughness can be obtained from one pre-cracked side-grooved Charpy-size specimen, while at present usually only one to three data on fracture toughness can be obtained from one Charpy-size specimen. Thus, it is found that the new method would improve the reliability of fracture toughness surveillance testing and evaluation. In addition, some factors that affect the optimum design of pre-cracked deep side-groove Charpy-size compound specimens have also been discussed. (author)

Tensile mechanical properties of a stainless steel irradiated up to 19 dpa in the Swiss spallation neutron source

Energy Technology Data Exchange (ETDEWEB)

Saito, Shigeru, E-mail: saito.shigeru@jaea.go.jp [JAEA, J-PARC Center, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Kikuchi, Kenji [Ibaraki Univ., iFRC, Tokai-mura, Ibaraki-ken 319-1106 (Japan); Hamaguchi, Dai [JAEA, J-PARC Center, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Usami, Kouji; Endo, Shinya; Ono, Katsuto; Matsui, Hiroki [JAEA, Dept. of Hot Laboratories, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Kawai, Masayoshi [KEK, Tsukuba-shi, Ibaraki-ken 305-0801 (Japan); Dai, Yong [PSI, Spallation Source Division, Villigen PSI (Switzerland)

2012-12-15

To evaluate the lifetime of the beam window of an accelerator-driven transmutation system (ADS), post irradiation examination (PIE) of the STIP (SINQ target irradiation program, SINQ; Swiss spallation neutron source) specimens was carried out. The specimens tested in this study were made from the austenitic steel Japan primary candidate alloy (JPCA). The specimens were irradiated at SINQ Target 4 (STIP-II) with high-energy protons and spallation neutrons. The irradiation conditions were as follows: the proton energy was 580 MeV, irradiation temperatures ranged from 100 to 430 Degree-Sign C, and displacement damage levels ranged from 7.1 to 19.5 dpa. Tensile tests were performed in air at room temperature (RT), 250 Degree-Sign C and 350 Degree-Sign C. Fracture surface observation after the tests was done by Scanning electron microscope (SEM). Results of the tensile tests performed at R.T. showed the extra hardening of JPCA at higher dose compared to the fission neutron irradiated data. At the higher temperatures, 250 Degree-Sign C and 350 Degree-Sign C, the extra hardening was not observed. Degradation of ductility bottomed around 10 dpa, and specimens kept their ductility until 19.5 dpa. All specimens fractured in ductile manner.

Design, fabrication and irradiation test report on HANARO instrumented capsule (03M-06U) for researches of universities in 2003

International Nuclear Information System (INIS)

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

2005-03-01

As a part of 2003 project for active utilization of HANARO, an instrumented capsule (03M-06U) was designed, fabricated and irradiated for the irradiation test of various nuclear materials under irradiation conditions requested by external researchers from universities. The basic structure of 03M-06U capsule was based on the 00M-01U, 01M-05U and 02M-05U capsules successfully irradiated in HANARO as 2000, 2001 and 2002 projects. However, because of the limited number of specimens and budget of 4 universities, the remained space of the capsule was charged with KAERI specimens for the development of the precise temperature control technology under irradiation. The material of the specimens is mainly Fe-based alloys partially mixed with Zr, Al and Cu-Ag alloys. The capsule is composed of 5 stages having many kinds of specimens and independent electric heater in each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. Various types of specimens such as tensile, Charpy, TEM, toughness, electrical resistance specimens were inserted in the capsule. The capsule was firstly irradiated in the CT test hole of HANARO of 30MW thermal output at 275∼500±10 .deg. C up to a fast neutron fluence of 5.4 x 10 20 (n/cm 2 ) (E>1.0MeV). The obtained results will be very valuable for the related researches of the users

Small Punch Test Techniques for Irradiated Materials in Hot Cell

International Nuclear Information System (INIS)

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

2006-06-01

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

Effect of cold work on tensile behavior of irradiated type 316 stainless steel

International Nuclear Information System (INIS)

Klueh, R.L.; Maziasz, P.J.

1986-01-01

Tensile specimens were irradiated in ORR at 250, 290, 450, and 500 0 C to produce a displacement damage of approx.5 dpa and 40 at. ppM He. Irradiation at 250 and 290 0 C caused an increase in yield stress and ultimate tensile strength and a decrease in ductility relative to unaged and thermally aged controls. The changes were greatest for the 20%-cold-worked steel and lowest for the 50%-cold-worked steel. Irradiation at 450 0 C caused a slight relative decrease in strength for all cold-worked conditions. A large decrease was observed at 500 0 C, with the largest decrease occurring for the 50%-cold-worked specimen. No bubble, void, or precipitate formation was observed for specimens examined by transmission electron microscopy (TEM). The irradiation hardening was correlated with Frank-loop and ''black-dot'' loop damage. A strength decrease at 500 0 C was correlated with dislocation network recovery. Comparison of tensile and TEM results from ORR-irradiated steel with those from steels irradiated in the High Flux Isotope Reactor and the Experimental Breeder Reactor indicated consistent strength and microstructure changes

Effect of Proton Irradiation on the Corrosion Behaviors of Ferritic/Martensitic Steel in Liquid Metal Environment

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeonghyeon; Kim, Tae Yong; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

2016-10-15

Liquid metal fast breeder reactors (LMFBRs) such as sodium-cooled fast reactor (SFR) and lead-cooled fast reactor (LFR) are the candidates of GEN-IV nuclear energy systems. Among various liquid metals that can be used as primary coolant material, sodium is a world widely used coolant for GEN-IV reactors. In this study, as-received Gr.92 and irradiated Gr.92 specimen in the oxygen-saturated liquid sodium were examined at high temperature for 300h. The microstructure results reveal the information of the effect of irradiation and effect of the chrome concentration in specimen. From the SRIM result, penetration distance of 40 μm in stainless steel and nominal sample thickness of 30 μm was used to avoid the damage peak and any proton implantation and From the microstructural evaluation, chromium-rich zones existed under the surface of the both of non-irradiated and irradiated materials. The irradiated materials showed chromium-rich zones with larger depths than the non-irradiated specimens.

Review of intense irradiation data and discussion on structural integrity

Energy Technology Data Exchange (ETDEWEB)

Suzuki, K.; Jitsukawa, S.; Okubo, N. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan)

2007-07-01

Full text of publication follows: Mechanical property data on austenitic stainless steels and F82H have been reviewed to discuss for developing structural integrity methodologies of intensely irradiated components such as first walls. The following have been already clarified: (1) Fracture ductility is still high even though tensile rupture elongation is reduced remarkably. (2) Strain-hardening occurs in true stress-logarithmic strain (true strain) relationship. Work-softening behavior observed in nominal stress-nominal strain curves is simply resulted from a reduction of work hardening rate accompanied by the increase of flow stress level by irradiation. The review lead to an innovative design concept for application to intensely irradiated components. A special consideration is given to unique feature of bending moment in developing design methodology for preventing ductile fracture of intensely irradiated materials. Another discussion is also made on how to simulate mechanical behavior of intensely irradiated components, because mechanical testing of component-wise specimens after intense irradiation is inevitable for the development of design concepts, although irradiation on such a large scale specimen seems to be almost impossible with current irradiation facilities. (authors)

Shear Punch Testing of BOR-60 Irradiated TEM Specimens

Energy Technology Data Exchange (ETDEWEB)

Saleh, Tarik A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Quintana, Matthew Estevan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Tobias J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-06-13

As a part of the project “High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation” an Integrated Research Program (IRP) project from the U.S. Department of Energy, Nuclear Energy University Programs (NEUP), TEM geometry samples of ferritic cladding alloys, Ni based super alloys and model alloys were irradiated in the BOR-60 reactor to ~16 dpa at ~370°C and ~400°C. Samples were sent to Los Alamos National Laboratory and subjected to shear punch testing. This report presents the results from this testing.

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

Irradiation-Induced Nanostructures

Energy Technology Data Exchange (ETDEWEB)

Birtcher, R.C.; Ewing, R.C.; Matzke, Hj.; Meldrum, A.; Newcomer, P.P.; Wang, L.M.; Wang, S.X.; Weber, W.J.

1999-08-09

This paper summarizes the results of the studies of the irradiation-induced formation of nanostructures, where the injected interstitials from the source of irradiation are not major components of the nanophase. This phenomena has been observed by in situ transmission electron microscopy (TEM) in a number of intermetallic compounds and ceramics during high-energy electron or ion irradiations when the ions completely penetrate through the specimen. Beginning with single crystals, electron or ion irradiation in a certain temperature range may result in nanostructures composed of amorphous domains and nanocrystals with either the original composition and crystal structure or new nanophases formed by decomposition of the target material. The phenomenon has also been observed in natural materials which have suffered irradiation from the decay of constituent radioactive elements and in nuclear reactor fuels which have been irradiated by fission neutrons and other fission products. The mechanisms involved in the process of this nanophase formation are discussed in terms of the evolution of displacement cascades, radiation-induced defect accumulation, radiation-induced segregation and phase decomposition, as well as the competition between irradiation-induced amorphization and recrystallization.

Titanium implants in irradiated dog mandibles

International Nuclear Information System (INIS)

Schweiger, J.W.

1989-01-01

The use of osseointegrated titanium implants has been a great benefit to selected cancer patients who otherwise would not be able to wear conventional and/or maxillofacial prostheses. Cognizant of the risk of osteoradionecrosis, we used an animal model to seek experimental evidence for successful osseointegration in bone irradiated to tumoricidal levels. Five healthy male beagle dogs received 60 gray to a previously edentulated and healed area of the right hemimandible. The left hemimandible was kept as a nonirradiated control. After 9 months, titanium implants were placed and allowed an additional 5 1/2 months to osseointegrate. At that time, block specimens were obtained, radiographed, photographed, and analyzed histologically. Although statistical significance cannot be attached to the results, osseointegration was achieved in half of the irradiated specimens

Positron trapping in heavily irradiated semiconductors

International Nuclear Information System (INIS)

Moser, P.; Pautrat, J.L.; Corbel, C.; Hautojarvi, P.

1985-01-01

Vacancy processes are studied in several heavily irradiated semiconductors. Specimens are ZnTe, CdTe, CdTe (In), InP, InP (Cr), InP (Zn) and Ge. Irradiations are made at 20 K using a 3 MeV Van de Graaff electron accelerator. Doses are 4 x 10 18 e - /cm 2 . Lifetime measurements are made at 77 K at each step of an isochronal annealing (30 min 20 K). In each specimen, the results show a significant increase of the lifetime (+ 30 at + 50 ps) which anneals out in different steps restoring the initial lifetime. The steps are sharp (ΔT/T=0.3) with the exception of InP, InP(Cr), InP(Zn), (ΔT/T=0.9). Tentative interpretations are given

Superplastic characteristics and microstructure of neutron irradiated 3Y-TZP

International Nuclear Information System (INIS)

Shibata, Taiju; Motohashi, Yoshinobu; Ishihara, Masahiro; Baba, Shinichi; Sawa, Kazuhiro

2006-01-01

Fast neutrons (energy > 1.6 x 10 -13 J) were irradiated to 3Y-TZP specimens, typical superplastic ceramics, at the fluence of 2.5 x 10 24 and 4.3 x 10 24 m -2 at JMTR of JAEA. The Vickers hardness with indentation load of 4.9 and 9.8 N at room temperature was seemed to be slightly increased by the irradiation. Through the superplastic tensile tests in a temperature range from 1623 to 1773 K with initial strain rates of 5.0 x 10 -4 and 1.0 x 10 -3 s -1 , it was found that the superplastic flow stress is decreased with increasing the neutron fluence. The microstructural features of the fractured specimens were observed by a SEM. It implies that the grain boundary microstructure of the irradiated specimens would be changed by annealing in the superplastic tests are elevated temperatures. It is quite probable that the irradiation-induced vacancy clusters might play an important role to weaken the grain boundary cohesion which may be an important factor to determine the superplastic properties, and hence they would decrease the superplastic flow stress. (author)

Conclusions regarding fracture mechanics testing and evaluation of small specimens - As evidenced by the finnish contribution to the IAEA CRP3 programme

Energy Technology Data Exchange (ETDEWEB)

Wallin, K; Valo, M; Rintamaa, R; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Ahlstrand, R [Imatran Voima Oy (IVO), Helsinki (Finland)

1994-12-31

An extensive mechanical property evaluation has been carried out on various specimens (a Japanese steel plate (JRQ), a French forging material (FFA) and a Japanese forging material (JFL)) in the as-received and irradiated conditions. The mechanical properties measured at different temperatures include Charpy-V notch and instrumented pre-cracked Charpy data and static and dynamic elastic-plastic fracture toughness based on the J-integral, with various specimen size and geometry. Test analysis lead to conclusions regarding the use of small specimen fracture mechanical tests for investigating irradiation effects: CVN{sub pc} and RCT type specimens are suitable for determining the materials fracture toughness even in the ductile/brittle transition region provided the elastic-plastic parameter K{sub JC} is applied together with a statistical size correction. These two specimen types yield equivalent results for the fracture toughness transition shift. Charpy-V appears not to be suitable for estimating the static fracture toughness transition shift. 8 refs., 11 figs.

Correlation of clinical data with fallopian tube specimen immune cells and tissue culture capacity.

Science.gov (United States)

Ramraj, Satish Kumar; Smith, Katie M; Janakiram, Naveena B; Toal, Coralee; Raman, Ankita; Benbrook, Doris Mangiaracina

2018-06-01

Human fallopian tube fimbria secretory epithelial cells (hFTSECs) are considered an origin of ovarian cancer and methods for their culture from fallopian tube specimens have been reported. Our objective was to determine whether characteristics of the donors or surgeries were associated with the capacities of fimbria specimens to generate hFTSEC cultures or their immune profiles. There were no surgical complications attributable to fallopian tube removal. Attempts to establish primary hFTSEC cultures were successful in 37 of 55 specimens (67%). Success rates did not differ significantly between specimens grouped by patient or surgery characteristics. Established cultures could be revived after cryopreservation and none became contaminated with microorganisms. Two cultures evaluated for long term growth senesced between passages 10 and 15. M1 macrophages were the predominant cell type, while all other immune cells were present at much lower percentages. IL-10 and TGF-β exhibited opposing trends with M1 and M2 macrophages. Plasma IL-10 levels exhibited significant positive correlation with patient age. In conclusion, fallopian tube fimbria specimens exhibit a pro-inflammatory phenotype and can be used to provide a source of hFTSECs that can be cultured for a limited time regardless of the donor patient age or race, or the type of surgery performed. Copyright © 2018 Elsevier Ltd. All rights reserved.

Intergranular attack observed in radiation-enhanced corrosion of mild steel

International Nuclear Information System (INIS)

Reda, R.J.; Kelly, J.L.; Harna, S.L.A.

1988-01-01

Experiments were conducted to determine the effects of gamma radiation on the corrosion of AISI 1018 mild steel in deaerated brine solutions of various sodium, magnesium, and chloride ion concentrations. Immersed metal specimens were irradiated at an exposure rate of 3 x 10/sup 5/ R/h (0.3 MR/h) for up to 1250 h at a temperature of --25 C. The corrosion rates of the irradiated specimens were found to be roughly a factor of 10 greater than the rates for the non-irradiated specimens. The radiation-enhanced corrosion rate was also found to have increased with the chloride concentration. Electron micrographs revealed that the surface morphology of the specimens exposed to irradiated brines differed greatly from the non-irradiated specimens. The non-irradiated specimens had undergone uniform corrosion, while the irradiated specimens exhibited intergranular corrosion (IGC), a phenomenon not yet observed in mild steel. An explanation for this observation is offered in terms of the relative rates of formation and recombination of radiolytic species

Dual ion beam irradiation system for in situ observation with electron microscope

International Nuclear Information System (INIS)

Tsukamoto, Tetuo; Hojou, Kiiti; Furuno, Sigemi; Otsu, Hitosi; Izui, Kazuhiko.

1993-01-01

We have developed a new in situ observation system for dynamic processes under dual ion beam irradiation. The system consists of a modified 400 keV analytical electron microscope (JEOL, JEM-4000FX) and two 40 kV ion beam accelerators. This system allows evaluation of microscopic changes of structure and chemical bonding state of materials in the dynamic processes under two kinds of ion beam irradiations, that is required for the simulation test of the first wall of nuclear fusion reactors onto which He + , H + , and H 2 + ions are irradiated simultaneously. These two ion accelerators were equipped symmetrically both sides of the electron microscope and individually controlled. Each ion beam extracted from a duo-plasmatron ion gun is bent downward by an angle of 30deg with a mass-separating magnet, and introduced into specimen chamber of the electron microscope. Inside the specimen chamber the beam is deflected again by an angle of 30deg with an electrostatic prism so as to be incident on the specimen surface. Finally, two ion beams from both side are incident on the specimen surface at an angle of 60deg. The maximum ion current density of helium is more than 250μA/cm 2 at the specimen at an ion energy of 17 keV. Images of the electron microscope during dual ion beam irradiation are observed through a TV camera and recorded with a VTR. (author)

Fractography evaluation of impact and tensile specimens from the HFBR [High Flux Beam Reactor

International Nuclear Information System (INIS)

Czajkowski, C.J.

1989-10-01

The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) has performed a fractographic examination of neutron irradiated and unirradiated tensile and Charpy ''V'' notch specimens. The evaluation was carried out using a scanning electron microscope (SEM) to evaluate the fracture mode. Photomicrographs were then evaluated to determine the extent of ductility present on the fracture surfaces of the unirradiated specimens. Ductility area measurements ranged from 4.6--9.5% on typical photomicrographs examined. 12 figs

Mechanical properties of irradiated and non-irradiated Zr1%Nb and Zircaloy claddings

International Nuclear Information System (INIS)

Griger, Agnes

2004-01-01

The mechanical properties of irradiated and non-irradiated Zr1%Nb were determined and they were compared with the analogous properties of Zircaloy-4 to establish connections between the evolution of mechanical parameters of Zr1%Nb and Zircaloy-4 cladding materials and the measure of irradiation. Samples were irradiated in the vertical channels of the Budapest Research Reactor for different time periods at 50-65 C temperature. The measure of irradiation (fluent) for different samples was estimated by means of flux measurement and using the effective irradiation time. Post irradiation uniaxial tension tests in transverse direction were carried out on ring specimens. The mechanical parameters of the Zr1%Nb alloy significantly improve due to the effect of irradiation. However, the values of mechanical parameters do not further increase when the fluent increases above 10 20 n/cm 2 . These results are in good accordance with the Russian ones [1]. Contrary to the behaviour of Zr1%Nb alloy, the mechanical parameters of the Zircaloy practically do not change on the effect of irradiation. The originally high values of ultimate tensile strength and yield stress change only slightly with the increasing fluent in the investigated fluent-region. (Author)

Change of mechanical properties of irradiated silicon iron in dependence of preliminary deformation

International Nuclear Information System (INIS)

Chirkina, L.A.; Okovit, V.S.; Khinkis, B.A.

1979-01-01

Presented are the data on the influence of the 225 MeV electron irradiation on flow limit and specific elongation of silicon iron specimens preliminary deformed by slipping and twinning. The irradiaton was carried out at the temperature up to 350 K with integral dose up to 7x10 18 el/cm 2 . The specimens were tested in the temperature range of 4-450 K. It is found that the ductile brittle transition temperature Tsub(c) and plastic deformation mode of the irradiated material heavily depends on the preliminary deformation mode. The irradiation of specimens deformed by slipping leads to the increase in transition temperature (Tsub(c)) by 80 deg and it reaches 420 K. The preliminary deformation by twinning results in the Tsub(c) increase up to 320 K

Detection of radiation-induced changes in electrochemical properties of austenitic stainless steels using miniaturized specimens and the single-loop electrochemical potentiokinetic reactivation method

International Nuclear Information System (INIS)

Inazumi, T.; Bell, G.E.C.; Kenik, E.A.; Kiuchi, K.

1993-01-01

Single-loop electrochemical potentiokinetic reactivation testing of miniaturized (TEM) specimens can provide reliable data comparable to data obtained with larger specimens. Significant changes in electrochemical properties (increased reactivation current and Flade potential) were detected for PCA and type 316 stainless steels irradiated at 200--420 degrees C up to 7--9 dpa. Irradiations in the FFTF Materials Open Test Assembly and in the Oak Ridge Research Reactor are reported on. 45 figs., 5 tabs., 52 refs

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-12-15

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

Effect of LASER Irradiation on the Shear Bond Strength of Zirconia Ceramic Surface to Dentin

Directory of Open Access Journals (Sweden)

Sima Shahabi

2012-09-01

Full Text Available Background and Aims: Reliable bonding between tooth substrate and zirconia-based ceramic restorations is always of great importance. The laser might be useful for treatment of ceramic surfaces. The aim of the present study was to investigate the effect of laser irradiation on the shear bond strength of zirconia ceramic surface to dentin. Materials and Methods: In this experimental in vitro study, 40 Cercon zirconia ceramic blocks were fabricated. The surface treatment was performed using sandblasting with 50-micrometer Al2O3, CO2 laser, or Nd:YAG laser in each test groups. After that, the specimens were cemented to human dentin with resin cement. The shear bond strength of ceramics to dentin was determined and failure mode of each specimen was analyzed by stereo-microscope and SEM investigations. The data were statistically analyzed by one-way analysis of variance and Tukey multiple comparisons. The surface morphology of one specimen from each group was investigated under SEM. Results: The mean shear bond strength of zirconia ceramic to dentin was 7.79±3.03, 9.85±4.69, 14.92±4.48 MPa for CO2 irradiated, Nd:YAG irradiated, and sandblasted specimens, respectively. Significant differences were noted between CO2 (P=0.001 and Nd:YAG laser (P=0.017 irradiated specimens with sandblasted specimens. No significant differences were observed between two laser methods (P=0.47. The mode of bond failure was predominantly adhesive in test groups (CO2 irradiated specimens: 75%, Nd:YAG irradiated: 66.7%, and sandblasting: 41.7%. Conclusion: Under the limitations of the present study, surface treatment of zirconia ceramics using CO2 and Nd:YAG lasers was not able to produce adequate bond strength with dentin surfaces in comparison to sandblasting technique. Therefore, the use of lasers with the mentioned parameters may not be recommended for the surface treatment of Cercon ceramics.

Microstructure and mechanical properties of neutron irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Ishitsuka, E.; Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Terai, T.; Tanaka, S.

1998-01-01

Microstructure and mechanical properties of the neutron irradiated beryllium with total fast neutron fluences of 1.3 - 4.3 x 10{sup 21} n/cm{sup 2} (E>1 MeV) at 327 - 616degC were studied. Swelling increased by high irradiation temperature, high fluence, and by the small grain size and high impurity. Obvious decreasing of the fracture stress was observed in the bending test and in small grain specimens which had many helium bubbles on the grain boundary. Decreasing of the fracture stress for small grain specimens was presumably caused by crack propagation on the grain boundaries which weekend by helium bubbles. (author)

Thiel embalming fluid--a new way to revive formalin-fixed cadaveric specimens.

Science.gov (United States)

Hunter, Amanda; Eisma, Roos; Lamb, Clare

2014-09-01

By soft fixing cadavers using the Thiel embalming method, our cadavers now exhibit a greater degree of flexibility and color retention compared to that of traditional formalin-fixed cadavers. The aim of this experiment was to discover whether Thiel embalming fluid could be used to revive and soften the muscles of formalin-fixed prosected specimens. Earlier this year, two severely dehydrated formalin-fixed forearm and hand specimens were fully submerged in a tank containing Thiel embalming fluid. After a period of six months the specimens were removed from the tank and noticeable changes were observed in flexibility, quality of the tissue, and color of the specimens. © 2014 Wiley Periodicals, Inc.

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-01

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

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

International Nuclear Information System (INIS)

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

1994-12-01

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

Effect Of Irradiation Temperature and Dose On Mechanical Properties And Fracture Characteristics Of Cu//SS Joints For ITER

International Nuclear Information System (INIS)

Fabritsiev, S.A.; Pokrovsky, A.S.; Peacock, A.; Roedig, M.; Linke, J.; Gervash, A.; Barabash, V.

2007-01-01

Full text of publication follows: By now, a number of technologies have been proposed for the production of Cu//SS joints for ITER, such as brazing, friction welding, HIP and cast-copper-to-steel (CC). The two last-mentioned technologies ensure sufficiently high mechanical properties and a high joint quality, when unirradiated. The data, however, on mechanical characteristics of irradiated of Cu//SS HIP joints are limited. In this paper, the authors present the results of investigations into the mechanical characteristics after irradiation of GlidCopAl25/316L(N) and Cu-Cr-Zr/316L(N)-type joints produced by the HIP and CC technologies. Specimens of the joints were irradiated in the RBT-6 reactor in the dose range of 10 -3 - 10 -1 dpa at T irr = 200 deg. C and 300 deg. C. The tensile stress-strain curves for irradiated and unirradiated joint specimens show deformation processes occurring in both the Cu and SS parts of the specimens. Irradiation at T irr = 200 deg. C causes strengthening of the joints specimens (by about 100 MPa at the maximum dose). The uniform elongation drops from 8% in the initial state to 2-3 %. But the total elongation remains at a relatively high level of ∼ 7%. Irradiation at T irr = 300 deg. C causes a slight strengthening of the joints specimens (∼30 MPa). The uniform elongation remains unchanged at ∼ 7%. The total elongation also maintains a relatively high level of ∼9-13%. SEM investigations revealed that fracture occurs only in the copper part of the irradiated specimens, and ductile trans-crystalline fracture predominates in the joints. 3D finite element analysis of the tensile test indicates that the concentration of stresses and deformations in the copper layer adjacent to the joint line is responsible for this typical failure of the irradiated joints specimens. Comparison of the behavior of the joints irradiated at T irr = 200 deg. C and 300 deg. C indicate an increased embrittlement at lower irradiation temperatures. At a

Short Communication on "In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures"

Science.gov (United States)

Miao, Yinbin; Harp, Jason; Mo, Kun; Bhattacharya, Sumit; Baldo, Peter; Yacout, Abdellatif M.

2017-02-01

The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

Temperature dependence of the damage microstructures in neutron-irradiated vanadium

Energy Technology Data Exchange (ETDEWEB)

Horton, L.L.; Farrell, K.

1983-01-01

Vanadium and vanadium with boron carbide additions (V-B/sub 4/C) were irradiated to approx. 1 dpa in the Oak Ridge Research Reactor at controlled temperatures ranging from 455 to 925 K. The V-B/sub 4/C alloy was enriched in /sup 10/B, which produced approx. 3900 at. ppM helium. In the vanadium specimens, the dislocation microstructures varied from clusters of small (< 50 nm diam) dislocation loops (455 to 625 K) to larger, homogeneously distributed loops at higher temperatures. Their Burgers vectors were a/2<111>. The V-B/sub 4/C specimens contained only tangled dislocation segments. Cavities were observed in all specimens. The cavity concentration decrease and the average diameter increased with increasing irradiation temperature. At 725 K, the maximum swelling was observed in both the vanadium (0.1%) and V-B/sub 4/C (1.4%). At comparable temperatures the cavities in the V-B/sub 4/C specimens were smaller and more numerous than those in the vanadium specimens. Helium bubbles were found on the grain boundaries in all of the V-B/sub 4/ specimens.

HTCAP-1: a program for calcuating operating temperatures in HFIR target irradiation experiments

International Nuclear Information System (INIS)

Kania, M.J.; Howard, A.M.

1980-06-01

The thermal modeling code, HTCAP-1, calculates in-reactor operating temperatures of fueled specimens contained in the High Flux Isotope Reactor (HFIR) target irradiation experiments (HT-series). Temperature calculations are made for loose particle and bonded fuel rod specimens. Maximum particle surface temperatures are calculated for the loose particles and centerline and surface temperatures for the fuel rods. Three computational models are employed to determine fission heat generation rates, capsule heat transfer analysis, and specimen temperatures. This report is also intended to be a users' manual, and the application of HTCAP-1 to the HT-34 irradiation capsule is presented

Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Aguilar H, F.

2015-09-01

The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO 3 ) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10 18 n/cm 2 , which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

ATF Neutron Irradiation Program Technical Plan

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-01

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

LOCA scenario tests of irradiated fuel rod specimens

International Nuclear Information System (INIS)

Scott, Harold

2004-01-01

Full text: The NRC's cladding performance program at Argonne National Laboratory (ANL) is testing fueled high-burnup segments subjected to LOCA integral phenomena. The data are provided to NRC and the nuclear industry for their independent assessment of the adequacy of licensing criteria for LOCA events. The tests are being conducted with high-burnup 30 cm segments from Limerick (9x9 Zry-2) and H.B. Robinson (15x15 Zry-4) reactors. Prior to testing, sibling samples are characterized with respect to fuel morphology, fuel-cladding bond, cladding oxide layer thickness, hydrogen content and high-temperature steam oxidation kinetics. Specimens that survive quench are subjected to four-point bend tests, followed by local diametral compression tests. The retention of post-quench ductility is a more limiting requirement than surviving thermal stresses during quench. Companion tests are conducted with unirradiated cladding to generate baseline data for comparison with the high-burnup fuel results. LOCA integral tests have the following sequential steps: stabilization of temperature, internal pressure and steam flow at 300 C, ramping of temperature (∼5C/s) through ballooning and burst to 1204 C, hold at 1204 C for 1-5 minutes, slow-cooling (∼3C/s) to 800 C, and water quenching at ∼800C. Two high-burnup tests were completed in 2002 with Limerick BWR rod segments: ramp to burst in argon followed by slow cooling; and the LOCA test with 5-minute hold time at 1204 C, followed by slow cooling. With the exception of burst-opening shape, results for burst temperature, burst pressure, burst length, and ballooning strain profile are more similar to, than different from, results for unirradiated Zry-2 cladding exposed to the same time-temperature history. The 3rd Limerick test with quench was performed in December 2003, and a 4th Limerick test was performed in March 2004. Tests on high-burnup Robinson PWR fuel segments are scheduled to begin in June 2004. The presentation points

The relaxation phenomena of radicals induced in irradiated fresh mangoes

International Nuclear Information System (INIS)

Kikuchi, Masahiro; Morishita, Norio; Kobayashi, Yasuhiko; Ogawa, Hideyuki; Shimoyama, Yuhei; Ukai, Mitsuko

2009-01-01

Using the γ-irradiated fresh mangoes followed by freeze-drying and powderization, electron spin resonance spectrometry of specimens was performed. As a result, a strong single peak in the flesh, the pericarp and the seed was observed at g=2.004 and attributed to organic free radicals. When relaxation times of the peak was calculated using the method of Lund et al., T 2 showed dose responses according to increasing doses while T 1 was almost constant. Dose responsibility of the relaxation time T 2 obtained from flesh specimens of the mangoes could be measured regardless of the preservation period of 1 to 9 days following γ-irradiation. Therefore, there might be possible to detect the irradiation treatment of fresh mangoes using relaxation time T 2 . (author)

Instrumented indentation for characterization of irradiated metals at room and high temperatures

International Nuclear Information System (INIS)

Sacksteder, Irene

2011-01-01

The reliability and sustainability of future fusion power plants will highly depend on the aptitude of materials to withstand severe irradiation conditions induced by the burning plasma in reactors. The so-called reduced-activation ferritic-martensitic (RAFM) steels are the current promising candidates for the structural applications considering the reactor's first wall. These steels exhibit irradiation embrittlement and hardening for defined irradiation conditions that are mainly characterized by the irradiation temperature and the irradiation dose. A proper characterization of such irradiated steels implies the use of adapted mechanical testing tools. In the present study, the instrumented indentation technique makes use of a post-processing tool based on neural networks. This technique has been selected for its ability to examine tensile properties by multistage indents on miniaturized irradiated metallic samples. The steel specimens studied in this project have been neutron-irradiated up to a dose of 15 dpa. They have been subsequently tested at room temperature in a Hot Cell by means of an adapted commercial indentation device. The significant irradiation-induced hardening effect present in the range of 250-350 deg C could be observed in the hardness and material's strength parameters. These two material parameters show a similar evolution with increasing irradiation temperatures. Post-irradiation annealing treatments of Eurofer97 have been realized and leads to a partial recovery of the irradiation damage. Considering the demands for characterization in irradiated steels at high temperature and for post-irradiation annealing experiments, the existing instrumented indentation device has been further developed during this work. A conceptual design has been proposed for an indentation testing machine, operating at up to 650 deg C, while remaining the critical temperature limit for tensile strength of the newly developed oxide dispersion strengthening ferritic

Final Report on Design, Fabrication and Test of HANARO Instrumented Capsule (07M-13N) for the Researches of Irradiation Performance of Parts of X-Gen Nuclear Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

2008-08-15

An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of a X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens of control rod materials of AP1000 reactor requested by Westinghouse Co. were inserted in this capsule as a preliminary irradiation test and Polyimide specimens requested by Hanyang university were also inserted. 463 specimens such as buckling and spring test specimens of cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718, Polyimide, Ag and Ag-In-Cd alloys were placed in the capsule. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of neutron fluence monitors installed in the capsule. A new friction welded tube between STS304 and Al1050 alloys was introduced in the capsule to prevent a coolant leakage into a capsule during a capsule cutting process in HANARO. The capsule was irradiated for 95.19 days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 230 {approx} 420 .deg. C. The specimens were irradiated up to a maximum fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1.0MeV) and the dpa of the irradiated specimens were evaluated as 1.21 {approx} 1.97. The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell and the obtained results will be very valuable for the related researches of the users.

Effects of irradiation on the interface between U-Mo and zirconium diffusion barrier

Science.gov (United States)

Jue, Jan-Fong; Keiser, Dennis D.; Miller, Brandon D.; Madden, James W.; Robinson, Adam B.; Rabin, Barry H.

2018-02-01

Irradiated fuel plates were characterized by microscopy that focused on the interface between U-Mo and Zr. Before irradiation, there were three major sub-layers identified in the U-Mo/Zr interface, namely, UZr2, Mo2Zr, and U with low Mo. The typical total thickness of this U-Mo/Zr interaction is 2-3 μm. The UZr2 sub-layer formed during fuel plate fabrication remains stable after irradiation, without large bubbles/porosity accumulation. However, this sub-layer becomes increasingly discontinuous as burnup increases. The low-Mo sub-layer exhibits numerous sub-micron bubbles/porosity at low burnup. Larger, interconnected porosity in this sub-layer was observed in a medium-burnup fuel specimen. However, at higher burnup, regions with the extra-large bubbles/porosity (i.e., larger than 5 μm) were observed in the U-Mo fuel foil at least 5 μm away from the original location of this sub-layer. The mechanism for the formation of the extra-large bubbles/porosity is still unclear at this time. In general, the U-Mo/Zr interface in monolithic U-Mo fuels is relatively stable after irradiation. No large detrimental defects, such as large interfacial bubbles or cracks/delamination, were observed in the fuel plates characterized.

Small specimen technique for assessing mechanical properties of metallic components

Energy Technology Data Exchange (ETDEWEB)

Lobo, Raquel M.; Andrade, Arnaldo H.P.; Morcelli, Aparecido E., E-mail: rmlobo@ipen.br, E-mail: morcelliae@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-11-01

Small Punch Test (SPT) is one of the most promising techniques of small specimen test, which was originally applied in testing of irradiated materials in nuclear engineering. Then it was introduced to other fields as an almost nondestructive method to measure the local mechanical properties that are difficult to be obtained using conventional mechanical tests. Most studies to date are focused on metallic materials, although SPT applications are recently spreading to other materials. The small punch test (SPT) employs small-sized specimens (for example, samples measuring 8 mm in diameter and 0.5 mm thick). The specimen is firmly clamped between two circular dies and is bi-axially strained until failure into a circular hole using a hemispherical punch. The 'load-punch displacement' record can be used to estimate the yield strength, the ultimate tensile strength, the tensile elongation, and the temperature of the ductile-to-brittle transition. Recently, some researchers are working on the use of miniature notched or pre-cracked specimens (denoted as p-SPT) to validate its geometry and dimensions for obtaining the fracture properties of metallic materials. In a first approach, the technique makes it possible to convert primary experimental data into conventional mechanical properties of a massive specimen. In this paper a comprehensive review of the different STP applications is presented with the aim of clarifying its usefulness. (author)

Small specimen technique for assessing mechanical properties of metallic components

International Nuclear Information System (INIS)

Lobo, Raquel M.; Andrade, Arnaldo H.P.; Morcelli, Aparecido E.

2017-01-01

Small Punch Test (SPT) is one of the most promising techniques of small specimen test, which was originally applied in testing of irradiated materials in nuclear engineering. Then it was introduced to other fields as an almost nondestructive method to measure the local mechanical properties that are difficult to be obtained using conventional mechanical tests. Most studies to date are focused on metallic materials, although SPT applications are recently spreading to other materials. The small punch test (SPT) employs small-sized specimens (for example, samples measuring 8 mm in diameter and 0.5 mm thick). The specimen is firmly clamped between two circular dies and is bi-axially strained until failure into a circular hole using a hemispherical punch. The 'load-punch displacement' record can be used to estimate the yield strength, the ultimate tensile strength, the tensile elongation, and the temperature of the ductile-to-brittle transition. Recently, some researchers are working on the use of miniature notched or pre-cracked specimens (denoted as p-SPT) to validate its geometry and dimensions for obtaining the fracture properties of metallic materials. In a first approach, the technique makes it possible to convert primary experimental data into conventional mechanical properties of a massive specimen. In this paper a comprehensive review of the different STP applications is presented with the aim of clarifying its usefulness. (author)

Solute segregation and void formation in ion-irradiated vanadium-base alloys

International Nuclear Information System (INIS)

Loomis, B.A.; Smith, D.L.

1985-01-01

The radiation-induced segregation of solute atoms in the V-15Cr-5Ti alloys was determined after either single- dual-, or helium implantation followed by single-ion irradiation at 725 0 C to radiation damage levels ranging from 103 to 169 dpa. Also, the effect of irradiation temperature (600-750 0 C) on the microstructure in the V-15Cr-5Ti alloy was determined after single-ion irradiation to 200 and 300 dpa. The solute segregation results for the single- and dual-ion irradiated alloy showed that the simultaneous production of irradiation damage and deposition of helium resulted in enhanced depletion of Cr solute and enrichment of Ti, C and S solute in the near-surface layers of irradiated specimens. The observations of the irradiation-damaged microstructures in V-15Cr-5Ti specimens showed an absence of voids for irradiations of the alloy at 600-750 0 C to 200 dpa and at 725 0 C to 300 dpa. The principle effect on the microstructure of these irradiations was to induce the formation of a high density of disc-like precipitates in the vicinity of grain boundaries and intrinsic precipitates and on the dislocation structure. 8 references, 4 figures

Investigation of TIG welding characteristics with a dual cooled rod for the fuel irradiation test

International Nuclear Information System (INIS)

Kim, Soo Sung; Kim, Hyung Kyu

2008-01-01

To establish the fabrication process, and for satisfying the requirements of the irradiation test, an TIG(Tungsten Inert Gas) welding machine for the dual cooled rods specimens was developed, and the preliminary welding experiments were performed to optimize the welding process conditions. Cladding tubes of 15.9 and 9 mm for the outer and inner diameters, respectively with a 0.57 mm thickness and end caps were used for the specimens. This paper describes the experimental results of the TIG welds and the micrograph examinations of the TIG welded specimens corresponding to various welding conditions for the dual cooled fuel irradiation test. The investigations revealed that the present TIG process satisfied the requirements for the fuel irradiation test in the HANARO research reactor

Fracture toughness of irradiated beryllium

International Nuclear Information System (INIS)

Beeston, J.M.

1978-01-01

The fracture toughness of nuclear grade hot-pressed beryllium upon irradiation to fluences of 3.5 to 5.0 x 10 21 n/cm 2 , E greater than 1 MeV, was determined. Procedures and data relating to a round-robin test contributing to a standard ASTM method for unirradiated beryllium are discussed in connection with the testing of irradiated specimens. A porous grade of beryllium was also irradiated and tested, thereby enabling some discrimination between the models for describing the fracture toughness behavior of porous beryllium. The fracture toughness of unirradiated 2 percent BeO nuclear grade beryllium was 12.0 MPa m/sup 1 / 2 /, which was reduced 60 percent upon irradiation at 339 K and testing at 295 K. The fracture toughness of a porous grade of beryllium was 13.1 MPa m/sup 1 / 2 /, which was reduced 68 percent upon irradiation and testing at the same conditions. Reasons for the reduction in fracture toughness upon irradiation are discussed

Detection of irradiated spice in blend of irradiated and un-irradiated spices using thermoluminescence method

International Nuclear Information System (INIS)

Goto, Michiko; Yamazaki, Masao; Sekiguchi, Masayuki; Todoriki, Setsuko; Miyahara, Makoto

2007-01-01

Five blended spice sample were prepared by mixing irradiated and un-irradiated black pepper and paprika at different ratios. Blended black pepper containing 2%(w/w) of 5.4 kGy-irradiated black pepper showed no maximum at glow1. Irradiated black pepper samples, mixed to 5 or 10%(w/w), were identified as 'irradiated' or 'partially irradiated' or 'un-irradiated'. All samples with un-irradiated pepper up to 20%(w/w) were identified as irradiated'. In the case 5.0 kGy-irradiated paprika were mixed with un-irradiated paprika up to 5%(w/w), all samples were identified as irradiated'. The glow1 curves of samples, including irradiated paprika at 0.2%(w/w) or higher, exhibited a maximum between 150 and 250degC. The results suggest the existence of different critical mixing ratio for the detection of irradiation among each spices. Temperature range for integration of the TL glow intensity were compared between 70-400degC and approximate 150-250degC, and revealed that the latter temperature range was determined based on the measurement of TLD100. Although TL glow ratio in 150-250degC was lower than that of 70-400degC range, identification of irradiation was not affected. Treatment of un-irradiated black pepper and paprika with ultraviolet rays had no effect on the detection of irradiation. (author)

New irradiation devices at the FRN reactor

International Nuclear Information System (INIS)

Stark, W.

1980-01-01

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

Demonstration of finite element simulations in MOOSE using crystallographic models of irradiation hardening and plastic deformation

Energy Technology Data Exchange (ETDEWEB)

Patra, Anirban [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wen, Wei [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez Saez, Enrique [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tome, Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-05-31

This report describes the implementation of a crystal plasticity framework (VPSC) for irradiation hardening and plastic deformation in the finite element code, MOOSE. Constitutive models for irradiation hardening and the crystal plasticity framework are described in a previous report [1]. Here we describe these models briefly and then describe an algorithm for interfacing VPSC with finite elements. Example applications of tensile deformation of a dog bone specimen and a 3D pre-irradiated bar specimen performed using MOOSE are demonstrated.

Microstructural evolution in austenitic stainless steel irradiated with triple-beam

Energy Technology Data Exchange (ETDEWEB)

Hamada, Shozo; Miwa, Yukio; Yamaki, Daiju [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yichuan, Zhang

1997-03-01

An austenitic stainless steel was simultaneously irradiated with nickel, helium and hydrogen ions at the temperature range of 573-673 K. The damage level and injected concentration of He and H ions in the triple-beam irradiated region are 57 dpa, 19000 and 18000 at.ppm, respectively. Following to irradiation, the cross sectional observation normal to the incident surface of the specimen was carried out with a transmission electron microscope. Two bands parallel to the incident surface were observed in the irradiated specimen, which consist of dislocation loops and lines of high number density. These locate in the range of the depth of 0.4 to 1.3 {mu}m and 1.8 to 2.4 {mu}m from the incident surface, respectively. The region between two bands, which corresponds to the triple beam irradiated region, shows very low number density of dislocations than that in each band. Observation with higher magnification of this region shows that fine cavities with high number density uniformly distribute in the matrix. (author)

The effect of irradiation and irradiation temperature on the fracture toughness of cold-worked Zr-2.5 wt percent Nb

International Nuclear Information System (INIS)

Simpson, L.A.; Ellis, R.B.; Stark, D.J.; Shillinglaw, A.J.

1984-09-01

The use of fracture mechanics methods and small specimens to assess the effect of metallurgical variables on fracture toughness and critical crack length in reactor pressure tubes is reviewed. Fracture toughness tests on specimens irradiated in the NRU research reactor at 260 degrees C are described and compared with results from a previous irradiation in the WR-1 research reactor at 350 degrees C. The J-resistance curve is used as the measure of fracture toughness, and is shown to be very sensitive to the metallurgical state. The lower irradiation temperature (260 degrees C), characteristic of the operating temperature range for power reactors, has a significant effect on fracture toughness. Circumferential hydrides also have an effect. Estimates of critical crack length are made using the J-resistance data, and are seen to slightly underestimate the actual critical crack length as determined in full-scale burst tests. This conservatism is not large enough to impose a significant penalty in design applications

Weldability of neutron-irradiated stainless steel and nickel-base alloy

International Nuclear Information System (INIS)

Koyabu, Ken; Asano, Kyoichi; Takahashi, Hidenori; Sakamoto, Hiroshi; Kawano, Shohei; Nakamura, Tomomi; Hashimoto, Tsuneyuki; Koshiishi, Masato; Kato, Takahiko; Katsura, Ryoei; Nishimura, Seiji

2000-01-01

Degradation of of weldability caused by helium, which is generated by nuclear transmutation irradiated material, is an important issue to be addressed in planning of proactive maintenance of light water reactor core internal components. In this work, the weldability of neutron.irradiated stainless steel and nickel-base alloy, which are major constituting materials for components, was practically evaluated. The weldability was first examined by TIG welding in relation to the weld heat input and helium content using various specimens (made of SUS304 and SUS316L) sampled from reactor internal components. The specimens were neutron irradiated in a boiling water reactor to fluences from 4 x 10 24 to 1.4 x 10 26 n/ m 2 (E> l MeV ), and resulting helium generation ranged from 0.1 to 103 appm. The weld defects were characterized by dye penetrant test and cross-sectional metallography. The weldability of neutron-irradiated stainless steel was shown to be better at lower weld heat input and lower helium content. To evaluate mechanical properties of welded joints, thick plates (20 mm) specimens of SUS304 and Alloy 600 were prepared and irradiated in Japan Material Test Reactor (JMTR). The helium content of the specimens was controlled to range from 0.11 to 1.34 appm selected to determine threshold helium content to weld successfully. The welded joints had multiple passes by TIG welding process at 10 and 20 kJ/cm heat input. The welded joints of thick plate were characterized by dye penetrant test, cross-sectional metallography, tensile test, side bend test and root bend test. It was shown that irradiated stainless steel containing below 0.14 appm of helium could be welded with conventional TIG welding process (heat input below 20 kJ/cm). Nickel-base alloy, which contained as much helium as stainless steel could be welded successfully, could also be welded with conventional TIG welding process, These results served as basis to evaluate the applicability of repair welding to

Proton irradiation effects on organic polymers

International Nuclear Information System (INIS)

Seguchi, T.; Sasuga, T.; Kawakami, W.; Hagiwara, M.; Kohno, I.; Kamitsubo, H.

1987-01-01

Organic polymer films(100 μm thickness) of polyethylene, polypropylene, polyethyleneterephtalate, and polyethersulfone were irradiated by protons of 8 MeV using a cyclotron, and their radiation effects were investigated by the changes of mechanical properties. In order to irradiate protons uniformly over wide area of polymer films, specimens were scanned during proton irradiation using a special apparatus. The absorbed dose was measured by CTA and RCD film dosimeters, and can be determined that 1 μC/cm 2 of 8 MeV proton fluence is equivalent to 54 kGy. For polyethylene and polypropylene, there was no significant difference between proton and electron irradiation for same doses. However, for polyethersulfone the decay of mechanical property was observed to be less than that of irradiation by electron. (author)

SEM-analysis of grain boundary porosity in three S-176 specimens

International Nuclear Information System (INIS)

Malen, K.; Birath, S.; Mattsson, O.

1980-10-01

Porosity in UO 2 -fuel has been studied in scanning electron microscope (SEM). The aim was to obtain a basis for evaluation of porosity in high burnup power reactor fuel. Three specimens have been analyzed. In the high temperature zones porosity can be seen both on grain boundaries and at grain edges. In the low temperature regions very little changes seem to have occurred during irradiation. (author)

JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

International Nuclear Information System (INIS)

G. Borges

2006-01-01

The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long term microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus

JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

Energy Technology Data Exchange (ETDEWEB)

G. Borges

2006-01-31

The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long term microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.

Effect of helium irradiation on fracture modes

International Nuclear Information System (INIS)

Hanamura, T.; Jesser, W.A.

1982-01-01

The objective of this work is to determine the crack opening mode during in-situ HVEM tensile testing and how it is influenced by test temperature and helium irradiation. Most cracks were mixed mode I and II. However, between 250 0 C and room temperature the effect of helium irradiation is to increase the amount of mode I crack propagation. Mode II crack opening was observed as grain boundary sliding initiated by a predominantly mode I crack steeply intersecting the grain boundary. Mode II crack opening was absent in irradiated specimens tested between 250 0 C and room temperature, but could be restored by a post irradiation anneal

Fabrication of three-dimensional platinum microstructures with laser irradiation and electrochemical technique

International Nuclear Information System (INIS)

Kikuchi, T.; Takahashi, H.; Maruko, T.

2007-01-01

Three-dimensional (3D) platinum microstructures were fabricated by successive procedures: aluminum anodizing, laser irradiation, nickel/platinum electroplating, and removal of the aluminum substrate, the oxide films, and the nickel metal layer. Aluminum plates and rods were anodized in an oxalic acid solution to form porous type oxide films. The anodized specimens were immersed in a nickel electroplating solution, and then irradiated with a pulsed Nd-yttrium aluminum garnet (YAG) laser beam to remove the anodic oxide film with a three-dimensional XYZθ stage. The specimens were cathodically polarized in the nickel and a platinum electroplating solution to form the metal micropattern at the laser-irradiated area. The electroplated specimens were immersed in NaOH solution to dissolve the aluminum substrate and the oxide films, and then immersed in HCl solution to dissolve the nickel deposits. A platinum grid-shaped microstructure, a microspring, and a cylindrical network microstructure with 50-100 μm line width were obtained successfully

Fabrication of three-dimensional platinum microstructures with laser irradiation and electrochemical technique

Energy Technology Data Exchange (ETDEWEB)

Kikuchi, T. [Graduate School of Engineering, Hokkaido University, N13, W8, Kita-Ku, Sapporo (Japan)]. E-mail: kiku@elechem1-mc.eng.hokudai.ac.jp; Takahashi, H. [Graduate School of Engineering, Hokkaido University, N13, W8, Kita-Ku, Sapporo (Japan); Maruko, T. [Furuya Metal Co. Ltd., R and D Group, Shimodate Daiichi Kogyodanchi 1915, Morisoejima, Chikusei, Ibaraki (Japan)

2007-02-01

Three-dimensional (3D) platinum microstructures were fabricated by successive procedures: aluminum anodizing, laser irradiation, nickel/platinum electroplating, and removal of the aluminum substrate, the oxide films, and the nickel metal layer. Aluminum plates and rods were anodized in an oxalic acid solution to form porous type oxide films. The anodized specimens were immersed in a nickel electroplating solution, and then irradiated with a pulsed Nd-yttrium aluminum garnet (YAG) laser beam to remove the anodic oxide film with a three-dimensional XYZ{theta} stage. The specimens were cathodically polarized in the nickel and a platinum electroplating solution to form the metal micropattern at the laser-irradiated area. The electroplated specimens were immersed in NaOH solution to dissolve the aluminum substrate and the oxide films, and then immersed in HCl solution to dissolve the nickel deposits. A platinum grid-shaped microstructure, a microspring, and a cylindrical network microstructure with 50-100 {mu}m line width were obtained successfully.

Irradiation Embrittlement Monitoring Programs of RPV's in the Slovak Republic NPP's

International Nuclear Information System (INIS)

Kupca, Ludovik

2006-01-01

Four types of surveillance programs were (are) realized in Slovak NPP's: 'Standard Surveillance Specimen Program' (SSSP) was finished in Jaslovske Bohunice V-2 Nuclear Power Plant (NPP) Units 3 and 4, 'Extended Surveillance Specimen Program' (ESSP), was prepared for Jaslovske Bohunice NPP V-2 with aim to validate the SSSP results, For the Mochovce NPP Unit 1 and 2 was prepared completely new surveillance program 'Modern Surveillance Specimen Program' (MSSP), based on the philosophy that the results of MSSP must be available during all NPP service life, For the Bohunice V-1 NPP was finished 'New Surveillance Specimen Program' (NSSP) coordinated by IAEA, which gave arguments for prolongation of service life these units for minimum 20 years, New Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1, 2) is approved now. ASSP is dealing with the irradiation embrittlement of heat affected zone (HAZ) and RPV's austenitic cladding, which were not evaluated till this time in surveillance programs. SSSP started in 1979 and was finished in 1990. ESSP program started in 1995 and will be finished in 2007, was prepared with aim of: increasing of neutron fluence measurement accuracy, substantial improvement the irradiation temperature measurement, fixed orientation of samples to the centre of the reactor core, minimum differences of neutron dose for all the Charpy-V notch and COD specimens, the dose rate effect evaluation. In the year 1996 was started the new surveillance specimen program for the Mochovce RPV's unit-1 and 2, based on the fundamental postulate - to provide the irradiation embrittlement monitoring till the end of units operation. The 'New Surveillance Specimen Program' (NSSP) prepared in the year 1999 for the Bohunice V-1 NPP was finished in the year 2004. Main goal of this program was to evaluate the weld material properties degradation due to the irradiation and recovery efficiency by annealing too. The

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Sub ablative Er: YAG laser irradiation on surface roughness of eroded dental enamel.

Science.gov (United States)

Curylofo-Zotti, Fabiana Almeida; Lepri, Taísa Penazzo; Colucci, Vivian; Turssi, Cecília Pedroso; Corona, Silmara Aparecida Milori

2015-11-01

This study evaluated the effects of Er:YAG laser irradiation applied at varying pulse repetition rate on the surface roughness of eroded enamel. Bovine enamel slabs (n = 10) were embedded in polyester resin, ground, and polished. To erosive challenges, specimens were immersed two times per day in 20mL of concentrated orange juice (pH = 3.84) under agitation, during a two-day period. Specimens were randomly assigned to irradiation with the Er:YAG laser (focused mode, pulse energy of 60 mJ and energy density of 3.79 J/cm(2) ) operating at 1, 2, 3, or 4 Hz. The control group was left nonirradiated. Surface roughness measurements were recorded post erosion-like formation and further erosive episodes by a profilometer and observed through atomic force microscopy (AFM). Analysis of variance revealed that the control group showed the lowest surface roughness, while laser-irradiated substrates did not differ from each other following post erosion-like lesion formation. According to analysis of covariance, at further erosive episodes, the control group demonstrated lower surface roughness (P > 0.05), than any of the irradiated groups (P enamel eroded. The AFM images showed that the specimens irradiated by the Er:YAG laser at 1 Hz presented a less rough surface than those irradiated at 2, 3, and 4 Hz. © 2015 Wiley Periodicals, Inc.

X-ray irradiation and Rho-kinase inhibitor additively induce invasiveness of the cells of the pancreatic cancer line, MIAPaCa-2, which exhibits mesenchymal and amoeboid motility

International Nuclear Information System (INIS)

Fujita, Mayumi; Otsuka, Yoshimi; Yamada, Shigeru; Iwakawa, Mayumi; Imai, Takashi

2011-01-01

Tumor cells can migrate and invade tissue by two modes of motility: mesenchymal and amoeboid. X-ray or γ-ray irradiation increases the invasiveness of tumor cells with mesenchymal motility through the induction of matrix metalloproteinases (MMP), and this increase is suppressed by MMP inhibitors (MMPI). However, the effects of X-ray or γ-ray irradiation on the invasiveness of tumor cells with amoeboid motility remain unclear. We investigated the effect of irradiation on amoeboid motility by using cells of the human pancreatic cancer line, MIAPaCa-2, which exhibits both modes of motility. The X-ray-induced invasiveness of MIAPaCa-2 cells was associated with the upregulation of MMP2 at both the RNA and protein levels and was inhibited by MMPI treatment. Amoeboid-mesenchymal transition was slightly induced after irradiation. The MMPI treatment caused mesenchymal-amoeboid transition without significant increase in invasiveness, while the ROCK inhibitor (ROCKI) stimulated amoeboid-mesenchymal transition and enhanced invasiveness under both non-irradiated and irradiated conditions. This ROCKI-induced transition was accompanied by the upregulation of MMP2 mRNA and protein. Exposure to both irradiation and ROCKI further enhanced MMP2 expression and had an additive effect on the invasiveness of MIAPaCa-2 cells. Additionally, exposure to MMPI led to significant suppression of both radiation-induced and the basal invasiveness of MIAPaCa-2 cells. This suggests that ROCKI treatment, especially with concomitant X-ray irradiation, can induce invasion of cancer cells and should be used only for certain types of cancer cells. Simultaneous use of inhibitors, ROCKI and MMPI may be effective in suppressing invasiveness under both X-ray-irradiated and non-irradiated conditions. (author)

Development of high time-resolution laser flash equipment for thermal diffusivity measurements using miniature-size specimens

International Nuclear Information System (INIS)

Shikama, Tatsuo; Namba, Chusei; Kosuda, Michinori; Maeda, Yukio.

1994-01-01

For measurements of thermal diffusivity of miniature-size specimens heavily irradiated by neutrons, a new Q-switched laser-flash instrument was developed. In the present instrument the time-resolution was improved to 0.1 ms by using a laser-pulse width of 25 ns. The realization of high time-resolution made it possible to measure the thermal diffusivity of thin specimens. It is expected that copper of 0.7 mm thick, and SUS 304 of 0.1 mm could be used for the measurements. In case of ATJ graphite, 0.5 mm thick specimen could be used for the reliable measurement in the temperature range of 300-1300 K. (author)

Effects of irradiation on low cycle fatigue properties for reduced activation ferritic/martensitic steel

International Nuclear Information System (INIS)

Kim, S.W.; Tanigawa, H.; Hirose, T.; Kohyama, A.

2007-01-01

Full text of publication follows: In materials life decision for a commercial blanket, thermal fatigue property of materials is a particularly important. The loading of structural materials in fusion reactor is, besides the plasma surface interactions, a combined effect of high heat fluxes and neutron irradiation. Depending on the pulse lengths, the operating conditions, and the thermal conductivity, these oscillating temperature gradients will cause elastic and elastic-plastic cyclic deformation giving rise to (creep-) fatigue in structural first wall and blanket components. Especially, investigation of the fatigue property in Reduced Activation Ferritic/Martensitic (RAF/M) steel and establishment of the evaluation technology are demanded in particular immediately for design/manufacturing of ITER-TBM. And also, fatigue testing after irradiation will be carried out in hot cells with remote control system. Considering limited ability of specimen manipulation in the cells, the specimen and the test method need to be simple for operation. The existing data bases of RAF/M steel provide baseline data set including post-irradiation fatigue data. However, to perform the accurate fatigue lifetime assessment for ITER-TBM and beyond utilizing the existing data base, the mechanical understanding of fatigue fracture is mandatory. It has been previously reported by co-authors that dislocation cell structure was developed on low cycle fatigued RAF/M steel, and led the fatigue crack to develop along prior austenitic grain boundary. In this work, the effects of nuclear irradiation on low cycle fatigue properties for RAF/M steels and its fracture mechanisms were examined based on the flow stress analysis and detailed microstructure analysis. Fracture surfaces and crack initiation site were investigated by scanning electron microscope (SEM). Transmission electron microscopy (TEM) was also applied to clarify the microstructural features of fatigue behavior. It is also important to

Effects of irradiation on low cycle fatigue properties for reduced activation ferritic/martensitic steel

Energy Technology Data Exchange (ETDEWEB)

Kim, S.W. [Kyoto Univ., Graduate School of Energy Science (Japan); Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan)

2007-07-01

Full text of publication follows: In materials life decision for a commercial blanket, thermal fatigue property of materials is a particularly important. The loading of structural materials in fusion reactor is, besides the plasma surface interactions, a combined effect of high heat fluxes and neutron irradiation. Depending on the pulse lengths, the operating conditions, and the thermal conductivity, these oscillating temperature gradients will cause elastic and elastic-plastic cyclic deformation giving rise to (creep-) fatigue in structural first wall and blanket components. Especially, investigation of the fatigue property in Reduced Activation Ferritic/Martensitic (RAF/M) steel and establishment of the evaluation technology are demanded in particular immediately for design/manufacturing of ITER-TBM. And also, fatigue testing after irradiation will be carried out in hot cells with remote control system. Considering limited ability of specimen manipulation in the cells, the specimen and the test method need to be simple for operation. The existing data bases of RAF/M steel provide baseline data set including post-irradiation fatigue data. However, to perform the accurate fatigue lifetime assessment for ITER-TBM and beyond utilizing the existing data base, the mechanical understanding of fatigue fracture is mandatory. It has been previously reported by co-authors that dislocation cell structure was developed on low cycle fatigued RAF/M steel, and led the fatigue crack to develop along prior austenitic grain boundary. In this work, the effects of nuclear irradiation on low cycle fatigue properties for RAF/M steels and its fracture mechanisms were examined based on the flow stress analysis and detailed microstructure analysis. Fracture surfaces and crack initiation site were investigated by scanning electron microscope (SEM). Transmission electron microscopy (TEM) was also applied to clarify the microstructural features of fatigue behavior. It is also important to

The modelling of irradiation embrittlement in submerged-arc welds

International Nuclear Information System (INIS)

Bolton, C.J.; Buswell, J.T.; Jones, R.B.; Moskovic, R.; Priest, R.H.

1996-01-01

Until very recently, the irradiation embrittlement behavior of submerged-arc welds has been interpreted in terms of two mechanisms, namely a matrix damage component and an additional component due to the irradiation-enhanced production of copper-rich precipitates. However, some of the weld specimens from a recent accelerated re-irradiation experiment have shown high Charpy shifts which exceeded the values expected from the measured shift in yield stress. Microstructural examination has revealed the occurrence of intergranular fracture (IGF) in these specimens, accompanied by grain boundary segregation of phosphorus. Theoretical models were developed to predict the parametric dependence of irradiation-enhanced phosphorus segregation on experimental variables. Using these parametric forms, along with the concept of a critical level of segregation for the onset of IGF instead of cleavage, a three mechanism trend curve has been developed. The form of this trend curve, taking into account IGF as well as matrix and copper embrittlement, is thus mechanistically based. The constants in the equation, however, are obtained by a statistical fit to the actual Charpy shift database

X-ray irradiation of RC-MAP pre-stored for various numbers of days. Effect of X-ray irradiation on RC-MAP and reuse of RC-MAP after irradiation

International Nuclear Information System (INIS)

Yamada, Naotomo; Nagumo, Fumio; Kawasaki, Seiji; Matsuzaki, Miwako; Tadano, Jutaro

1995-01-01

X-ray irradiation is currently in wide use as a means of preventing post-transfusion graft-versus-host disease (PT-GVHD). In this study, we evaluated the effects of X-ray irradiation on RC-MAP pre-stored for various numbers of days, and assessed how long irradiated RC-MAP can be stored. RC-MAP was irradiated at a dosage of 15 Gy at 1, 7, 14, 21 or 28 days after blood collection. These specimens were referred to as group I, II, III, IV and V (X-ray-irradiated groups), respectively. Non-irradiated RC-MAP was used as the control. Results showed that plasma K concentration increased after X-ray irradiation. It is therefore advisable that RC-MAPs be used immediately in infants and in renal failure. However, to maximize the efficiency of blood product use, it seemed possible that groups I and II could be used within two weeks after irradiation, and group III within one week. On the basis of this increase in K concentration, however, groups IV and V plasma should be used immediately after irradiation. (author)

Properties of vanadium-base alloys irradiated in the Dynamic Helium Charging Experiment*1

Science.gov (United States)

Chung, H. M.; Loomis, B. A.; Smith, D. L.

1996-10-01

One property of vanadium-base alloys that is not well understood in terms of their potential use a fusion reactor structural materials, is the effect of simultaneous generation of helium and neutron damage. In the present Dynamic Helium Charging Experiment (DHCE), helium was produced uniformly in the specimen at linear rates of ≈ 0.4 to 4.2 appm helium/dpa by the decay of tritium during irradiation to 18-31 dpa at 425-600°C in Li-filled capsules in a sodium-cooled fast reactor. This paper presents results of postirradiation examination and tests of microstructure and mechanical properties of V5Ti, V3Ti1Si, V8Cr6Ti, and V4Cr4Ti (the latter alloy has been identified as the most promising candidate vanadium alloy). Effects of helium on tensile strength and ductility were insignificant after irradiation and testing at > 420°C. However, postirradiation ductilities at irradiation. Ductile—brittle transition behavior of the DHCE specimens was also determined from bend tests and fracture appearance of transmission electron microscopy (TEM) disks and broken tensile specimens. No brittle behavior was observed at temperatures > - 150°C in DHCE specimens. Predominantly brittle-cleavage fracture morphologies were observed only at - 196°C in some specimens that were irradiated to 31 dpa at 425°C during the DHCE. For the helium generation rates in this experiment (≈ 0.4-4.2 appm He/dpa), grain-boundary coalescence of helium microcavities was negligible and intergranular fracture was not observed.

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-09-15

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

Programmed temperature control of capsule in irradiation test with personal computer at JMTR

International Nuclear Information System (INIS)

Saito, H.; Uramoto, T.; Fukushima, M.; Obata, M.; Suzuki, S.; Nakazaki, C.; Tanaka, I.

1992-01-01

The capsule irradiation facility is one of various equipments employed at the Japan Materials Testing Reactor (JMTR). The capsule facility has been used in irradiation tests of both nuclear fuels and materials. The capsule to be irradiated consists of the specimen, the outer tube and inner tube with a annular space between them. The temperature of the specimen is controlled by varying the degree of pressure (below the atmospheric pressure) of He gas in the annular space (vacuum-controlled). Beside this, in another system the temperature of the specimen is controlled with electric heaters mounted around the specimen (heater-controlled). The use of personal computer in the capsule facility has led to the development of a versatile temperature control system at the JMTR. Features of this newly-developed temperature control system lie in the following: the temperature control mode for a operation period can be preset prior to the operation; and the vacuum-controlled irradiation facility can be used in cooperation with the heater-controlled. The introduction of personal computer has brought in automatic heat-up and cool-down operations of the capsule, setting aside the hand-operated jobs which had been conducted by the operators. As a result of this, the various requirements seeking a higher accuracy and efficiency in the irradiation can be met by fully exploiting the capabilities incorporated into the facility which allow the cyclic or delicate changes in the temperature. This paper deals with a capsule temperature control system with personal computer. (author)

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)

Dislocation and void segregation in copper during neutron irradiation

DEFF Research Database (Denmark)

Singh, Bachu Narain; Leffers, Torben; Horsewell, Andy

1986-01-01

); the irradiation experiments were carried out at 250 degree C. The irradiated specimens were examined by transmission electron microscopy. At both doses, the irradiation-induced structure was found to be highly segregated; the dislocation loops and segments were present in the form of irregular walls and the voids...... density, the void swelling rate was very high (approximately 2. 5% per dpa). The implications of the segregated distribution of sinks for void formation and growth are briefly discussed....

Iodine stress-corrosion cracking in irradiated Zircaloy cladding

International Nuclear Information System (INIS)

Mattas, R.F.; Yaggee, F.L.; Neimark, L.A.

1979-01-01

Irradiated Zircaloy cladding specimens, which had experienced fluences from 0.1 to 6 x 10 21 n/cm 2 (E>0.1 MeV), were gas-pressure tested in an iodine environment to investigate their stress-corrosion cracking (SCC) susceptibility. The test temperatures and hoop stresses ranged from 320 to 360 0 C and 150 to 500 MPa, respectively. The results indicate that irradiation, in general, increases the susceptibility of Zircaloy to iodine SCC. For specimens that experienced fluences >2 x 10 21 n/cm 2 (E>0.1 MeV), the 24-h failure stress was 177+-18 MPa, regardless of the preirradiation metallurgical condition. An analytical model for iodine SCC has been developed which agrees reasonably well with the test results

Mechanical properties of irradiated beryllium

International Nuclear Information System (INIS)

Beeston, J.M.; Longhurst, G.R.; Wallace, R.S.

1992-01-01

Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 x 10 25 n/m 2 (E > MeV) at an irradiation temperature of 75deg C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium. (orig.)

Mechanical properties of irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Beeston, J.M.; Longhurst, G.R.; Wallace, R.S. (EG and G Idaho, Inc., Idaho Falls, ID (United States). Idaho National Engineering Lab.); Abeln, S.P. (EG and G Rocky Flats, Inc., Golden, CO (United States))

1992-10-01

Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 x 10[sup 25] n/m[sup 2] (E > MeV) at an irradiation temperature of 75deg C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium. (orig.).

Mechanical properties of irradiated beryllium

Science.gov (United States)

Beeston, J. M.; Longhurst, G. R.; Wallace, R. S.; Abeln, S. P.

1992-10-01

Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 × 10 25 n/m 2 ( E > 1 MeV) at an irradiation temperature of 75°C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium.

Irradiation effects on aluminium and beryllium

International Nuclear Information System (INIS)

Bieth, M.

1992-01-01

The High Flux Reactor (HFR) in Petten (The Netherlands) is a 45 MW light water cooled and moderated research reactor. The vessel was replaced in 1984 after more than 20 years of operation because doubts had arisen over the condition of the aluminium alloy construction material. Data on the mechanical properties of the aluminium alloy Al 5154 with and without neutron irradiation are necessary for the safety analysis of the new HFR vessel which is constructed from the same material as the old vessel. Fatigue, fracture mechanics (crack growth and fracture toughness) and tensile properties have been obtained from several experimental testing programmes with materials of the new and the old HFR vessel. 1) Low-cycle fatigue testing has been carried out on non-irradiated specimens from stock material of the new HFR vessel. The number of cycles to failure ranges from 90 to more than 50,000 for applied strain from 3.0% to 0.4%; 2) Fatigue crack growth rate testing has been conducted: - with unirradiated specimens from stock material of the new vessel; - with irradiated specimens from the remnants of the old core box. Irradiation has a minor effect on the sub-critical fatigue crack growth rate. The ultimate increase of the mean crack growth rate amounts to a factor of 2. However crack extension is strongly reduced due to the smaller crack length for crack growth instability (reduction of K IC ). - Irradiated material from the core box walls of the old vessel has been used for fracture toughness testing. The conditional fracture toughness values K IQ ranges from 30.3 down to 16.5 MPa√m. The lowermost meaningful 'K IC ' is 17.7 MPa√m corresponding to the thermal fluence of 7.5 10 26 n/m 2 for the End of Life (EOL) of the old vessel. - Testing carried out on irradiated material from the remnants of the old HFR core box shows an ultimate neutron irradiation hardening of 35 points increase of HSR 15N and an ultimate tensile yield stress of 589 MPa corresponding to the

Some physical properties of irradiated and non-irradiated oxide glasses containing uranium

International Nuclear Information System (INIS)

Simon, V.; Ardelean, I.; Simon, S.; Cozar, O.; Milea, I.; Lupsa, I.; Mih, V.

1995-01-01

The x U O 3 (1-x) [2 P 2 O 5 · Na 2 O] non-irradiated and gamma irradiated glasses (0 3+ , U 4+ and U 5+ ions. The gamma irradiation induces paramagnetic defects around the glass network forming sites occupied by phosphorous atoms. The non-irradiated samples are weak paramagnetic up to x = 0.1. For higher U O 3 concentration (0.1 < x ≤ 0.2) the magnetic measurements indicated a larger number of paramagnetic ions which are magnetically isolated and exhibit a Curie type behaviour. (author) 5 figs., 14 refs

Irradiation creep induced anisotropy in a/2 dislocation populations

International Nuclear Information System (INIS)

Gelles, D.S.

1984-05-01

The contribution of anisotropy in Burgers vector distribution to irradiation creep behavior has been largely ignored in irradiation creep models. However, findings on Frank loops suggest that it may be very important. Procedures are defined to identify the orientations of a/2 Burgers vectors for dislocations in face-centered cubic crystals. By means of these procedures the anisotropy in Burgers vector populations was determined for three Nimonic PE16 pressurized tube specimens irradiated under stress. Considerable anisotropy in Burgers vector population develops during irradiation creep. It is inferred that dislocation motion during irradiation creep is restricted primarily to a climb of a/2 dislocations on 100 planes. Effect of these results on irradiation creep modeling and deformation induced irradiation growth is considered

Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

International Nuclear Information System (INIS)

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

1977-01-01

A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this obervation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. Measurements of radiation enhanced diffusion are less time consuming and expensive than irradiation creep tests and information on this property can be obtained rather quickly, prior to the selection of stainless steel alloys for creep tests. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. Finally, a few uniaxial tensile creep tests will be performed in fully instrumented rigs. (Auth.)

Tritium release from neutron-irradiated Li2O: Transport in porous sintered pellets

International Nuclear Information System (INIS)

Tanifuji, Takaaki; Yamaki, Daiju; Jitsukawa, Shiro

2006-01-01

The tritium release behavior from Li 2 O sintered pellets (81-88% T.D.) is examined by isothermal heating tests. (1) For the 88% T.D. specimens, the fraction of residual tritium is found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (2) For the 81% T.D. specimens, which are annealed after irradiation at 630 K for 4 h, the fraction of residual tritium is also found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (3) For the 81% T.D. specimens as irradiated, the tritium release rate is found to follow the square-root law of the annealing time. The rate-determining process is controlled by Kohlrauch stretched exponential form. Tritium trapped in irradiation defects released with recovering the defects by isothermal heating

Degradation mechanisms of optoelectric properties of GaN via highly-charged 209Bi33+ ions irradiation

Science.gov (United States)

Zhang, L. Q.; Zhang, C. H.; Xian, Y. Q.; Liu, J.; Ding, Z. N.; Yan, T. X.; Chen, Y. G.; Su, C. H.; Li, J. Y.; Liu, H. P.

2018-05-01

N-type gallium nitride (GaN) epitaxial layers were subjected to 990-keV Bi33+ ions irradiation to various fluences. Optoelectric properties of the irradiated-GaN specimens were studied by means of Raman scattering and variable temperature photoluminescence (PL) spectroscopy. Raman spectra reveal that both the free-carrier concentration and its mobility generally decrease with a successive increase in ion fluence. Electro-optic mechanisms dominated the electrical transport to a fluence of 1.061 × 1012 Bi33+/cm2. Above this fluence, electrical properties were governed by the deformation potential. The appearance of vacancy-type defects results in an abrupt degradation in electrical transports. Varying temperature photoluminescence (PL) spectra display that all emission lines of 1.061 × 1012 Bi33+/cm2-irradiated specimen present a general remarkable thermal redshift, quenching, and broadening, including donor-bound-exciton peak, yellow luminescence band, and LO-phonon replicas. Moreover, as the temperature rises, a transformation from excitons (donor-acceptor pairs' luminescence) to band-to-band transitions (donor-acceptor combinations) was found, and the shrinkage effect of the band gap dominated the shift of the peak position gradually, especially the temperature increases above 150 K. In contrast to the un-irradiated specimen, a sensitive temperature dependence of all photoluminescence (PL) lines' intensity obtained from 1.061 × 1012 Bi33+/cm2-irradiated specimen was found. Mechanisms underlying were discussed.

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

International Nuclear Information System (INIS)

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

2004-11-01

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

Characterization of BOR-60 Irradiated 14YWT-NFA1 Tubes

Energy Technology Data Exchange (ETDEWEB)

Saleh, Tarik A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Quintana, Matthew Estevan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Tobias J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-02-15

Tubes of FCRD 14YWT-NFA1 Alloy were placed in the BOR-60 reactor and irradiated under a fast flux neutron environment to two conditions: 7 dpa at 360-370 °C and 6 dpa at 385-430 °C. Small sections of the tube were cut and sent to UC Berkeley for nanohardness testing and focused ion beam (FIB) milling of TEM specimens. FIB specimens were sent back to LANL for final FIB milling and TEM imaging. Hardness data and TEM images are presented in this report. This is the first fast reactor neutron irradiated information on the 14YWT-NFA1 alloy.

The influence of low dose irradiation on the creep properties of type 316 welds

International Nuclear Information System (INIS)

Marshall, P.; Steeds, J.W.; Lin, Y.P.; Finlan, G.T.

1987-01-01

Fully instrumented creep and stress rupture tests have been performed at 873K for times up to 20,000h on a series of type 316 steel/17Cr 8Ni 2Mo weld metal specimens in the unirradiated and thermal neutron irradiated conditions. The specimens tested included all weld metal longitudinal and transverse composites in the as-welded condition and following a stress relief heat treatment of 10h at 1075K. Simulated heat affected zone (HAZ) specimens were also tested. Analysis of the creep results combined with metallography, autoradiography and TEM established that the decrease in properties of irradiated samples is caused by an increasing secondary strain rate due to enhanced helium induced grain boundary fracture of the simulated HAZ and enhanced interdendritic fracture in the weld metal. Implications of strength reductions on the design of welded structures subjected to thermal irradiation are briefly assessed. (author)

Defects investigation in neutron irradiated reactor steels by positron annihilation

International Nuclear Information System (INIS)

Slugen, V.

2003-01-01

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

Influence of specimen design on the deformation and failure of zircaloy cladding

International Nuclear Information System (INIS)

Bates, D.W.; Koss, D.A.; Motta, A.T.; Majumdar, S.

2000-01-01

Experimental as well as computational analyses have been used to examine the deformation and failure behavior of ring-stretch specimens of Zircaloy-4 cladding tubes. The results show that, at least for plastically anisotropic unirradiated cladding, specimens with a small gauge length l to width w ratio (l/w ∼ 1) exhibit pronounced non-uniform deformation along their length. As a result, specimen necking occurs upon yielding when the specimen is fully plastic. Finite element analysis indicates a minimum l/w of 4 before a significant fraction of the gauge length deforms homogeneously. A brief examination of the contrasting deformation and failure behavior between uniaxial and plane-strain ring tension tests further supports the use of the latter geometry for determining cladding failure ductility data that are relevant to certain reactivity-initiated accident conditions

Effect on fast neutron irradiation to 4 dpa at 400{degrees}C on the properties of V-(4-5)Cr-(4-5)Ti alloys

Energy Technology Data Exchange (ETDEWEB)

Zinkle, S.J.; Alexander, D.J.; Robertson, J.P. [Oak Ridge National Lab., TN (United States)] [and others

1997-04-01

Tensile, Charpy impact and electrical resistivity measurements have been performed at ORNL on V-4Cr-4Ti and V-5Cr-5Ti specimens that were prepared at ANL and irradiated in the lithium-bonded X530 experiment in the EBR-II fast reactor. All of the specimens were irradiated to a damage level of about 4 dpa at a temperature of {approximately}400{degrees}C. A significant amount of radiation hardening was evident in both the tensile and Charpy impact tests. The irradiated V-4Cr-4Ti yield strength measured at {approximately}390{degrees}C was >800 MPa, which is more than three times as high as the unirradiated value. The uniform elongations of the irradiated tensile specimens were typically {approximately}1%, with corresponding total elongations of 4-6%. The ductile to brittle transition temperature of the irradiated specimens was less than the unirradiated resistivity, which suggests that hardening associated with interstitial solute pickup was minimal.

Development of reconstitution technique of irradiated specimen. 2. Annual report for FY1994 on JAERI-IHI cooperated research program

Energy Technology Data Exchange (ETDEWEB)

Nishiyama, Yutaka; Fukaya, Kiyoshi; Onizawa, Kunio; Suzuki, Masahide; Shibata, Katsuyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaihara, Shoichiro; Nakamura, Terumi; Sato, Akira; Yoshida, Kazuo

1996-02-01

A surface-activated joining method to construct Charpy impact specimens from a limited volume of broken specimens is being developed. The method is likely to decrease the thermal input led to annealing and metallurgical changes. This paper describes the technical qualification process of the joining parameters and surface configuration of joined specimens. All tests have been done with A533B cl.1. The joining machine with higher vacuum than that previously used was prepared for the tests. Precise control of joining parameters led to heat-affected zone as small as 1mm in each side. In the case of joining the square shaped (10x10mm) and circular shaped ({phi} 16mm) specimens, overall joining was achieved by an attached envelope to the square shaped specimen. In addition, the grooved surface of the circular shaped specimen brought out uniformly distributed heat-affected zone. The specification of hot-use joining machine which involves the joining sequence and restrictions of the dimension was also examined. (author).

Effect of Heat Flux on the Specimen Temperature of an LBE Capsule

International Nuclear Information System (INIS)

Kang, Y. H.; Park, S. J.; Cho, M. S.; Choo, K. N.; Lee, Y. S.

2011-01-01

For application of high-temperature irradiation tests in the HANARO reactor for Gen IV reactor material development, a number of newly designed LBE capsules have been investigated at KAERI since 2008. Recent study on heat transfer experiment of an LBE capsule with a single heater has shown that the specimen temperature of the mock-up increased linearly with an increase of heat input. The work highlighted only the heat transfer capability of an LBE capsule with a single heater as a simulated specimen in a liquid metal medium. Hence, a new LBE capsule with multi specimen sets has been designed and fabricated for the heat transfer experiment of an LBE capsule of 11M-01K. In this paper, a series of thermal analyses and heat transfer experiments for a newly designed LBE capsule was implemented to study the effect of an increase in the value of heat input and its influence on temperature distribution in the capsule mock-up

Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

International Nuclear Information System (INIS)

Dethloff, Christian; Gaganidze, Ermile; Svetukhin, Vyacheslav V.; Aktaa, Jarir

2012-01-01

Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different 10 B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

Energy Technology Data Exchange (ETDEWEB)

Dethloff, Christian, E-mail: christian.dethloff@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gaganidze, Ermile [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Svetukhin, Vyacheslav V. [Ulyanovsk State University, Leo Tolstoy Str. 42, 432970 Ulyanovsk (Russian Federation); Aktaa, Jarir [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2012-07-15

Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different {sup 10}B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

Continuous in-situ measurements of fission fragment irradiation induced void swelling in Ni

International Nuclear Information System (INIS)

Lefakis, H.

1980-01-01

A novel simulation technique has been developed to study the early stages of irradiation induced void formation in metals. The technique makes use of fission fragment irradiation produced by doping with 235 U and irradiating in a thermal neutron flux under highly controlled irradiation-environmental conditions. Employment of a computer and a high temperature radiation resistant LVDT resulted in a high volumetric sensitivity and the production of continuous, in-situ void swelling data for bulk specimens. Results for Ni, used as a test-metal served to corroborate the technique in a number of ways including comparisons with (a) reactor data, (b) direct post-irradiation specimen length measurements and (c) TEM examinations of irradiated samples. The technique has several unique advantages and, in conjunction with other conventional methods, it offers the possibility of detailed evaluation of void nucleation and growth theories. In view of the present results no definitive answer may be given on the issue of the incubation period while checks with two theoretical models have yielded an order-of-magnitude agreement

Study of the Effect of Swelling on Irradiation Assisted Stress Corrosion Cracking

Energy Technology Data Exchange (ETDEWEB)

Teysseyre, Sebastien Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

This report describes the methodology used to study the effect of swelling on the crack growth rate of an irradiation-assisted stress corrosion crack that is propagating in highly irradiated stainless steel 304 material irradiated to 33 dpa in the Experimental Breeder Reactor-II. The material selection, specimens design, experimental apparatus and processes are described. The results of the current test are presented.

Statistical analyses of fracture toughness results for two irradiated high-copper welds

International Nuclear Information System (INIS)

Nanstad, R.K.; McCabe, D.E.; Haggag, F.M.; Bowman, K.O.; Downing, D.J.

1990-01-01

The objectives of the Heavy-Section Steel Irradiation Program Fifth Irradiation Series were to determine the effects of neutron irradiation on the transition temperature shift and the shape of the K Ic curve described in Sect. 6 of the ASME Boiler and Pressure Vessel Code. Two submerged-arc welds with copper contents of 0.23 and 0.31% were commercially fabricated in 215-mm-thick plates. Charpy V-notch (CVN) impact, tensile, drop-weight, and compact specimens up to 203.2 mm thick [1T, 2T, 4T, 6T, and 8T C(T)] were tested to provide a large data base for unirradiated material. Similar specimens with compacts up to 4T were irradiated at about 288 degrees C to a mean fluence of about 1.5 x 10 19 neutrons/cm 2 (>1 MeV) in the Oak Ridge Research Reactor. Both linear-elastic and elastic-plastic fracture mechanics methods were used to analyze all cleavage fracture results and local cleavage instabilities (pop-ins). Evaluation of the results showed that the cleavage fracture toughness values determined at initial pop-ins fall within the same scatter band as the values from failed specimens; thus, they were included in the data base for analysis (all data are designated K Jc )

Irradiation of defected SAP clad UO2 fuel in the X-7 organic loop

International Nuclear Information System (INIS)

Robertson, R.F.S.; Cracknell, A.G.; MacDonald, R.D.

1961-10-01

This report describes an experiment designed to test the behaviour under irradiation of a UO 2 fuel specimen clad in a defected SAP sheath and cooled by recirculating organic liquid. The specimen containing the defect was irradiated in the X-7 loop in the NRX reactor from the 25th of November until the 13th of December 1960. Up to the 13th of December the behaviour was analogous to that seen with defected UO 2 specimens clad in zircaloy which were irradiated in water loops. Reactor power transients resulted in peaking of gamma ray activities in the loop, but on steady operation these activities tended to fall to a steady state level, Over this period the pressure drop across the fuel increased by a factor of two, the increases occurring after reactor shut downs and start ups. On 13th December the pressure drop increased rapidly, after a reactor shut down and start up, to over five times its original value and the activities in the loop rose to a high level. The specimen was removed and examination showed that the sheath was very badly split and that the volume between the fuel and the sheath was filled with a hard black organic substance. This report gives full details of the irradiation and of the post -irradiation examination. Correlation of the observed phenomenon is attempted and a preliminary assessment of the problems which would be associated with defect fuel in an organic reactor is given. (author)

Preliminary investigation of candidate specimens for the Egyptian environmental specimen bank

International Nuclear Information System (INIS)

Shawky, S.; Amer, H.; Schladot, J.D.; Ostapczuk, P.; Emons, H.; Abou El-Nour, F.

2000-01-01

In the frame of establishing an environmental monitoring program related to environmental specimen banking in egypt, some candidate specimens from the aquatic environment (Fish muscle, fish liver; mussels) were investigated. The selection of specimens and sampling sites is described. Specimens are chemically characterised with respect to some major and trace elements and the results are compared with data obtained from comparable specimens collected in aquatic ecosystems of germany

Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

International Nuclear Information System (INIS)

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

1977-01-01

A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this observation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. The diffusion tests will be performed at the Euratom Joint Research Centre in Ispra, Italy, and the irradiation creep tests will be carried out in the High Flux Reactor /9/ of the Euratom Joint Research Centre in Petten, The Netherlands. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. In the first type of rig about 50 samples can be tested uniaxially under tension with various combinations of irradiation temperature and stress. The second type of rig holds up to 70 samples which are tested in bending, again with various combinations of irradiation temperature and stress

Steam-chemical reactivity for irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Anderl, R.A.; McCarthy, K.A.; Oates, M.A.; Petti, D.A.; Pawelko, R.J.; Smolik, G.R. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States)

1998-01-01

This paper reports the results of an experimental investigation to determine the influence of neutron irradiation effects and annealing on the chemical reactivity of beryllium exposed to steam. The work entailed measurements of the H{sub 2} generation rates for unirradiated and irradiated Be and for irradiated Be that had been previously annealed at different temperatures ranging from 450degC to 1200degC. H{sub 2} generation rates were similar for irradiated and unirradiated Be in steam-chemical reactivity experiments at temperatures between 450degC and 600degC. For irradiated Be exposed to steam at 700degC, the chemical reactivity accelerated rapidly and the specimen experienced a temperature excursion. Enhanced chemical reactivity at temperatures between 400degC and 600degC was observed for irradiated Be annealed at temperatures of 700degC and higher. This reactivity enhancement could be accounted for by the increased specific surface area resulting from development of a surface-connected porosity in the irradiated-annealed Be. (author)

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Annealing of dislocation loops in neutron-irradiated copper investigated by positron annihilation

International Nuclear Information System (INIS)

Gauster, W.B.; Mantl, S.; Schober, T.; Triftshauser, W.

1975-01-01

Positron annihilation angular correlation measurements were carried out on neutron-irradiated copper as a function of annealing temperature. Two types of specimens were used: single crystals irradiated with fast neutrons, and 10 B-doped polycrystalline samples irradiated with thermal neutrons. All irradiations were at approximately 320 0 K. A structure in the annealing curve, not previously observed by other techniques, indicates that between 460 and 600 0 K the dislocation loops present after irradiation dissociate and more effective positron trapping sites are formed. (auth)

The effects of irradiance and exposure time on the surface roughness of bulk-fill composite resin restorative materials

Science.gov (United States)

Alkhudhairy, Fahad I.

2018-01-01

Objectives: To evaluate the surface roughness of 4 different bulk-fill resin-based composites cured using different irradiance levels. Methods: This in vitro study was performed in February 2017 to August 2017 at the College of Dentistry, King Saud University. Twenty-four specimens were prepared from each of the bulk-fill materials [Tetric N-Ceram (TNC), SonicFill (SF), Smart Dentin Replacement (SDR), and Filtek Bulk-Fill (FB)] using a brass metal mold, resulting in a total of 96 specimens, cured using a Bluephase N light curing unit. Half of the total number of specimens (N=48) were cured using high-power irradiance (1200 mW/cm2) for 20 seconds, while the remaining half (N=48) were cured using low power irradiance (650 mW/cm2) for 40 seconds. After 24 hours, baseline surface roughness of each specimen was analyzed using a profilometer, then polished using Sof-lex abrasive disks, and the surface roughness of all groups was assessed. Results: Post-polished SonicFill cured at high irradiance had the highest mean surface roughness (0.23±0.03), whereas pre-polished Smart Dentin Replacement (0.11±0.01) and SonicFill (0.11±0.02) cured at low irradiance had the lowest mean surface roughness. Conclusion: High curing irradiance (1,200 mW/cm2) had no positive influence on the surface roughness of Filtek Bulk Fill and Tetric N-Ceram bulk-fill RBCs compared with lower curing irradiance (650 mW/cm2). However, the difference of curing irradiance significantly affected the surface roughness in SDR and sonic fill RBCs. PMID:29436570

Spectroscopy of electron irradiated polymers in electron microscope

International Nuclear Information System (INIS)

Faraj, S.H.; Salih, S.M.

1981-01-01

The damage induced by energetic electrons in the course of irradiation of polymers in a transmission electron microscope was investigated spectroscopically. Damage on the molecular level has been detected at very low exposure doses. These effects have been induced by electron doses less than that received by the specimen when it is situated at its usual place of the specimen stage in the electron microscope by a factor of 1,000. (author)

Irradiation of copper alloys in FFTF

International Nuclear Information System (INIS)

Brager, H.R.; Garner, F.A.

1984-01-01

Nine copper-base alloys in thirteen material conditions have been inserted into the MOTA-18 experiment for irradiation in FFTF at approx.450 0 C. The alloy Ni-1.9Be is also included in this experiment, which includes both TEM disks and miniature tensile specimens

Microstructures of beta-silicon carbide after irradiation creep deformation at elevated temperatures

International Nuclear Information System (INIS)

Katoh, Yutai; Kondo, Sosuke; Snead, Lance L.

2008-01-01

Microstructures of silicon carbide were examined by transmission electron microscopy (TEM) after creep deformation under neutron irradiation. Thin strip specimens of polycrystalline and monocrystalline, chemically vapor-deposited, beta-phase silicon carbide were irradiated in the high flux isotope reactor to 0.7-4.2 dpa at nominal temperatures of 640-1080 deg. C in an elastically pre-strained bend stress relaxation configuration with the initial stress of ∼100 MPa. Irradiation creep caused permanent strains of 0.6 to 2.3 x 10 -4 . Tensile-loaded near-surface portions of the crept specimens were examined by TEM. The main microstructural features observed were dislocation loops in all samples, and appeared similar to those observed in samples irradiated in non-stressed conditions. Slight but statistically significant anisotropy in dislocation loop microstructure was observed in one irradiation condition, and accounted for at least a fraction of the creep strain derived from the stress relaxation. The estimated total volume of loops accounted for 10-45% of the estimated total swelling. The results imply that the early irradiation creep deformation of SiC observed in this work was driven by anisotropic evolutions of extrinsic dislocation loops and matrix defects with undetectable sizes

Correlations between Standard and Miniaturised Charpy-V Specimens

International Nuclear Information System (INIS)

Lucon, E.; Van Walle, E.; Fabry, A.; Puzzolante, J.-L.; Verstrepen, A.; Vosch, R.; Van de Velde, L.

1998-12-01

A total of 565 instrumented impact tests (232 performed on full-size and 333 on sub-size Charpy-V specimens) have been analysed in order to derive meaningful assumptions on the correlations existing between test results obtained on specimens of different size. Nine materials (pressure vessel steels) have been considered, in both as-received and irradiated state, for a total of 19 conditions examined. For the analysis of data, conventional as well novel approaches have been investigated; former ones, based on a review of the existing literature, include predictions of USE values by the use of normalization factors (NF), shifts of index temperatures related to energy/lateral expansion/shear fracture levels, and a combination of both approaches (scaling and shifting of energy curves). More original and recent proposals have also been verified, available in the literature but also proposed by SCK-CEN in the frame of enhanced surveillance of nuclear reactor pressure vessels. Conclusions have been drawn regarding the applicability and reliability of these methodologies, and recommendations have been given for future developments of the activities on this topic

Awareness on food irradiation among students

International Nuclear Information System (INIS)

Seri Chempaka Mohd Yusof; Foziah Ali; Salahbiah Abdul Majid; Ros Anita Ahmad Ramli; Zainab Harun

2009-01-01

This survey was conducted to determine the level of understanding on radiation and irradiated food products amongst students during an exhibition in conjunction with Nuclear Malaysia Innovation Day 2008, on 16-18 July 2008 at Dewan Tun Dr. Ismail, Malaysian Nuclear Agency. Data were collected from 180 respondents comprising students from various schools visiting the exhibition. The results revealed that 55.56 % of the respondents knew of radiation and 81.11 % agreed that food could be irradiated. However, 53.33 % respondents misunderstood that there was presence of radioactivity in the food after irradiation. The results also showed that respondents knew that various foods can be irradiated and the type of radiation used in irradiation of food products. This survey indicates more aggressive work must be done to educate and introduce the public the application of nuclear technology in modern life, including in food preservation. (Author)

Influence of tensile stress on cavity growth in nickel under helium irradiation

International Nuclear Information System (INIS)

Kusanagi, Hideo; Hide, Koichiro; Takaku, Hiroshi

1989-01-01

The influence of tensile stress on cavity behavior in pure nickel under helium irradiation was investigated by in-situ observation using the transmission electron microscope (TEM) in which an ion gun is installed. Specimens were irradiated at 500 0 C with 20 keV helium in the TEM. The dose rate was about 10 14 He/cm 2 s, and the angle between the helium beam and the normal direction of the specimens was about 60 0 . The damage rate estimated by the E-DEP-1 code was about 0.6x10 -3 dpa/s at its peak position. The main results are as follows: (1) cavity nucleation was accelerated by applying tensile stress, and cavity size in stressed specimens was several times larger than that in stress-free specimens; (2) cavity density in the stressed specimen increased more rapidly than in the stress-free specimen, and then decreased by cavity coalescences; (3) depth of cavity nucleation in the stress-free specimen was about 160 nm, while that in the stressed specimen was about 320 nm; that is, cavities nucleated in deeper regions in the stressed specimen than in the stress-free specimen. This result indicates that helium atoms and vacancies can migrate into the deeper region by applying tensile stress. (4) The experimental results obtained in this study can be explained qualitatively by the mechanism that mobile dislocations drag He-V complexes to the deeper region. This implies that there are similar phenomena in the case of compressive stress. (orig.)

Irradiation and annealing behavior of 15Kh2MFA reactor pressure vessel steel

International Nuclear Information System (INIS)

Popp, K.; Bergmann, U.; Bergner, F.; Hampe, E.; Leonhardt, W.D.; Schuetzler, H.P.; Viehrig, H.W.

1992-01-01

This work deals with the mechanical properties of RPV steels used WWER-440. The materials under investigation were a forging (base metal 15Kh2MFA) and the corresponding weld. Charpy V-notch specimens and tensile test specimens were irradiated in the WWER-2 Rheinsberg at about 270 C up to the two neutron fluence levels of 4 x 10 18 and 5 x 10 19 n/cm 2 (E>1MeV). Post-irradiation annealing heat treatments were performed, among others a 475 C/152 h treatment of technical interest. (orig.)

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

Hardening of ODS ferritic steels under irradiation with high-energy heavy ions

Science.gov (United States)

Ding, Z. N.; Zhang, C. H.; Yang, Y. T.; Song, Y.; Kimura, A.; Jang, J.

2017-09-01

Influence of the nanoscale oxide particles on mechanical properties and irradiation resistance of oxide-dispersion-strengthened (ODS) ferritic steels is of critical importance for the use of the material in fuel cladding or blanket components in advanced nuclear reactors. In the present work, impact of structures of oxide dispersoids on the irradiation hardening of ODS ferritic steels was studied. Specimens of three high-Cr ODS ferritic steels containing oxide dispersoids with different number density and average size were irradiated with high-energy Ni ions at about -50 °C. The energy of the incident Ni ions was varied from 12.73 MeV to 357.86 MeV by using an energy degrader at the terminal so that a plateau of atomic displacement damage (∼0.8 dpa) was produced from the near surface to a depth of 24 μm in the specimens. A nanoindentor (in constant stiffness mode with a diamond Berkovich indenter) and a Vickers micro-hardness tester were used to measure the hardeness of the specimens. The Nix-Gao model taking account of the indentation size effect (ISE) was used to fit the hardness data. It is observed that the soft substrate effect (SSE) can be diminished substantially in the irradiated specimens due to the thick damaged regions produced by the Ni ions. A linear correlation between the nano-hardeness and the micro-hardness was found. It is observed that a higher number density of oxide dispersoids with a smaller average diameter corresponds to an increased resistance to irradiation hardening, which can be ascribed to the increased sink strength of oxides/matrix interfaces to point defects. The rate equation approach and the conventional hardening model were used to analyze the influence of defect clusters on irradiation hardening in ODS ferritic steels. The numerical estimates show that the hardening caused by the interstitial type dislocation loops follows a similar trend with the experiment data.

Enhanced low-temperature oxidation of zirconium alloys under irradiation

International Nuclear Information System (INIS)

Cox, B.; Fidleris, V.

1989-01-01

The linear growth of relatively thick (>300 nm) interference-colored oxide films on zirconium alloy specimens exposed in the Advanced Test Reactor (ATR) coolant at ≤55 o C was unexpected. Initial ideas were that this was a photoconduction effect. Experiments to study photoconduction in thin anodic zirconium oxide (ZrO 2 ) films in the laboratory were initiated to provide background data. It was found that, in the laboratory, provided a high electric field was maintained across the oxide during ultraviolet (UV) irradiation, enhanced growth of oxide occurred in the irradiated area. Similarly enhanced growth could be obtained on thin thermally formed oxide films that were immersed in an electrolyte with a high electric field superimposed. This enhanced growth was found to be caused by the development of porosity in the barrier oxide layer by an enhanced local dissolution and reprecipitation process during UV irradiation. Similar porosity was observed in the oxide films on the ATR specimens. Since it is not thought that a high electric field could have been present in this instance, localized dissolution of fast-neutron primary recoil tracks may be the operative mechanism. In all instances, the specimens attempt to maintain the normal barrier-layer oxide thickness, which causes the additional oxide growth. Similar mechanisms may have operated during the formation of thick loosely adherent, porous oxides in homogeneous reactor solutions under irradiation, and may be the cause of enhanced oxidation of zirconium alloys in high-temperature water-cooled reactors in some water chemistries. (author)

An inverse method based on finite element model to derive the plastic flow properties from non-standard tensile specimens of Eurofer97 steel

Directory of Open Access Journals (Sweden)

S. Knitel

2016-12-01

Full Text Available A new inverse method was developed to derive the plastic flow properties of non-standard disk tensile specimens, which were so designed to fit irradiation rods used for spallation irradiations in SINQ (Schweizer Spallations Neutronen Quelle target at Paul Scherrer Institute. The inverse method, which makes use of MATLAB and the finite element code ABAQUS, is based upon the reconstruction of the load-displacement curve by a succession of connected small linear segments. To do so, the experimental engineering stress/strain curve is divided into an elastic and a plastic section, and the plastic section is further divided into small segments. Each segment is then used to determine an associated pair of true stress/plastic strain values, representing the constitutive behavior. The main advantage of the method is that it does not rely on a hypothetic analytical expression of the constitutive behavior. To account for the stress/strain gradients that develop in the non-standard specimen, the stress and strain were weighted over the volume of the deforming elements. The method was validated with tensile tests carried out at room temperature on non-standard flat disk tensile specimens as well as on standard cylindrical specimens made of the reduced-activation tempered martensitic steel Eurofer97. While both specimen geometries presented a significant difference in terms of deformation localization during necking, the same true stress/strain curve was deduced from the inverse method. The potential and usefulness of the inverse method is outlined for irradiated materials that suffer from a large uniform elongation reduction.

Quality evaluation of garlic irradiated in Argentine and stored in Brazil

International Nuclear Information System (INIS)

Curzio, Osvaldo A.; Croci, Clara A.; Domarco, Rachel E.; Spoto, Marta H.F.; Blumer, Lucimara; Walder, Julio M.M.

1997-01-01

This work was undertaken to evaluate the quality of garlic Colorado, irradiated in Argentina and stored for long period of time under environmental conditions in Brazil. Two samples of 100 kg each were selected from high quality garlic harvested in December/1995 from a region close to the Universidad Nacional del Sur. At 30 days after harvesting one of the samples were irradiated with a dose of 60 Gy. Both samples, irradiated and non irradiated, were transported by road from the Laboratorio de Radioisotopos of the Universidad Nacional del Sur, Bahia Blanca, Argentina, to the Centro de Energia Nuclear na Agriculture (CENA-USP), Piracicaba, Sao Paulo, Brazil. The effects of irradiation on weigh loss , discard, germination and sensorial analysis were nonthly observed on CENA among 30and 180 days of storage. The evaluation of germination evidences the benefits of the radioinhibition process. The irradiated bulbs did not exhibit any intern bud, however the non-irradiated bulbs exhibit any intern bud, however the non-irradiated bulbs exhibit any intern bud, however the non-irradiated bulbs exhibit 100% of germination on the period of storage. At the end of the period of storage, the weigh loss of the irradiated garlic was smaller than the non-irradiate one. The percentage of discard, evaluated as intern germination, scatying and whitering, was 43% on the irradiated sample. The analysis of sensorial parameters shows no difference between irradiated and non-irradiated garlic, so irradaition did not affect the sensorial quality of the product. (author). 12 refs., 8 tabs

Anomalous fracture toughness of irradiated Cr-MoV - Reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Ahistrand, R [Imatran Voima Oy (IVO), Helsinki (Finland)

1994-12-31

The base metal Crack Opening Displacement (COD) specimens of the irradiation-induced embrittlement surveillance programme in Loviisa 1 revealed an anomalous behaviour of K{sub JC} compared to the Charpy-V results and to expected results according to standards: about 20% of the COD specimens showed an exceptionally low fracture toughness. Abnormal test specimens were analyzed through fractography, metallography and repeated tests using reconstitution technique: the anomalous behaviour appears to be caused by incorrect pre-fatigue cracking of base metal COD specimens. 7 refs., 9 figs.

Optical properties and light irradiance of monolithic zirconia at variable thicknesses.

Science.gov (United States)

Sulaiman, Taiseer A; Abdulmajeed, Aous A; Donovan, Terrence E; Ritter, André V; Vallittu, Pekka K; Närhi, Timo O; Lassila, Lippo V

2015-10-01

The aims of this study were to: (1) estimate the effect of polishing on the surface gloss of monolithic zirconia, (2) measure and compare the translucency of monolithic zirconia at variable thicknesses, and (3) determine the effect of zirconia thickness on irradiance and total irradiant energy. Four monolithic partially stabilized zirconia (PSZ) brands; Prettau® (PRT, Zirkonzahn), Bruxzir® (BRX, Glidewell), Zenostar® (ZEN, Wieland), Katana® (KAT, Noritake), and one fully stabilized zirconia (FSZ); Prettau Anterior® (PRTA, Zirkonzahn) were used to fabricate specimens (n=5/subgroup) with different thicknesses (0.5, 0.7, 1.0, 1.2, 1.5, and 2.0mm). Zirconia core material ICE® Zircon (ICE, Zirkonzahn) was used as a control. Surface gloss and translucency were evaluated using a reflection spectrophotometer. Irradiance and total irradiant energy transmitted through each specimen was quantified using MARC® Resin Calibrator. All specimens were then subjected to a standardized polishing method and the surface gloss, translucency, irradiance, and total irradiant energy measurements were repeated. Statistical analysis was performed using two-way ANOVA and post-hoc Tukey's tests (pgloss was significantly affected by polishing (p<0.05), regardless of brand and thickness. Translucency values ranged from 5.65 to 20.40 before polishing and 5.10 to 19.95 after polishing. The ranking from least to highest translucent (after polish) was: BRX=ICE=PRTirradiant energy was: BRXirradiant energy, and thickness of zirconia and the amount was brand dependent (p<0.05). Brand selection, thickness, and polishing of monolithic zirconia can affect the ultimate clinical outcome of the optical properties of zirconia restorations. FSZ is relatively more polishable and translucent than PSZ. Copyright © 2015 Academy of Dental Materials

Neutron irradiation effect of thermally-sensitized stainless steels

Energy Technology Data Exchange (ETDEWEB)

Hide, Kouitiro [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

1998-03-01

Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated thermally-sensitized Type 304 Stainless Steels (SSs) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 1.1 x 10{sup 24} n/m{sup 2} (E > 1MeV) at the light water reactor temperature in the Japan Material Test Reactor. The irradiated specimens were examined by slow strain rate stress corrosion cracking tests in 290degC pure water of 0.2 ppm dissolved oxygen concentration and microhardness measurements. The IGSCC susceptibility of the irradiated specimens increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}. From an attempt to correlate the IGSCC susceptibility with the mechanical properties, an excellent correlation was identified between the susceptibility and microhardness increments at the grain boundary relative to the grain center. While intergranular corrosion rate of thermally sensitized SS increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}, that of solution annealed SS did not change. The incremental grain boundary hardening and degradation of intergranular corrosion resistance may presumably be the major factors affecting IGSCC performance. (author)

Response of irradiated diet fed rats to whole body X irradiation

International Nuclear Information System (INIS)

Hasan, S.S.; Kushwaha, A.K.S.

1985-01-01

The response to whole body X irradiation has been studied in the brain of rats fed both on a normal diet (consisting of equal parts of wheat and gram flour) and on a low protein irradiated diet (consisting of a part of normal diet and three parts of wheat). The activity of enzymes related to the glucose metabolism (glucose 6-phosphate dehydrogenase and fructose diphosphate aldolase) is reduced, while that of peroxidant enzymes (catalase and lipid peroxidase) increased in the brain of rats that received a diet poor in proteins and irradiated diets (normal or hypoproteic). DNA and RNA levels and protein content show a significant reduction in the brain of rats with hypoproteic and irradiated diets. The total body irradiation causes serious alterations in the brain in animals with a hypoproteic malnutritions due both to a low protein and an irradiated diet. The brain of rats fed on a low protein and irradiated diet exhibits after whole body irradiation damages more severe than those in rats fed on a normal irradiated diet

Augmentation of thermoelectric performance of VO2 thin films irradiated by 200 MeV Ag9+-ions

International Nuclear Information System (INIS)

Khan, G.R.; Kandasami, A.; Bhat, B.A.

2016-01-01

Swift Heavy Ion (SHI) irradiation with 200 MeV Ag 9+ -ion beam at ion fluences of 1E11, 5E11, 1E12, and 5E12 for tuning of electrical transport properties of VO 2 thin films fabricated by so–gel technique on alumina substrates has been demonstrated in the present paper. The point defects created by SHI irradiation modulate metal to insulator phase transition temperature, carrier concentration, carrier mobility, electrical conductivity, and Seebeck coefficient of VO 2 thin films. The structural properties of the films were characterized by XRD and Raman spectroscopy and crystallite size was found to decrease upon irradiation. The atomic force microscopy revealed that the surface roughness of specimens first decreased and then increased with increasing fluence. Both resistance as well as Seebeck coefficient measurements demonstrated that all the samples exhibit metal–insulator phase transition and the transition temperatures decreases with increasing fluence. Hall effect measurements exhibited that carrier concentration increased continuously with increasing fluence which resulted in an increase of electrical conductivity by several orders of magnitude in the insulating phase. Seebeck coefficient in insulating phase remained almost constant in spite of an increase in the electrical conductivity by several orders of magnitude making SHI irradiation an alternative stratagem for augmentation of thermoelectric performance of the materials. The carrier mobility at room temperature decreased up to the beam fluence of 5E11 and then started increasing whereas Seebeck coefficient in metallic state first increased with increasing ion beam fluence up to 5E11 and thereafter decreased. Variation of these electrical transport parameters has been explained in detail. - Highlights: • Thermoelectric properties of VO 2 thin films enhance upon SHI irradiation. • Structural properties show that crystallite size decrease upon SHI irradiation. • Metal–insulator phase